written historical and descriptive data reduced

HISTORIC AMERICAN ENGINEERING RECORD ~-

Location:

Date of Construction:

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Present Owners:

Present Use:

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Project Information:

HADDAM NECK NUCLEAR POWER PLANT (Connecticut Yankee Nuclear Power Plant)

HAER No. CT-185

362 Injun Hollow Road Haddam Middlesex County Connecticut

U.S. Geological Survey Deep River Quadrangle UTM Coordinates 18.708837.80 E -4595337.91 N +41 o 28' 56.866" latitude, -72° 29' 54.983" longitude1

1964-1966

Westinghouse Electric Company Stone & Webster Engineering Corporation

Connecticut Yankee Atomic Power Company (CY APCO) 362 Injun Hollow Road Haddam Neck CT 06424-3022

Demolished with some foundations left in place.

The Haddam Neck Power Plant was one of the earliest commercial-scale nuclear power stations in the United States and the first completed on the East Coast. During its operating history from 1967 to 1996, this plant established several records in electrical production. The plant was eligible for the National Register of Historic Places.

CY APCO ceased electrical generation at the Haddam Neck plant in 1996. Decommissioning operations started in 1998, subject to authority of the Nuclear Regulatory Commission (NRC). NRC authority brought the protection under the purview of federal acts and regulation protecting significant cultural

1 The Haddam Neck Nuclear Power Plant was located at latitude +41 o 28' 56.866", longitude -72° 29' 54.983". The coordinate represents the center point of the former reactor containment building. This coordinate was obtained on November 4, 2009 using a GPS unit accurate to+!- 5 meters. The coordinates were compared to values obtained on the Google Earth website and USGS Deep River Quadrangle and the accuracy ofthe coordinates is+/- 15 meters.

resources from adverse project effects. 2 This documentation was requested by the Connecticut State Historic Preservation Office to preclude the possibility of any adverse project effects.

Note: The material in this report is in large part based on Connecticut Yankee Atomic Power Company records that are archived at the University of Connecticut, Dodd Library. The records consist of plant design drawings, plant historical records, employee newsletters, environmental reports, regulatory correspondence, scrapbooks, plaques, historic photographs, and other audiovisual materials. The records are available to the public. For information contact the librarian at:

y niversity of Connecticut Thomas J. Dodd Research Center 405 Babbidge Road, Unit 1205 Storrs, Connecticut 06269-1205 860.486.4500/ 860.486.4521 (Fax) http:// dodd center . edu

Proj ect Manager and Historian Michael S. Raber Raber Associates 81 Dayton Road, P.O. Box 46 South Glastonbury, CT 06073 860/633-9026

Steam and Electric Power Historian G-erald Weinstein Photo Recording Associates 40 West 77th Street, Apt. 17b New York, NY 10024 212/431-6100

Industrial Archaeologist Robert C. Stewart Historical Technologies 1230 Copper Hill Road West Suffield, CT 06093 860/668-2928

~JJQJ~~rJ~Q}y~x CQn~!!lt~n! Gerry van N oordennen Dutchman Consulting 30 Miller Road Burlington, CT 06013

Graphics Consultant Gerry Loftus 5 Roberts Road Marlborough, CT

Connecticut Yankee Liaison John Arnold 362 Injun Hollow Road East Hampton, CT 06424

ConnecticutYankee ISFSI Manager James Lenois 362 Injun Hollow Road East Hampton, CT 06424

~ational Historic Preservation Act of 1966 (PL 89-655), the National Environmental Policy Act of 1969 (PL 91-190), the Archaeological and Historical Preservation Act (PL 93-291), Executive Order 11593, Procedures for the Protection of Historic and Cultural Properties (36 CFR Part 800).


HAERNo. CI-185 (Page 3)

Historical Information

The Haddam Neck Nuclear Power Plant had an important place in the history of American nuclear power generation as one of the first two large commercial plants in the United States, and the first completed on the East Coast. 1 The plant centered on a pressurized water reactor (PWR). In this design water, kept under pressure to prevent boiling, removed heat from the reactor core and moderated or slowed neutrons from the uranium "fuel" to energies at which the fission process could continue. Additional control was provided by control rods which could be inserted or removed from the core to absorb excess neutrons. Water passing through the core became radioactive and was cycled through a heat exchanger, called a steam generator. Water passing through a secondary system in the steam generator absorbed heat from the pressurized system. It flashed to steam which drove the turbine and generator, was condensed, then cycled back to the steam generator. During its operating history from 1967 to 1996, this plant established several records in electrical production. It typified the steam powered electric plants which combined a primary nuclear steam supply system based on two decades of post-World-War-II development, and secondary systems based on 19th and early 20th century technology for converting steam to el~ctrical energy and recycling condensed steam as feed water for the primary system. The design of the plant, and issues arising from that design, are significant examples of the limitations inherent in the first generation of American nuclear plants.

The Development of American Nuclear Steam Supply Systems, c1945-1960

Nuclear Submarines and the Beginnings of Commercial Nuclear Power

Perhaps the first electricity produced by a nuclear chain reaction occurred when one of the WorldWar -II -era reactors at Oak Ridge National Laboratory in Tennessee was fitted in the early postwar years with a boiler supplying steam to a small external turbine generator which lit up a light bulb? Most sources credit the Experimental Breeder Reactor (EBR-1), built by the Argonne National Laboratory at the National Reactor Testing Station (NRTS) in Idaho, with being the first nuclear reactor to produce electrical power in December 1951.3 Less than five years, later one of the J\TRTS reactors was used to provide all electric power for 1200 people in Areo, Idaho.4

It was, however, Cold War military planning for submarine propulsion which ultimately drove civilian American power reactor development. 5 The idea of a submarine that could travel at high speed underwater was proven by the German Type XXI U-Boat of World War II. The XXI boats had super battery plants and streamlined hulls designed to give greater submerged speed. 6 Taking that a step further, the German navy came up with a design that eliminated the air-breathing dieselcharged battery plants with their limited endurance. The Walter closed-cycle propulsion system in the Type XXVI submarines utilized hydrogen peroxide as an oxidizer to make steam which powered a turbine to drive the propeller, providing extended underwater operation. Both designs were evaluated by the U. S. and British navies after the war, but the peroxide system was unreliable


HAER No. CT -185 (Page 4)

and hazardous.7 The end of the war and the arrival of nuclear power doomed that technology. Less than six months after the close of the war, nuclear physicists were suggesting using nuclear power to drive ShipS.8 In March 1946, the Naval Research Laboratory recommended that nuclear submarines be constructed on high speed hulls based on the German XXVI types. 9 Captain (later Rear Admiral) Hyman Rickover, who had observed operations at Oak Ridge since 1946, took a lead role in the development of the nuclear submarine and American nuclear power plants.

1O As

head of the Naval Reactors Branch (Division of Reactor Development) he started working with the General Electric Company (GE) to develop a submarine reactor plant. ll GE scientists were involved with the isolation of Uranium-235 before the war, and the company took over management of the plutonium production operations at the U.S. Atomic Energy Commission's (AEC) Hanford Laboratories in Richland, Washington in 1946. 12 GE was an advocate of a Submarine Intermediate Reactor (SIR) that utilized relatively fast neutron velocity and was cooled by liquid metal sodium, Concerned about the viability of that technology, Rickover asked the Westinghouse Electric Corporation to develop an alternative design. Westinghouse had also been involved at an early stage of nuclear development as a supplier to the Manhattan Project. l3 They favored a reactor cooled with ordinary (light) water with slowed velocity (thermal type) neutrons. The genesis of both companies' designs was probably the CP-3 Heavy-Water C reactor built at Argonne in 1944 which had a core of uranium rods submerged in a water tank cooled by a heat exchanger system. 14 GE's commitments to production of weapons materials at Hanford and difficulties with the sodium system led Rickover to push Westinghouse to provide the first working plant. 15 Their Submarine Thermal Reactor (STR Mark I) propulsion system was designed at the Westinghouse Bettis Laboratory near Pittsburgh, and tested in a mock -up submarine hull at the NRTS.16 The developed power plant (STR Mark II) utilized a Pressurized Water Reactor (PWR) powered by highly-enriched U-235. The fuel in metallic form was clad with zirconium alloy forming tubes which were arranged in bundles. The nuclear chain reaction occurring in the fuel rods produced heat, and was controlled by hafnium neutron absorber rods inserted into the bundles. The coolant that took the heat out of the reactor core and moderated the reactions was ordinary

(light) water, pressurized to prevent boiling as it was pumped back and forth between the reactor and tubing in separate external heat exchanger vessels called steam generators. Feed water pumped around the heated tubes turned to steam which powered turbines geared to the propellers.

The PWR (also known as a closed-cycle water reactor)17 utilizing highly-enriched uranium was a good choice because its compact size allowed it to fit in the confines of a submarine hull. The light water was easy to handle, and the relatively low fluid and steam pressures outside of the reactor made the plant very reliable, a requirement for a vessel designed for undersea warfare. In comparison to other reactor types under development, the PWR was not very efficient, 18 but it did

CHeavy-water (D20), discovered in the 1930s, has the hydrogen atoms replaced with deuterium. It makes a good moderator and coolant and because it does not capture or waste as many neutrons as light water allowing the use offuel without enrichment (Oxford English Dictionary 1989: vA, p. 559 (Deuterium); Nero & Dennis 1984: 391).


HAERNo. CT-185 (Page 5)

not have to be commercially profitable in a naval vessel and was far better than the diesel-electric and peroxide drives that preceded it. Fears of the Soviet Union fielding a large fleet of captured German advanced U-Boats19 spurred intense, well-organized, and rapid design development among Rickover, Westinghouse and the Electric Boat Division of General Dynamics at Groton, CT. President Truman laid the keel of the first U.S. nuclear submarine, the Nautilus in 1952. The shakedown cruises in 1955 set records for underwater distances and the success of its plant was a tremendous impetus for use of that type at sea. Westinghouse subsequently supplied the reactors for the first U.S. Navy surface vessels and many of the second-generation submarines. Despite Westinghouse's success, Rickover still wanted another system as an alternative, and GE continued to work on their design. Its Submarine Intermediate Reactor (SIR)d was tested on land at Knolls Atomic Power Laboratory in Milton, New York before installation in Seawolf, launched in 1956 as the second nuclear submarine. Surplus power from that plant near Schenectady was distributed locally and may have been the first commercial electricity to be produced by nuclear energy.20 The 5'eawolfwas commissioned in 1957, but the SIR proved unreliable and was replaced by 1960 with a conventional Westinghouse PWR. 21 GE then switched to water-cooled reactors with its twin highpressure Submarine Advanced Reactors for the Triton. 22 Their later-model reactors powered a majority of the navy's aircraft carriers, cruisers and submarines.

PRESSURIZER

PRIMARY CIRCUIT

CONDENSER

MAIN TURBINE

Flow diagram for a Pressurized Water Reactor in a Submarine

,.. ,-.. . + • . : ..

dThe Nautilus reactor utilized "thermal" neutrons of reduced energy. The SIR used neutrons of intermediate energy_ Breeder reactors that create more fuel use high energy "fast" neutrons. (Weinberg and Wigner 1958: 12) Weinberg, Alvin M_ And Wigner, Eugene P. The Physical Theory of Neutron Chain Reactors. Chicago: The University of Chicago Press.



Atoms for Peace and Early Commercial Reactor Designs

The 1946 Atomic Energy Act encouraged civilian uses of nuclear power without specifying the means. President Eisenhower's Energy for Peace program and the 1954 Atomic Energy Act opened the way for military research and fissionable materials to reach civilian programs.23 In August of 1955 the International Conference on the Peaceful Uses of Atomic Energy opened in Geneva. Papers were presented on three broad categories of reactors being considered by seven nations: water-cooled, gas-cooled, and liquid-metal-cooled. Water-cooled reactors were divided into pressurized and boiling types. 24 Types of coolants were light water vs. heavy water in the water reactors, air vs. carbon dioxide in the gas reactors and sodium vs. bismuth in the liquid metal reactors. Moderators which slowed neutrons and aided the reaction included light water and heavy water. Carbon in the form of graphite was also proposed as a moderator in all but the boiling water types. The classes were further divided by types of nuclear fuel (natural vs. enrichedt and by fuel configuration (heterogeneous vs. homogeneous).f

Heterogeneous loading with uranium dioxide pellets stacked in stainless steel or zirconium alloy (Zircaloy) tubes was the most common arrangement. 25 Most of the reactors were "thermal" types that had neutrons slowed to thermal velocities. 26 An additional type, the Fast Fission Breeder that had no moderator and produced additional fuel, was also described. 27

eCivilian reactors using light water require uranium fuel in which the natural percentage ofU-235 (about 0.7 percent) has been slightly enriched to typical concentrations of 2-5 percent, less than the concentrations needed for naval submarine reactors. American civilian reactors have relied on government-owned gaseous diffusion enrichment plants; centrifuge-enrichment plants have been built in Europe, Japan, and South Africa. Reactors using heavy water, notably the Canadian Candu models discussed below, can use natural uranium without enrichment (McIntyre 1975; Power 1982b).

fIn the homogeneous fuel arrangement, the uranium or other fuel was suspended in liquid or formed into a slurry which could be pumped in and out ofthe reactor and replaced at will without shutting down the reactor. The design was adaptable to fuel breeding and recycling was an integral part of the process. The AEC decided not to pursue the concept, or its successor the molten salt breeder and instead advanced the fast breeder reactor to be built at Clinch River. (Weinberg 1994: 117-129)


HAERNo. CI-18S (Page 7)

Considering all the variables of cooling types, moderators, fuels and fuel configurations there were at least 100 feasible arrangements. 28 The United States reported on about ten reactor types completed or in construction under the ABC prototype program at the national laboratories or production plants at Argonne in Illinois, Brookhaven on Long Island, Hanford, Oak Ridge, and the NR TS to assess the various characteristics. A movable "package" reactor for the Army and an aircraft power plant for the Air Force were also being developed. GE and Westinghouse worked closely with some of these facilities and the results of their experimentation would profoundly affect American reactor economics for years.29 Containment building criteria developed in parallel with reactor types. The dangers from the release of fission materials were well recognized and each reactor design seemed to favor one type of construction over another. While remote siting had originally been the main safety feature, the Seawolf test plant near a population center was housed in one of the first vapor-tight steel containment shells. 30

Clearly, Cold War military competition drove American choices for power station reactors. The AEC wanted nuclear power to advance and the naval program offered the fastest and best chance for that to happen. Private industry may have been enticed by financial assistance from the commission's Power Demonstration Reactor Program (PDRP),31 and the exponentially greater power of fission: the energy of one pound offissioned U-235 was equal to the energy in 1000 tons of high quality coal. 32 The government's willingness to provide enriched fuel at nominal prices from its gaseous diffusion plants (built during the war to provide plutonium for bombs) allowed American designers to push the relatively less-efficient light-water reactor that required enriched fuel as the ideal power producer. Based on queries sent by the ABC to private industry in 1951, plans for four private, commercial electric-power generating plants were announced at the 1955 conference. 33 The stations- all with light-water reactors- were at Shippingport, Pennsylvania; Dresden, Illinois; Rowe, Massachusetts (Yankee); and Buchanan, New York (Indian Point).

American and British engineers rightly noted that light-water reactors were going to be limited in their steam pressure and temperature abilities. The parameters of the light-water coolant and moderator were limited by two factors: the difficulty of making large reactor shells pressure proof, and the requirement that the water temperature around the hottest fuel rod areas be kept low enough to prevent film boiling which could cause inadequate cooling. 34 Manufacturing limitations resulted in a maximum working pressure of about 2,000 pounds per square inch (psi) corresponding to a temperature of about 636 degrees Fahrenheit (F). A lower temperature had to be maintained around the fuel rods, resulting in the reactor producing only half as high a heat as that in a contemporary fossil-fueled boiler. It was expected that the poor quality steam would cause moisture problems with steam turbines. In addition, a technology available in fossil-fueled boilers, adding heat to the steam after generation (superheating) was not possible with the water reactors. In some cases plant designers added oil or gas-fired superheaters to raise the steam temperature. 35 The relative inefficiency of the water reactors had another downside that would cause problems later: they produced large amounts of waste heat that had to be dumped into bodies of water unless the utilities opted for large and expensive cooling towers.36 The engineers were concerned about



the safety ofPWRs because the large quantities of pressurized water in the reactor coolant loops increased the potential damage from accidents.37 The gas-cooled system favored by the British had its own share of safety concerns, however. Without plants to enrich uranium cheaply, or the ilumediate need for naval reactors, Britain chose graphite-moderated carbon-dioxide-cooled reactors fueled with natural uranium, and moved faster than the Americans in opening Calder Hall, the world's first commercial nuclear power station, in late 1956 at 65 megawatts (mw).38 France a] so started off with gas-cooled reactors but later went over to PWRs. Canada was an early nuclear power advocate, based on wartime research work at the Chalk River Nuclear Laboratories in Ontario. Assigned to work on natural-fuel, graphite-moderated, heavy-water cooled reactors, the Canadians stayed with that technology in their Nuclear Power Demonstration reactors. This technology led to the successful Candu (Canadian deuterium-uranium) type which uses heavy-water for both the moderator and coolant.39 The USSR inaugurated their power program with graphitemoderated water cooled PWRs.

Steam generators were a critical component ofPWRs. Unlike the steam-generating tubes offossil fueled boilers which were directly impinged by combustion flames and gases, steam generators were fluid heat exchangers. Heated water pumped out of the reactor was forced through steam generator tubes (primary side) without boiling and then returned to cool the reactor. Feed water was continuously pumped around the tubes and was heated to boiling by contact with their surfaces ( secondary side) The generated steam was passed through moisture removal devices and sent to the turbines. Because of the relatively low temperature of the reactor coolant in a PWR, it was necessary to have a large heat-transfer surface to insure reasonable efficiency, calling for almost 4, 000 tubes in the Westinghouse units and even more in those of other manufacturers. The cooling water had to flow evenly through all tubes and the feed water around them to insure full heat transfer, requiring complex perforated tube plates and baffles to channel the flow and insure that the primary and secondary water never merged. The tube bundles had to be supported to resist the flows on both sides with a network of braces and connections. The designs proved vulnerable to damage by various foreign substances.

Shippingport

In its prototype reactor program, the AEC supported in whole or in part the construction of small experimental reactor plants that included gas, polyphenol or sodium cooling, fast breeding, homogeneous fuels, etc. 40 It was no coincidence, however, that the first large nuclear electric power utility station in the United States was a PWR built by Westinghouse and supervised by Rickover, a team with a proven track record. 41 That plant, in Shippingport, was essentially a landbased version ofa projected naval aircraft carrier reactor and went on line in late 1957. A group of manufacturers got together to build this pioneer plant. Westinghouse was the designer and main supervising contractor of the primary (reactor) systems and fabricated the intricate core assembly consisting of almost 100, 000 fuel elements and the critical reactor coolant pumps. 42 Three established fossil-fuel boiler manufacturers supplied other hardware. The reactor vessel was built



by Combustion Engineering Inc. (CE). Foster-Wheeler Corp. (F-W) supplied two straight- tube steam generators, and The Babcock and Wilcox Co.(B&W) provided two u-tube generators, each of which was part of a coolant "loop" out of the reactor. 43 Stone and Webster were the architectengineers with construction shared by Dravo Corporation and Burns and Roe, Inc. Duquesne Light Company supplied the secondary systems (turbine generator, condenser and auxiliaries) and guaranteed to purchase a block of 60 mw power.

The economic efficiency was not expected to compete with conventional plants, even with the ABC supplying most of the development dollars and the enriched fue1. 44 Rickover's standards served as a model for the industry, 45 including 3000-hour core life, redundant fuel rods, backup safety systems, four separate coolant loops to ensure reliability, corrosion resistant materials in the reactor (but not in the secondary systems46), and commercially available equipment. 47 Placing safety in the forefront, he insisted that the reactor be partially buried below grade so that a safety injection system could immerse the core with cooling water without needing pumps.48 If one of the control rods failed, water containing boron (boric acid) to kill the nuclear reactions could be injected into the system. 49 The reactor was contained in its own gas-tight steel chamber, with pairs of steam generators, coolant pumps, and auxiliaries in separate steel chambers, the whole surrounded by more than 5 feet of concrete shielding. The steel shells were designed to resist internal missiles such as valves traveling at high speed. 50 The compact Nautilus reactor was made possible by the use of expensive highly-enriched uranium. The larger reactor for Shippingport had a more economical arrangement with highly-enriched uranium-zircaloy "seed" fuel rods surrounded by natural uranium "blanket" fuel elements of uraniulll dioxide in Zircaloy tubes. 51 The removable head of the reactor \vas penetrated for instrumentation and multiple fail safe control rods to start, maintain, and stop the fission reaction. Fuel rods could be individually replaced through fuel ports with the head in place. The lower section of the reactor vessel containing the core was surrounded by three feet of water

to reflect neutrons. Coolant water entered the bottom of the vessel, flowed up through the fuel rod bundles taking off their heat and out through nozzles above the core to the steam generators. Each reactor coolant pump was a "canned" leak-proofunit developed for the submarine reactors, without seals between the centrifugal pump and motor and cooled by the primary water. Pressure in the system was closely controlled by electric heaters or water sprays in a separate pressurizer vessel. Following the lead of the Experimental Boiling Water Reactor at Argonne, the spent fuel rod bundles were unloaded underwater (to contain the radiation) by remote handling devices and moved out of containment in a flooded canal to a storage pool in an adjacent fuel handling building which al so handled the new fuel. 52 Shippingport was very much a product of national priorities, Rickover's's driving leadership, and perhaps a desire to beat Britain and the USSR.53 The "forced" nature of the engineering and construction, with speed of design a significant factor, impacted a whole generation ofD.S. power reactors.

General Electric's Boiling Water Reactors



While Westinghouse was partnering with the ABC on land based PWR development, GE was working to perfect a boiling water reactor (BWR) to produce electric power from concepts tried out in experimental plants at Argonne and NR TS. 54 In their single cycle BWR, cooling water was allowed to boil in the reactor dome producing steam that was sent directly to the turbine. Starting early with private funding, GE built the Vallecitos boiling water plant in California (ABC license #1) which sent out a small block of power over the Pacific Gas and Electric Co. (PG&E) system in 1958.55 The benefits of reduced pressure in the core (compared to PWRs) and the elimination of what were to become troublesome steam generators were offset by the fact that the power regulation was poor and that irradiated steam traveled out of containment into the turbine, complicating environmental safeguards and turbine maintenance. 56 Despite these problems, the design had development potential, proved to be equally efficient per kilowatt-hour (kw-h), and became the second most common type in the United States.

The "honeymoon" that occurred during the development of Shippingport between Westinghouse, B&W, and CE did not last as their public relations departments touted the benefits of the reactors and components each was putting on the market. With GE, Alco Products Inc., AMF Atomics, The Martin Company, Allis-Chalmers Mfg. Co., North American Aviation (Atomics International), General Nuclear Engineering Corp., and General Atomic in the mix, there probably were too many suppliers promoting too many design variations. 57 By the early 1960s, power companies were likely attracted to the relatively lower costs of light-water designs, and to the more common models for which supplier profits and experience seemed to assure better potential customer support.

Initial Licensing

The ABC licensing program, established under the Atomic Energy Act of 1954, was critical to the siting, design, construction, start-up and power production of commercial plants. The Act gave the government extensive control over research and nuclear materials, but encouraged private industry to build plants. 58 Fuel and rods were leased from the ABC at rates that were designed to make nuclear power stations competitive with fossil-fueled plants. The ABC exercised its control through extensive licensing procedures. Provisional Operating Licenses, Full Term Licenses, amendments, modifications, safety evaluations, and violations or penalties were the ruling documents. The commission specified everything from the facility location to the limits of worker radiation exposure. Changes to equipment that could in any way lead to radiation releases had to be approved. Resident inspectors were constantly in the field checking start up/shutdown procedures, re-fueling and maintenance. The utilities had to respond to the documents with detailed descriptions of actions taken and also send in annual reports. Commission authority over operational issues was accorded by the Code of Federal Regulations (CFR) and license amendments were printed in the Federal Register. 59

Early Commercial Plants



All of the plants built into the early 1960s including Shippingport were heavily supported by the government, but some built under the PDRP were mostly financed and built by private utilities with the AEC providing research, development assistance and free fuel for five years. 60 The first of this series, commenced by Commonwealth Edison Company of Chicago at Dresden, Illinois, was GE's first large (180 mw) BWR plant. The dual-cycle design utilizing "secondary" steam generators to supply the low-pressure turbine had better power regulation than the earlier models. 61

Westinghouse began its first private venture with the Yankee Atomic Power Company plant in Rowe, Massachusetts on the Sherman Pond reservoir of the Deerfield River, the first plant built by a consortium of New England power companies including Connecticut Light & Power Company and other later affiliates of Northeast Utilities. Completed in 1960 as the third American nuclear power plant and the first in New England, the PWR plant with a net output of 167 mw set the trend for subsequent Westinghouse three- and four-loop plants including Connecticut Yankee. Consolidated Edison partnered with B&W to build the 163-mw PWR Indian Point plant on the Hudson River in Buchanan, NY which was designed to use uranium and thorium as fuels. 62 These stations were followed in 1962-3 by the 65- and 67-mw BWR's at Big Rock Point of Consumers Public Power Co. in Michigan and PG&E's Humboldt Bay in California. In 1966 two more stations were completed: Northern States Power's 60-mw Pathfinder BWR plant in Sioux Falls, South Dakota, and Philadelphia Electric Company's 40-mw Peach Bottom High Temperature Gas Cooled Reactor (HTGR) plant. The last plant completed before Connecticut Yankee was the 48-mw BWR La Cross Nuclear Generating Station built in 1967 by the Dairyland Power Company of Wisconsin. The outputs of most of these "demonstration" phase plants were generally less than the fossil-fueled stations already on their respective grids. 63 One other plant usually not included in the history of commercial plants was the N Reactor constructed at Hanford by the government in 1963. While producing weapons grade plutonium it also put a large block of electric power onto the Washington Public Power Supply System grid.64

The Yankee Nuclear Plant at Rowe was the direct precursor to Westinghouse's later plants including Connecticut Yankee. While based on the technologies worked out at Shippingport, the design basis was different enough so that proj ect engineers stated in the company magazine Westinghouse Engineer that no direct comparisons could be made.65 Among the changes at Rowe were:

• use of a single, above-ground steel containment sphere;

• modification of reactor coolant flow with entry and exit nozzles above the core to facilitate the admission of emergency core cooling water;


HAERNo. CI-i85 (Page 12)

• stainless steel cladding on fuel rods instead of Zircaloy;

• uniform slightly-enriched fuel loading instead of the seed and blanket arrangement;

• silver-indium-cadmium control rods (instead of hafnium) with supports extending below the core;

• no ability to load fuel with the reactor head in place.

Boron injection into the coolant aided in normal shutdowns and was also used in the safety injection system (later known as the emergency core cooling system.) Westinghouse-designed vertical Utube steam generators were used in place of the contractor-built horizontal straight- and u-tube types at Shippingport. The elevated position of the reactor required an inclined water filled chute in which a transfer car carried the spent or new fuel rod bundles and control rods to and from the fuel building on ground leve1. 66 Overall, the goals were safety (the location near the Vermont border was considered remote67) and low first cost to make the plant economically viable. 68 Construction was supervised by Stone and WebsteL

Containment Structures

The containment structures of these early commercial stations began to assume the features that were standardized in the 1970s.69 The predominant shapes were either spheres or cylinders with hemispherical tops/bottoms. Dresden had a steel sphere modeled on the earlier Milton shell. 70 Yankee Rowe containment was an above ground spherical steel vapor container surrounding a reinforced-concrete reactor support structure.71 Indian Point had a domed concrete cylinder with a separate internal steel vapor containment sphere. 72 These early containment structures were built under local building codes and American Society of Mechanical Engineers pressure vessel codes.73

While containing radiation was relatively easy, they also had to resist a pressure build up from a release of the stored energy from the coolants and moderators, requiring a pressure rating of around 20 to 30 pounds per square inch gageg (psig).74 Release to the atmosphere was to be restricted but not necessarily prohibited. 75

The AEC's first standardized requirements for siting distance and emissions control were not proposed until 1961.76 As a result of obj ections to aspects of that draft, actual criteria of the 1962 document loosened the definition of "population center distance," and led to a trend of reliance on engineered safeguards for protection rather than remote locations. 77 It suggested that meteorological conditions be considered, and that no reactor be located within a quarter mile of a

gSteam or gas pressure was stated as pounds per square inch gage (psig) which was the pressure over the nominal atmospheric pressure at sea level of 14.7 pounds per square inch (psi). Pressure over true 0 was known as pounds per square inch absolute (psia).



on all environmental factors of seismology, meteorology, geology, and hydrology in the CFR was lilnited to four paragraphs. 78 Planning engineers could use conventional methodology for assessing seismic and wind loading. 79 Assumptions were based on experience with non-mechanical-systemfilled structures such as offices. 80 Thus the earliest commercial plants were often sited in areas that later would probably have been prohibited. The 1959 Santa Susana Station was sited by the Southern California Edison Co. sixteen miles from the San Gabriel fault in an area that was described by project engineers to be " ... as free from seismic disturbances as any in the vicinity of Los Angeles.,,81 The siting of the Indian Point plant near the Ramapo fault was another example. The first detailed criteria from the AEC occurred well after the commercial phase plants were built. 82 It was also some years before effective models were devised to show how critical nuclear components would interact with structures during earthquakes. By 1970 prestressed concrete (previously instituted in French nuclear plants) had taken over from simple reinforced

. 83 constructIon.

Early Insurance Issues

While prevention of release via engineered containment buildings and safety systems was generally accepted by the industry, there was developing resistance to insurance and siting criteria in place during the demonstration phase of reactor construction. As early as the 1860s, private insurers in partnership with boiler makers had arrived at specifications and inspections procedures to protect the public from power boiler explosions. 84 The insurance industry was understandably uncertain about extending fossil-fuel-powered boiler insurance programs to nuclear reactors. 85 To spread the risk they set up insurance pools (syndicates) and instituted rating plans to assess various "nuclear perils,,86 As a result of a 1957 AEC report noting the possibility of up to four billion dollars in costs and thousands offatalities from a major accident, most public liability was transferred from reactor manufacturers and operators to taxpayers though an amendment to the Atomic Energy Act. 87 The substitution of engineered safeguards for remote siting led to challenges to the construction permit for the Fermi Station, located within 35 miles of Detroit and Toledo. A federal court of appeals' decision to rescind the AEC's construction permit ended up in the U.S. Supreme Court which reversed the lower court's decision. 88

Advanced Reactors

While the licensing trend was leaning towards water-cooled reactors, the AEC was still sponsoring attempts at developing systems that could develop higher pressure and superheated steam in commercial plants. The 1963 prototype Carolinas Virginia Nuclear Power Associates (CVNP A) plant in South Carolina which like reactors in Canada used a heavy water moderated, pressure tube design operated for just 4 years. 89 h Experimental superheating reactors were built at Argonne and

h In the Canadian reactors, the fuel rods were contained in individual pressure tubes through which the heavy water coolant flowed, eliminating the need for a reactor vessel containing a large volume of pressurized water surrounding the core. (Mcintyre 1975: 18)



Vallecitos but the technology was not adopted by the industry.9o Sodium cooling with graphite moderation and superheating was tried at the ABC-sponsored, Atomics International (AI) built, experimental Santa Susana plant and their subsequent 77-mw Hallam plant in Nebraska built in late 1963 for the Consumers Public Power District. 91 Hallam was followed by the AI built 61-mw, Detroit Edison Co., Enrico Fermi Liquid Metal Fast Breeder Reactor (LMFBR) plant in Monroe, Michigan which operated for over ten years. 92 There were start-up difficulties in these plants with damage and leaking. While such problems were to be expected with advanced technology, it may have soured the concept with utilities that were experiencing reliability with their light-water thermal plants. Many in the nuclear industry thought that the LMFBR technology was the only economical route in the long term due to limited uranium reserves. 93 In 1972, a major research and development project to create an advanced LMFBR was started by the ABC and two utilities at Clinch River, Tennessee.94 Three competitors, Westinghouse, GE and AI, combined forces for this project. In 1977 the DOE placed a new breeder core in the Shippingport reactor which operated until plant shutdown in 1982.95 The Fermi plant closed in 1972, Congress ended funding for the Clinch River project in 198396, and with ample enriched uranium supplies there was little impetus for further breeder development in the United States.97

The Economics and Efficiency of Early Nuclear Generation

All commercial nuclear plants were designed to meet efficiency objectives which originated over a century earlier The overall efficiency of a steam-powered electric generating station, regardless of fuel source, was determined by a close interaction of all components in the system from the heat source through the prime mover and condenser, with inputs to and from the feed water heating and other auxiliary systems. The goal was to achieve an overall operational efficiency based on an ideal number drawn from the 19th-century theoretical works of Carnot and Rankine. i Beginning in 1922, power station operators used the extraction method of feed heating, in which exhaust steam was withdrawn from the turbines to pre-heat the feed water going back to the boiler to boost overall st ation economy. 98 A few years later, re-heating the steam between separate high- and low-pressure turbine casings also improved efficiency while reducing erosion in turbine blades from moisture.99

With so many stages of heat utilization, the calculations required to achieve maximum efficiency were complex. A new concept called Heat Balancing, (also known as BTU auditing), treated the heat utilization as a balance sheet in which the usage in the components had to balance for maximum efficiency.lOo Tables to plot the heat flows were augmented in the late teens by heat balance diagrams in which all the important components producing and using heat in a plant were

ISadie Camot (1796-1832), a French natural philosopher, founded the science of thermodynamics in 1824. His work "Reflections on the Motive Power of Heat and on the Machines Adapted to Develop This Power" described an "ideal" cycle of steam through an engine and was later developed by Lord Kelvin (1824-1907) and others as the Camot Cycle, the basis of a practical measure of the maximum possible output from a given power system (Wilson 1981:137, Engineering 1907: 847). In England, William 1. M. Rankine (1820-1872) wrote a paper entitled "On the General Law of the Conservation of Energy" in 1853 followed by other writings expanding on thermodynamics. The cycle he described called the Rankine Cycle is used to measure the comparative efficiency of turbine power systems (Engineering 1873: 14, Babcock & Wilcox: 1960: 10-6).


HAER No. CT-185 (Page 15)

drawn in a simplified manner to illustrate the relationship of the components and heat flows. 101 As the technique was perfected engineers of some plants showed all the important parameters of temperature, pressure, and quantity offlow on the diagrams to help designers plan the most efficient arrangements. 102 With these factors included, plant designers and operators could clearly project how changes of temperature, pressure and flow quantity in one component would affect other components upstream and downstream, relative to the total plant output as measured in gross heat rate ofBTUs per kilowatt hour. Designers of commercial nuclear plants prepared diagrams for various heat rates, from initial startup through licensed maximum output. 103

Heat balance considerations were only one set of factors in determining the economic viability of the early commercial stations, whose owners used complex formulas based on assumptions regarding costs of construction, operation, and government-supplied fuel, measured against current and expected costs of coal, the main competitor. Nuclear fuel costs were part of a larger cycle including mining, enrichment, loading, burn up, and recycling or storage of spent fuel (Figure 1). In theory, it was expected that the government would take responsibility for all but the loading and burn up components of this cycle, but the unresolved issues of recycling and storage introduced costs to utilities which were not fully factored into early cost calculations.

While there were several methods of measuring the costs of power stations including cents per BTU, or dollars per kw, the standard measure was the mill (one thousandth ofa dollar) per kilowatt hour. 104 Plant costs were broken down into fixed (structures and equipment), operation, and fuel. Conventional fossil-fuel plant costs during the critical planning period of the first commercial stations were about seven mills total. 105 This was a number that nuclear plants had to approach to be viable. The fixed costs of nuclear plants were much higher due to their still-experimental nature and the requirements for remote siting or containment structures. Shippingport was projected to come in at 64 mills, but Duquesne Light Company was buying the steam at only eight mills reflecting the extent of the government support. 106

Nuclear planners expected that perfected (larger) designs, cheap nuclear fuel costs, a credit for burned fuel, increasing demand, and stable or rising coal costs would change that imbalance. 107 For several reasons their assumptions proved inaccurate. In 1964, the ABC amended the 1954 act to require private ownership of enriched fuel which would continue to be enriched at AEC facilities (known as toll enrichment) until at least 1970. 108 Private recycling facilities were to be set up and the fuel price was stabilized. 109 However, nuclear fuel cost projections failed at the tail end of the nuclear cycle because the expected credits from recycled fuel never materialized due to poor p lanningi .110 [Moved note 108 and deleted "The fuel economy ... "] Starting in 1958 the AEC sponsored the development of rail, truck and barge shipping casks for the spent fuel rods. 111 The failure early onl12 to set up storage locations, and possible public objection to projected routes,

jup to 1971, no commercial reactors had shipped any fuel for recycling (Osbourn and Larson 1971 :247)



resulted in the used fuel being stored in plants' increasingly- crowded spent-fuel pools.k This came to be known as the "stowaway cycle. ,,113 In projecting profitability for the nuclear industry, planners took as a given that demand for electricity in the US doubled every ten years and that coal burning plants would not be able to meet that load. 114 A graph provided by the ABC in a 1959 study of operating costs showed nuclear stations of increasing size (450 mw) dropping to meet a flat line for coal station costs. 115 More conservative projections aimed at establishing a nuclear parity with coal plants stated that to be competitive, the plants would have to reach 1,000 mw. 116 Any belief that coal costs which had been dropping from 1948 to 1958 would rise in the 1960S117 was also in error.

Seeing nuclear power as a second threat after oil to their hegemony, the coal industry came up with new methods and technology that generally kept prices stable and actually lowered costs in some areas. 118 Coal companies encouraged or built power plants close to coal mines that were worked with giant stripping shovels or advanced mining machines. Though these "Mine-Mouth" power plants were sited far from load centers, they were made feasible by extra high voltage (EHV) transmission lines (500 kilovolt amperes) which could send power economically hundreds of miles at five mils. 119 At the same time, railroads concerned about competition from oil-or natural gasfired stations, coal slurry pipelines, and the Mine-MouthlEHV technology came up with the unit/integral coal train concept. 120 The unit train of all coal cars had favorable rates (up to 350/0 cheaper than regular trains121

) as it shuttled between a mine and power station in the load center. The permanently coupled integral train took the concept even farther. The unloading process was streamlined by providing the coal cars with rotary couplers allowing automatic high speed discharge from car dumpers. 122 One utility, Commonwealth Edison, expected to save over four million dollars per year from these innovations. 123 At the same time improved fuel-burning technologies and superhigh-pressure boilers were driving down the number of BTUs required to produce a kw-h of power. 124 Thanks to these advances, and the use of large marine colliers for delivery of coal to plants on navigable waterways, costs of coal-powered generation were not much more than the Jersey Central Power & Light Company Oyster Creek Plant in Toms River, New Jersey, the nuclear industry leader at four mills. 125 As early as 1965 nuclear industry planners were acknowledging that ilnprovements in coal technology had changed the equation but they claimed that it benefitted the nation as a whole. 126 By 1967, the Tennessee Valley Authority (TVA) was projecting its new fossil-fuel and nuclear plants to come in at under three mills. 127 Thus during the critical early phase of commercial nuclear power, there was considerable pressure on profitability.

Another factor that had to be considered in nuclear station economics was the "fit" with the fossilfueled stations on the grid. The fact that nuclear stations had to be of large size to be economic128

posed a problem for the utilities since their intermittent fueling meant that during shut-down for refuel, a large block of power had to be replaced. Still committed to water reactors, Westinghouse

kBetween 1974 and 1980, Connecticut Yankee shipped a total of83 fuel assemblies to General Electric and Battelle, WIth the remaining 1,019 removed assemblies stored in the spent fuel pool (van Noordennen 2005). Battelle is a Columbus, Ohio based, global, non-profit scientific research and management enterprise founded in 1929 which assists the DOE in operating the national laboratories at Brookhaven, Oak Ridge, and Idaho. (Battelle 2003).



sponsored development of a homogeneous breeder reactor system in which nuclear fuel in a slurry loop could be replaced without shut-down. 129 Their Pennsylvania Advanced Reactor (PAR) concept never reached construction. Continuous on-line fueling was also a goal of Combustion Engineering's proposed heavy-water-moderated, organic-cooled SOO-mw plant. l CE was aiming for generating costs of three to five mills in full-scale plants but their technology never came to fruition.130 The only widespread use of continuous fueled reactors in North America are those of the Canadian Candu series. 131 Despite the AEC's encouragement of (and international use of) diverse and more efficient reactor types, the pressure of getting plants on line and making a profit led American utilities to concentrate on the apparently-reliable, somewhat-efficient light-water reactors.

The First Full-Scale Commercial Nuclear Plants and Construction of Connecticut Yankee

By 1963, the stage was set for the first nuclear power stations that could function in multi-station grids on a nearly equal power production and operational cost footing with coal- or oil-powered units. These were important criteria because it was clear that the nuclear stations were not economical unless they were putting large blocks of power into their respective grids. 132 The first "full scale" stations were the three-loop 436-mw San Onofre Nuclear Generating Station of the Southern California Edison Co. and San Diego Gas and Electric Co. in San Clemente, California and the four-loop 616-mw Connecticut Yankee Nuclear Generating Station in Haddam, Connecticut. A 1968 article in Scientific American titled "The Arrival of Nuclear Power" noted the ilnportance of these plants in the maturation of commercial atomic energy.133 Both were Westinghouse-designed plants and both began commercial operation on January 1, 1968.134 They vvere closely followed by the 6S0-mw Oyster Creek station and Niagara Mohawk Power Corp's N-ine Mile Point 610 mw-plant in New York State to round out what has been called the "commercial phase."B5

The long lead times, large size, and problems of engineering containment buildings virtually assured that only a few design and construction concerns in the United States would share the work.

IOrganics are a class of compounds (diphenyls, terphenyls, etc) derived from or containing hydrocarbon radicals. They w.::re first produced by Faraday in 1850 through compression of oil gas (Oxford English Dictionary 1989: v.x, p. 675[Phenyl] and v. XI, p.920 [Organic D. Organics have many benefits as a reactor coolant or moderator: providing a compact core, low system pressure, lack of reactions with fuels or water, compatibility with standard metals, and production of higher temperature steam (Balent 1959: 120).



Ebasco Services, Inc.; Sargent & Lundy, Engineers; Burns and Roe, Inc.; Stone & Webster Engineering Corporation, and Bechtel Power Corporation engineered or built many of the early and commercial phase plants and the latter two would figure in the history of Connecticut Yankee.

Planning and Corporate Organization for Connecticut Yankee

The early success of the Yankee Rowe station, which began commercial operations in 1961, and the increased demand for electricity in Connecticut prompted the state's three largest utilities -Connecticut Light & Power Company (CL&P), Hartford Electric Light Company (HELCO), and United Illuminating Company (UI) - to consider another PWR plant in April 1962. Initially organized as the Nutmeg Electric Companies Atomic Project, this consortium soon concluded that a nuclear station could be competitive against fossil-fuel generation over the life of the plant, using the then-common assumption that coal (and to a lesser extent oil) costs would rise. As discussed above, this assumption later proved false, although in regional terms the construction of more nuclear generating capacity contributed to lower costs in conjunction with pressure on coal prices, the introduction of larger generating units and higher-voltage long-distance transmission facilities, and increased coordination among power companies. Nutmeg Electric moved quickly to option the SOO-acre site of what became Connecticut Yankee in Haddam Neck, and by the end of 1962 selected Westinghouse to produce the major plant components and Stone and Webster to design, engineer, and build the plant. m At the same time, the considerable costs involved, plus the model of the Yankee Rowe consortium and the long history of cooperation among New England power producers,n led to the dissolution of Nutmeg Electric and the creation of a new corporation to build Connecticut Yankee in December 1962. The Connecticut Yankee Atomic Power Company (CY APCO) expanded the consortium beyond Connecticut to include eleven utilities. As plant construction was about to begin, the project became a factor in the negotiations leading to the 1966 creation of Northeast Utilities, an affiliation of CL&P, HELCO, and Western Massachusetts Electric Company (WMECO), the latter also an owner of CY APCO. The NU system, later expanded by the absorption of Holyoke Water Power Company and Public Service Company of New Hampshire, immediately became the largest utility in New England and one of the twenty largest in the nation. 136

m The Boston company was founded by Charles Stone and Edwin Webster in 1889 as an electrical testing lab. The company grew to provide worldwide engineering consulting with a particular emphasis on design and construction of power stations. In 2000, it became a subsidiary of the Shaw Group of Baton Rouge (European Construction Institute 2005: Website, 1; Hoovers 2005: Website, 1).

nThe Connecticut Valley Power Exchange, consisting ofCL&P, HELCO, and Western Massachusetts Electric Company (WMECO), was the nation's first electric power pool when created in 1925 (Northeast Utilities 2005).



Connecticut Yankee Initial Licensing, Construction, and Initial Operation

Provisional and final construction permits for Connecticut Yankee were issued in May and June 1964. Plant siting had to conform to the 1962 ABC Reactor Site Criteria which allowed engineered safeguards to replace remote siting. An additional document, Calculation of Distance Factors for Power or Test Reactors (TID-1844) guided the process which included the maximum allowed releases, containment capability, and environmental conditions at the proposed site to arrive at an exclusion zone. 137 The efficacy of the safeguards including structures, safety injection, water sprays, and filters was balanced against the TID's recommended distance factor. In the case of Connecticut Yankee, the plant's engineered safeguards reduced the exclusion radius from about a mile to 1,700 feet. 138 In addition to the exclusion zone, low population zones and population centers were considered. For instance, the low population zone was one where it could be expected that the residents could be protected from a hypothesized major accident while receiving only a specified radiation limit. 139 Although Connecticut was considered to be a seismically stable area, borings were taken and the plant was designed to be able to shut down safely in a "moderately strong earthquake ... ,,14o

Concrete pouring for the containment and turbine pedestal foundations began in August 1964. The reactor vessel was installed in May 1966 and construction was completed in early 1967. The plant received its provisional license (No. DPR-14) from the ABC in June of that year and initial reactor criticality followed in August141 ABC licenses governed the power level of the reactor which was measured in megawatts thermal (abbreviated mwt).o For startup, the reactor was limited to 1473 mwt out of a possible 1825 mwt. 142 Electricity generation began in August. Under the provisional license the ABC closely monitored start-up activities. During the period before full power operation, adjustments were tnade to equipment, leaks were sealed and turbine stop valves were modified on two occasions. Most of the work was done on the secondary systems outside of containment with some power production continuing. A repair to a steam generator access door a few months after start-up did require an output drop to less than 50 mw. Commercial operation began in 1968. The amendment to the provisional license for full power operation was not granted until February-March 1969 and 600-mwe generation was not achieved until January 1970. 143

OF rom the beginnings of the electric power industry, the power of boilers and turbines was rated in horsepower and the power of generators in watts, kilowatts (a thousand watts), and megawatts (a thousand kilowatts.) From their inception, the output of power reactors was measured in megawatts of heat (Ford 1955: 492.) and later in thermal megawatts. The electrical output in megawatts was lower than the thermal number due to loses in the nuclear skarn supply system and generator. In this historical overview, the capacities of nuclear power stations are given in megawatts of electrical output as described in contemporary and later documents. Published figures for a station can vary because the AECINRC often allowed increases in output over the life ofthe plant. The abbreviation "mwe" came into use in the 1960s and was common in the 1970s. Dates of stations may vary between sources due to the length of time between completion and commercial generation.



Connecticut Yankee Containment and Primary Systems

Connecticut Yankee containment structure and primary systems combined some elements of the Shippingport and Yankee Rowe stations reflecting ten years of development. Shippingport's proximity to a population center and undeveloped standards for mechanical engineered safeguards led the designers to place the steel reactor, steam generator and auxiliary system vessels largely underground. The Yankee Rowe plant had all its reactor systems surrounded by concrete in a completely aboveground steel sphere. 144 Connecticut Yankee reverted to a sub-grade reactor location inside a newer above- and below-ground, industry-standard reinforced-concrete straightwalled cylinder with a hemispherical top known as a "right circular cylinder.,,145 The steel vapor shell was attached to the inside surface of the outer concrete wall. The main components within containment were the reactor; four steam generators and coolant pumps; pressurizer, emergency core cooling system (ECCS), ventilation and filter equipment; refueling systems; and overhead crane. The reactor was enclosed in a separate concrete chamber by a primary shield wall which isolated the coolant pumps and steam generators from radiation, allowing access shortly after shutdown. 146 Additional protective walls separated the pumps and generators into pairs. A second concentric circular concrete wall (secondary shield) further isolated the primary system from the containment shell and provided the support for the "polar" overhead crane that rotated and traversed to cover all the equipment areas. A concrete floor over the reactor and pumps provided a surface for access to the reactor head for refueling. Between the secondary shield and the outer wall of containment were auxiliary systems. The containment building was closely abutted to the spent fuel building and turbine building to ease fuel bundle transfers and keep steam pipe runs short.

The Connecticut Yankee reactor was generally similar to the Rowe reactor but the increased output required a wider and higher vessel, weighing over twice as much and operating at greater pressure. The lower control rod supports used at Rowe were eliminated so the core sat lower in the reactor. lJnlike Rowe, the bottom head of the reactor was penetrated for instrumentation devices. The

nuclear fuel was clad with stainless steel. While the Zircaloy cladding used at Shippingport had superior nuclear properties, it was considered hard to fabricate and not worth the cost at that time. 147 In later years, Zircaloy rods were tested at Rowe, and two complete assemblies ofZircaloy clad rods were included in early Connecticut Yankee cores for testing. 148 As a result of these tests vvhich showed the potential for longer core life, the fuel rods were being completely converted to Zircaloy cladding in the years before shutdown. 149 Control rod materials were the same as at Rowe. The "canned" main coolant pumps used in the first two Westinghouse stations were succeeded by a

new shaft-seal design with almost three times more output. They also incorporated flywheels which insured vital extra seconds of pumping power after a power failure. Since the reactor was nearly at sub-grade a horizontal fuel canal connected the reactor cavity and the spent fuel pool.



Secondary Systems for Electrical Generation and Feed water Control

Steam Turbine Design, Construction, Operations

At Connecticut Yankee, the heat energy to rotative energy converSIon device was a WestinghouselKraftwerke Union (KWU ) three-casing, tandem- compound turbine directconnected to a Westinghouse generator. The nominal turbine output was 619,328 kw with a maximum output of648,527 kw or over 800,000 hp.150 At that load, the turbine was taking 7.463 million pounds of steam per hour. The turbine included one high-pressure and two low-pressure units, which was typical of many large nuclear turbines of the era. The turbines of the Connecticut Y'ankee unit were all on a single shaft with a high-pressure element exhausting into twin lowpressure units. This design was known as a tandem-compound arrangement to distinguish it from cross-compound types which had two or three separate turbine shafts. 151 Turbines were also typed according to the directional flow of steam through the casing. The Connecticut Yankee units were double-axial flow types, in which the steam entered the center of the casing and flowed outward to each end. 152 One advantage of this design was that the thrust on the blades was well balanced which simplified the design of the support bearings. 153 The Connecticut Yankee turbines were also categorized by their exhaust arrangements. The steam exhausted each casing from two ports at each end, called quadruple exhaust. The splitting of the exhaust path allowed a greater flow vvithout greatly increasing the size of the casing ends. This was an additional benefit of the doubleflow design. 154 The steam flow volume dictated the size of the exhaust ports, which in tum dictated the length of the last row of blades in each stage. As discussed below, blade size is a critical factor in turbines because of centrifugal forces acting to pull the blades out by the roots. Casing size and blade length had to be increased as steam pressure dropped and steam volume increased during the flow of steam through the turbine. 155 The constraint of relatively poor steam conditions from the pressurized water reactor generators on exhaust-port design and blade-tip speed required largerdiameter blading in the last stages than were found in fossil-fuel power stations.

Developing steam turbine designs to operate with the first generation of full-scale nuclear reactors of the late 1960's proved be to an engineering challenge for Westinghouse and General Electric. 156 The pressurized water reactors (PWR) favored by Westinghouse, Babcock & Wilcox (B&W) and Combustion Engineering had a design limitation: their use of ordinary water as the reactor coolant severely limited the pressure and temperature of delivered steam. 157 The transfer of heat from the reactor to the steam generators by an indirect heat exchange loop contributed to this problem. 158 Even the boiling water reactors (BWR) of General Electric were limited in their output temperature. 159 The Connecticut Yankee reactor produced steam at 690 psi and 5010 F. Coal and oil fired central stations of the early 1960's generally had boilers operating at over 3000 psig and 1000 0 F. 160 The direct impingement of combustion gases on the water filled generating tubes explained part of their higher operating conditions. In addition, fossil fuel plants utilized superheaters to add extra heat to the steam by running the steam back through the boiler before it went to the turbines. The high temperature steam was very dry which simplified the engineering of



the turbines. The pressurized water reactors of the 1960s could not provide any superheat. In an attempt to achieve higher temperatures, some early plants such as Con Edison's B&W-built Indian Point Plant of 1965 had an oil-fired superheater to improve the steam conditions. 161 The nuclear power industry did not pursue that solution. B& W turned to "Once Through" steam generators in the early 1970's which gave a modest degree of superheat. 162 Westinghouse and Combustion Engineering continued on with their proven U-tube generators producing saturated steam, with a temperature the same as that of the water from which it was liberated. 163 Having made that decision, Westinghouse attempted to design turbogenerators that could effectively utilize huge alTIounts of relatively poor quality steam. These were machines that were as large or larger than existing high-pressure, high-temperature fossil-fuel designs - and had to be because the economics of relatively small nuclear plants were poor. 164

In designing the Connecticut Yankee turbines for relatively low-pressure, low-temperature steam conditions, Westinghouse had to build units working on steam conditions not common in large power stations since the late 1920s,165 by which time a number of reliable designs were available. The Connecticut Yankee use of the three-casing, tandem-compound turbine direct-connected to a generator was a direct descendant of the groundbreaking reaction turbine design patented by Sir Charles Parsons in England in 1884, for which the Westinghouse Electric Corporation of East Pittsburgh was the original American licensee. 166 The three-casing arrangement was an efficient, practical way of handling the huge increase in volume that occurs as steam works its way through the turbine. 167 In Parsons' early machines, a number of increasing diameter blade wheels in a single unit utilized the energy of the steam as it flowed through the blades, losing pressure and gaining in volume. As steam pressures got higher in the twentieth century, builders split the turbine blade stages into high pressure (hp) and low pressure (lp) casings (called compounding) with the steam passing out of the hp turbine via exhaust ports and into the lp turbine. 168 Each casing was larger to accommodate the increase in volume. In addition, each casing was enlarged at the exhaust end to provide free flow for the steam. 169 The stationary and moving turbine blades increased in length along the steam flow path to fill the casings.

By 1920 the first large central station turbines in the 15,000-60,000 kw range built by General Electric and Westinghouse had several yearsil operational experience which included a spate of serious accidents. In some cases blades were completely shed from their mounting discs, in others the discs burst at high speed and wrecked the casings which also cracked from temperature stress. 170 It was clear that engineering had not kept up with the size of the machines. The turbine rotors of that period were of built-up design, including forged spindles with cast steel blade attachment discs bolted or pressed on.171 The bores of the discs were actually machined slightly slnaller than their mating spindle diameter. For assembly, the discs were heated to expand the hole and then forced on the spindle with hydraulic pressure. When the components returned to normal temperature they were locked together producing considerable stress at the mating surfaces. A machined steel key was inserted into a slot cut in both the spindle and disc to prevent rotative separation. Investigators used high speed photography on test rotors which showed that the disc



vvheels were flexing. Metallurgical examinations showed cracks were emanating from keyways, balancing holes and any rough discontinuities in surfaces. Remedies included stiffening the discs, using forgings instead of castings and rounding off corners in the key way areas. It was discovered that cast iron casings "grew" from the higher temperature steam produced by pulverized coal boilers requiring substitution of cast steel. There were also problems with blading clearance, oiling systems and bearings that had to be addressed. At that time the primary problem of turbine builders was blade design. Securing the rotating turbine blades from destruction by vibration required advanced metallurgy and specialized mechanical fastenings. Erosion of the blades from wet steam in the last stages also became a problem. 172

The solutions to these problems emerged from cooperative engineering between boiler manufactures and the turbine makers. Westinghouse started building one-piece forged rotors in the early 1920s which reduced the risk of assembly flaws. 173 Boiler designers increased the superheat so that the steam stayed dry through the turbine cycle reducing the chance of wet steam damaging the elements. They also increased pressures, which helped turbine designers tackle another highrisk area: the blades in the last rows of the low pressure sections. 174 The longer blades in those areas were particularly susceptible to stress cracking at their attachment roots, wearing along their ilnpingement surfaces (erosion) , and centrifugal force working to tear them out of the discs. At the same time it was recognized that steel under stress was particularly vulnerable to corrosive media, a condition first called "season cracking" in the early twentieth century due to its occurrence during wet weather. It was later known as Stress Corrosion Cracking. 175 This phenomenon was observed by jewelers in the 19th century,176 and was seen in brass cartridge cases shipped from Britain to arsenals in India in the early 20th century. It resulted from corrosive media attacking metal parts under mechanical stress produced by applied forces, forming operations, or expansion! contraction. 177

The higher pressures generated by advanced fossil-fuel boilers of the mid-twentieth century mitigated stress corrosion cracking and allowed turbine designers to build smaller machines with high outputs. The smaller machines operating at 3600 rpm had solid forged rotors which were resistant to mechanical failure. It was easier to engineer blading near the smaller exhaust ports and the blades edges were protected from erosion by attached hard metal alloy strips. 178 The high speed engineering also saved money in manufacturing and foundations. 179 By 1950 there were reliable standardized designs putting out 100,000 kw (100mw) at 3600 rpm. The largest turbines of the period still required built-up rotors, and had occasional failures,180 but the excellent steam conditions produced by fossil fuel boilers of that era ensured reasonable reliability.18I Output pressure and temperature were slowly increased through the 1950s. By the time the Connecticut Yankee unit was ordered, coal-or oil-fired boiler/turbine generators were producing 600 mw. 182

Thus the trend in the first hundred years of turbine development was to produce the smallest possible machines producing the highest power at high speed with reliability. This achievement was ai ded by boiler designs that produced huge amounts of steam at very high pressures with high superheat.



A 1966 article on the Connecticut Yankee turbines and their contemporaries in Westinghouse Engineer, the company's house magazine, described the extra lengths that their engineers took to make the huge nuclear power station turbines function in the poor steam conditions.

183 Inlet and

exhaust ports had to be larger than those at contemporary fossil-fuel plants to do the same work. Rotor speed had to be cut back to 1800 rpm to prevent erosion of the very long last row blades resulting from the large ports. 184 General Electric (Westinghouse's chief competitor) used the same rational for its low-speed nuclear turbines. 185 The lack of superheat required steam drying between the stages to protect the vulnerable low-pressure units from moisture. Live steam from the reactor was then used to bring the temperature back up. This process of reheating between stages was used in fossil fuel stations, but the 90-100 degrees of reheat obtained in the Connecticut Yankee plant and contemporary nuclear plants was very low in comparison to levels obtained in fossil fueled st ations. Live steam was even sent direct to the exhaust ends in an attempt to pull out entrained water. Large amounts of steam were used in these areas, but it was not really a problem because the reactor was sized to produce so much more steam than was needed for actually powering the turbines. More water removal occurred at extraction points where steam was bled off to heat the feed water going back into the steam generators.

Turbine Governing

All the calculations for over speed paralneters depended on the turbine governing devices doing their job within specified limits. The function of the turbine governing system was to control the speed of the unit to ensure that the generator was producing even, continuous, high quality electric power. In addition the governing devices prevented over speeding which could lead to explosive destruction of critical components. The Connecticut Yankee governing system evolved from the flyball governors used by millers in the 17th century to control speed in their corn mills. 186 This was an early feedback device: a self-regulating mechanism. 187 James Watt later took that design and patented it for steam engine governing. He used it to open and shut a valve on the steam pipe. 188

Later designers adapted the flyball governor to control the admission of steam by varying the settings of the steam admission valves with mechanical linkages. Westinghouse used the same device on its first turbines. 189 By the nineteen teens, the necessity of precise speed control for electricity generation from large turbines led to the oil pressure relay type govemor. 190 This utilized the pressurized bearing lubricating oil as a working fluid. The Connecticut Yankee governor dispensed with the mechanical complication of a large rotating flyball governor and instead activated the relays with speed sensitive control oil pressure supplied by a pump impeller on the turbine shaft. 191 A key element in the Connecticut Yankee turbine governor was the servo control system developed in the 1920s in which a powerful control motion was produced from a remote and relatively weak signal via sensors, amplifiers, and servo-motors. In

Condensate and Feed water Components



After the steam finished its work in the turbines, it was condensed back to water and recycled. Two surface condensers (nos. 1A and 1B) stood directly below the low-pressure turbines. The condensers were of shell-and-tube construction in which cooling water and exhaust steam were not mixed, a standard design evolved from mid-19th -century steamships which needed fresh water for feeding high pressure boilers. 193 In principal the Connecticut Yankee condensers simply reversed the heat exchange of the steam generators. Water pumped from the Connecticut River flowing through tubes cooled and condensed the surrounding steam. At fun load, 93,000 gallons per minute (gpm) of water was required for condensation. 194 The condensed steam ( condensate) was the main source of feed water for the steam generators. The collection points were the hot wells which constituted the lower two feet of the condenser shell, and normally held 33,000 gallons -- enough for 2.75 minutes of steaming. 195 Two condensate feed pumps on the Turbine Building ground floor removed water from the hot wells and directed it to Reactor Containment. 196

The Connecticut Yankee condensate component design was a "once through" system for condensing the used steam and returning it to the boilers. All the cooling water needed was drawn from the Connecticut River and sent back to the river in a heated condition. 197 This was a common choice in the less environmentally-aware early 1960s. The other, more expensive option would have been a closed system in which the condensing water would be cooled in towers and sent back into the condensers. 198 Westinghouse was an early advocate of marine-type shell-and-tube surface condensers for utility steam turbines like those supplied to Connecticut Yankee. 199 Surface condensers were originally necessary for preserving fresh water boiler feed in steam ships operating in salt water. Most early land turbine installations used simpler condenser types operating on barometric or jet mixing principals. The increasing size of turbines in the twentieth century led to widespread reliance on the ability of surface types to condense large amounts of steam and provide high levels ofvacuum?OO Their heavy water flow required plant siting near rivers, lakes, or oceans. Because they were originally designed to operate in corrosive ocean salt environments, they had

non-ferrous metal tubes to resist wastage. This technology transferred well to power plants in tidal estuaries where salt or brackish water was the rule. Their complex construction with thousands of closely spaced tubes was still vulnerable to corrosion and fouling by biological organisms?OI Tube material had to be carefully chosen to suit the particular local water chemistry. Choice of a closed cycle cooling system would have eliminated biofouling and reduced the chance of corrosion. 202 In the Connecticut Yankee units, the bulk of the original tubing was fabricated of Admiralty Brass. The brass tubes deteriorated due to ammonia induced stress cracking, but operation continued by plugging the affected tubes. This could only be considered a stop-gap repair since output would ultimately drop.203 Failing condenser tubes was a problem in many aging American power plants, and as discussed below eventually led to complete tube replacement at Connecticut Yankee. 204

Studies done to determine that there would be no impact on fish and bird life in the river adjacent to and downstream from the plant were completed after plant design and construction. 205

Generator and Transformer Designs



Connecticut Yankee turbines drove a 1,000,000-hp generator which proved far more reliable than the low pressure turbines and condensers. Although the generator also had to be scaled up for such large output at half speed, improvements in mechanical construction, metallurgy, insulation, and cooling during the previous sixty years of development kept pace with the engineering requirements.

The basic form of the Connecticut Yankee generator evolved from designs developed for European systems of alternating-current high-tension transmission. George Westinghouse recognized the superiority of this system over direct current as early as 1885 and he aggressively purchased patents and licenses from several engineers on the continent and from Tesla in this country in an attempt to lock in the technology. 206 The main benefit of AC high-tension distribution was in economy of copper transmission wire, which was a maj or portion of the capital expense of electrification. 207 A strong influence on emerging technology was the experimental 108-mile 25,000-volt polyphase transmission from Lauffen to Frankfort in Germany in 1891. This installation pioneered high tension, three-phase transmission, with a water powered revolving-field generator and step-up transformers. 208 The use of a three-phase generator gave smoother power, greater capacity and saved money in conductors. 209 In addition the first reliable AC motors worked better on a polyphase system. 210 The main constructional feature was the use of a revolving-field magnet surrounded by stationary armature conductors. This arrangement (which reversed earlier practice in vvhich the armature revolved inside the stationary field magnets), disposed the main elements where they could add to structural simplicity and strength. The copper conductors arrayed in the stationary armature were easier to brace against displacement by electromotive forces. This also eliminated the difficulties of taking high voltages and currents from a moving element. 211 The invention of silicon steel for the conductor-supporting laminations greatly reduced stray currents allowing more output per pound of metal. 212 Placing the relatively simple field magnet wiring on the rotating armature allowed this element to be strongly built to resist centrifugal forces.

Increasingly efficient insulating materials for the copper conductors also played a part in making the early nuclear era generators possible. Varnished paper and cloth used in the first generators gave way to mica in the 1890s.213 In the 1920s asphalt-bonded mica was the norm. Resin-bonded mica and fiberglass came into use around 1950?14 The most important factor in making large generators like the Connecticut Yankee unit possible was hydrogen cooling. In the late 1920s, designers realized that improvements in heat removal in natural circulation air-cooled generators would enable them to get higher outputs from smaller machines. 215 Water cooling the air helped, but the Swiss invention of hydrogen cooling in the 1920's paved the way to greater outputS.216 Hydrogen's lower density and higher heat transfer boosted outputs. At first the pressure was just high enough to keep out air. 217 By increasing the pressure to 60 psi and ducting gas through the conductors, reliable machines of 600 to 1000 mw served the Westinghouse built plants.



Essential to the economics of long-distance transmission was the adoption of alternating currents and step- up transformers. One of the technologies that Westinghouse acquired was the transformer design of Gaulard and Gibbs in Britain. 218 They utilized the principal of electromagnetic induction: current entering the "primary" coil of copper discs at one end of a magnetic circuit produced an electro-magnetic flux which induced a current in an opposite "secondary" coil. Westinghouse engineers rapidly improved this design using copper wire coils and stacked iron plates for the magnetic circuit. 219 By increasing the number of wire turns on the second leg, a relatively low voltage/high current ( amperage) incoming current produced an opposite high voltage/low amperage output. 220 This had two features that aided long distance transmission: it allowed for more powerful generators which did not need hard-to-engineer high-voltage connections, and economized the use of copper transmission wire. The high voltage/low amp output of the new transformers allowed much more electricity to be sent though a given wire size that could be economically strung for hundreds of miles. 221 At the receiving end, the same type of transformers reduced the voltage to a safe level for industry or home use. The main areas of development were similar to those for generators: core, insulation, and cooling. The critical core metallurgy was a challenge early on to designers because with ordinary iron, electrical losses increased with time. 222 The same silicon steel used in armature cores solved that problem.223

Insulation materials evolved along much the same lines as in generators. By the 1930s designers began to design units that could withstand lightning strikes which required a new order of testing, mechanical integrity, and insulation surge resistance. 224 Early transformers tended to be cooled by either forced air or oil, with oil becoming the predominant method for power stations. Natural convection of the heated oil gave way to water cooling of the oil and later to forced oil circulation in external tubed coolers. 225 By mid-century, thermostat-activated fans were added to draw the heat off from, the oil in the cooling banks.226 The Connecticut Yankee output transformer was derived from those "double- and triple-rated" units, as was a step-down transformer which produced a lower voltage to supply the reactor coolant pumps.

Summary of Connecticut Yankee Operations and Repair Issues 1970-1974

The first refueling shutdown began in April 1970 and took about two months, with later refuelings scheduled roughly every year. Each new core required extensive design and engineering of the fuel arrangement, control rods and moderator chemistry to ensure the required power output. Refueling shutdowns also allowed for operational improvements and introduction of other new or re-designed facilities. During the first refueling episode, a new Diesel Generator Building was completed to enhance auxiliary power supply. Enhanced or enlarged facilities to process gaseous and liquid nuclear waste were completed in 1973-74, largely during the fourth refueling shutdown. Until midI 973, plant designers and operators proclaimed satisfaction with what appeared to be trouble-free operations and the production of some 21 billion kilowatt hours, with particular satisfaction expressed about turbine performance. 227 At about the same time, a GE engineer stated that the erosion rate of their nuclear turbines was no worse than their fossil fuel units,228 and Nuclear Safety, the bimonthly review of the Atomic Energy Commission, indicated no turbine erosion or



corrosion problems after reviewing the performance of twenty-eight light water reactors.229

Soon after these articles appeared, however, significant design problems appeared in the steam generators and low-pressure turbines.

Turbine Repairs 1973-1974 - Emerging Issues of Stress Corrosion Cracking and Erosion

The low-pressure turbines at Connecticut Yankee began to fail in the Spring of 1973.230 The first repair on the No.2 unit was in June 1973, and the spindles on both low-pressure units were replaced between July-December of that year, possibly using spares provided by Westinghouse. The rotors must have been severely degraded since the repair was not done during refueling. 231 The station was out of service for over five months. Another month-long repair requiring shut-down began in 1974. The problems included disc cracking, blade root cracking, and erosion of the stationary and rotating blading. 232 Indirect evidence from other plants suggests these repairs reflected stress corrosion problems which began soon after Connecticut Yankee began full-power operations in 1969, if not earlier and a re-emergence of the blade erosion problems that had occurred in the early 20th century.

Beginning in 1965, while Connecticut Yankee was under construction, the steam generators and turbines in one of the pioneer nuclear power stations of the Central Electricity Generating Board (C.E.G.B.) in England began to show damage from feed water impurities. 233 Just before the lowpressure turbine problems at Connecticut Yankee became evident, a groundbreaking report on a turbine failure in another C.E.G.B. station reached the engineering journals. In 1969 the Hinkley Point 'A' Nuclear Station had a catastrophic failure of the low-pressure blade discs on one of its turbines. The unit was very similar in layout and construction to the Connecticut Yankee unit, though smaller and running at 3000 rpm. A 21/2-year investigation revealed that the discs failed due to stress corrosion cracking. The report found that the causes of the corrosion were minute ilnpurities in the steam, attacking very tiny defects in the disc attachment points. 234 By this time, stress corrosion cracking was well understood at the molecular level, although it was many years before the engineering caught up with the science?35 Failure from stress corrosion cracking takes about four years to develop - about the length of time between full-power operation and first

b' . C . Y k 236 tur Ine repaIrs at onnectlcut an ee.

The problems afflicting the Connecticut Yankee low-pressure turbines were common in the first generation of Westinghouse units. The near-sister plant to Connecticut Yankee at San Onofre, California had problems with cracking in the keyways that locked in the blade discs to the spindles. The carbon steel of the rotors could not handle the relatively high moisture content. 237 Brookwood #1 (now Ginna) of Rochester Power and Light suffered blade failures with blade ejection requiring

operation with the last row blades removed while engineers tried to find solutions. 238 Ten years after delivery of the Connecticut Yankee turbines, some engineers still felt that conditions in lowpressure nuclear units were not very different from fossil-fired units and did not require new engineering?39 Westinghouse's (and GE's) assumption that lower rotor speeds would reduce the



erosion of the last row blades may have been in error as later observations found that high revolutions led to longer blade life. 240 Until c 1973, GE turbines operating at PWR and BWR plants

. f . k' 241 had not shown sIgns 0 stress corrosIon crac lng.

While the incidents undoubtedly caused economic harm to the utilities, public notice was probably muted by the lack of safety issues, the subsequent Three Mile Island accident, and later ongoing problems with the more newsworthy steam generator tubing. It is clear that the Connecticut Yankee low-pressure turbine problems were caused by unforeseen engineering decisions and manufacturing methods used by Westinghouse. Excluding the C.E. G.B stations, these types of turbine failures were very much a United States problem that also affected non-nuclear turbines with increasing frequency from 1964 to 1973.242 For various reasons, similar turbines, even American-made ones, did not fail in Germany or Japan?43 A critical component in stress corrosion control is the chemical quality of the feed water going into the system. Very slight rises in salinity or minerals could exacerbate the deterioration of components in the steam path. 244 The Hinkley Point investigators even found that stress corrosion cracking could also be induced by certain water purity control chemicals. 245 However, in reviewing all the C.E. G.B. stations that had cracking, they found that though water quality varied it was still within operating specifications. In addition they felt that it would have been impossible to expect the controls to be any better. Their recommendations were that the details of the highly vulnerable disc keyways had to be better engineered. 246 The cooling water intakes for Connecticut Yankee were in theory sited upstream of observed salinity, and the operating engineers kept fairly close controls to prevent its entry into the system. 247 Water quality issues were also central to problems with steam generators at Connecticut Y-ankee, and at many other nuclear plants.

Connecticut Yankee and Worldwide Steam Generator Problems

Shortly after full power operation began in 1970, leaks in the steam generator tubes were detected. 248 Tube leaks allowed irradiated primary coolant water past the Reactor Coolant System (RCS) boundary, creating a safety hazard. Consequently, federal regulations set maximum gallons per day leakage rate for all steam generators.249 Initially no repairs were needed. Between June of 1973 and April of 1974, during repairs on the low-pressure turbines, steam generator tube leaks were stopped by explosive plug welding. 250 This technique was developed around 1970 in answer to the difficulties of closing tubes in an irradiated area with manual plugging or welding. On detonation of a small nitroglycerine-based charge, molten metal cleaned the inside of the tube, and weld positioned a wooden plug to block loss of coolant. 251 While plugging was effective in preventing coolant loss, it could only be considered a stop-gap as plant output would drop if too many tubes were plugged.

The history and causes of steam generator problems were related to the issues noted for turbine blades and rotors, and were found in nuclear plants built before Connecticut Yankee. In early 1958, after only a few months of operation, one of the B& W horizontal u-tube steam generators at the



Shippingport station developed a tube leak. Testing with an eddy-current device!' showed extensive surface stress corrosion cracking. An analysis by Westinghouse and Duquesne concluded that the chemicals then being used in Duquesne's fossil-fueled boilers to control oxygen levels (sulfite) and the pH of the feed water (phosphate) were not as effective in the smaller nuclear (in comparison to fossil-fuel) steam generators. 252 Changes in the type of phosphate, amounts of sulfite and modifications to the tube arrangements controlled the problems.

The British investigations of feed water contaminants in the late 1960s, noted above, found that st eam generator tubes were scarred by corrosion.253 An intensive investigation by the British board indicated several causes including steam generator tube configurations, breakdown of the feed water heater tubing and carryover of the particles, insufficient air removal from condensed steam, and failure of resin filter beds to remove organic compounds from lake-sourced feed water. Improving the filter media, eliminating air from the feed water and adding phosphate water treatment largely resolved the problems. 254 Six years after Yankee Rowe began operation in 1960, leaks were detected in its steam generators with stainless steel tubes, but operators controlled the problems. 255 A few years later, utilities with Westinghouse, B& W, and CE units tubed with Inconel 600 (nickel-iron-chrome alloy256) tubes began to have an epidemic of tube and tube support failures that eventually affected forty stations. Some plants had steam generators with 20% of their tubes plugged to prevent leakage. 257 The types of degradation included wastage, pitting, denting, cracking of the tubes, support plate damage, and mechanical damage from vibration and loose parts.

In analyzing steam generator problems, a 1988 NRC report found a complex interaction between the mechanical design, materials, fabrication methods, water treatment chemistry and corrosion products from the plant secondary systems, mainly the condensers.258 The construction of the steam generators, with thousands of tubes, tube seal plates, and bracing, provided many areas for corrosion materials to accumulate and do damage. Designers of the first PWRs saw the water in the primary reactor/steam generator loop as a potential source of reactor or steam-generator tube damage due to the addition of acidic boron to help control the nuclear reactions. While they specified precise chemical parameters for that system, less attention was given to secondary systetTI feed water contro1. 259 In both fossil- and nuclear-fueled stations, the secondary system condensers that returned the steam leaving the turbines to water were designed to be the "first line of defense" against the entry of corrosion products, yet they also could be their main source.260 The thousands of tubes providing the condensing surface area had to be sealed into tube plates at both ends and were acknowledged by American operators to be practically impossible to keep leak-free?61 The

PEddy-current testing of boiler tubes is a remote, non-destructive procedure that utilizes electro-magnetic fields from a probe to find faults by a change in signal intensity caused by variances in wall thickness and cracking. (Singley et al1959: 753)



fact that the condenser shell was operating in a vacuum meant impurities and air in the cooling water drawn from oceans, rivers, or lakes would be sucked out of any leaking tubes to mix with the condensing steam. As a result, air, chlorides, hydroxides and up to ninety other chemical compounds could enter the system and induce stress corrosion cracking in the steam generator tubes and also the turbine blade roots and mounting discs. 262

"Deaerating" sections of the condensers, air ejectors and chemical water treatment were necessary to mitigate these problems. Recognizing that condensers and mechanical devices alone could not bring oxygen levels low enough to prevent corrosion, engineers offossil-fueled stations had started using extensive chemical feed water treatment around 1950. Most stations used phosphate to keep water pH below corrosion thresholds. 263 Phosphate treatment did not alleviate the dissolved oxygen content of the water, so some utilities added sulfites or hydrazine as oxygen scavengers in the mid 1950s.264 Later ammonia or morpholine-q were added to control the pH of the water. 265 Initially positive results were mitigated by increasing evidence that the hydrazine or ammonia attacked the copper tubes in feed heaters and condensers, leading to suggestions that those units be entirely tubed with carbon or stainless steel. 266

American nuclear plants through the generation including Connecticut Yankee used phosphate treatment until it became evident that the phosphate built up as sludge on tube sheets causing wastage and thinning of the steam generator tubes. 267 For better control of oxygen content and to forestall tube deterioration, both Connecticut Yankee and San Onofre tried hydrazine injection as early as 1970.268 With the blessing of the manufacturers, American nuclear plants switched to an alnmonia- and hydrazine-based "all volatile treatment" (AVT) around 1974?69 At first AVT reduced tube plugging, but it then began to cause other problems. The ammonia injected into the feed water (or that produced by the breakdown of the hydrazine), in combination with the dissolved oxygen, attacked the copper-based tubing that had been specified for the demonstration- and commercial-phase feed water heaters (feed train) and condensers.27o The copper shed from the tubes ended up in the steam generators where it acted as both an oxidizer and catalyst for pitting of the tubes. 271 At the same time, the feed train materials and the chlorides from condenser leaks would lodge in the spaces between the steam generator tubes and their drilled, carbon-steel support plates. A resulting buildup of oxide would then squeeze the tubes (denting) leading to cracking and nlptures. 272 Steel and copper corrosion products would also accumulate on top of the lower tube sheet as a sludge that would lead to stress corrosion cracking (also known as inter-granular attack) damage of tubes. 273 The stress set up when the tubes were bent to form an inverted V-shape made that area particularly vulnerable to stress corrosion cracking. Even rigid control of pH did not

q Morpholine is a common additive, in ppm concentrations, for pH adjustment in fossil fuel and nuclear power plant steam systems systems. Morpholine's volatility is about the same as water, so once it is added to the water, its concentration becomes distributed evenly in both the water and steam. Its pH adjusting qualities then become distributed throughout the steam plant to provide corrosion protection.



guarantee steam generator health because the right level needed to control copper corrosion would allow iron oxidation. 274 The chemistry problems were exacerbated by operations of the plants at low loads and during startup. The mechanical damage was caused either by loose or foreign parts (some left in the units by assembly or repair teams) impacting the tubes or by vibration due to inadequate supports. Clearance between the tubes and their support plates and bars was necessary in some models of generators but it allowed the large volumes of water and gases traveling through the generators at high speeds to cause relative movement (fretting) leading to tube damage. 275

These problems perplexed manufacturers and plant operators. From 1968 to 1975, San Onofre modified its phosphate chemistry four times, switching between A VT and phosphate before settling on A VT at the request of Westinghouse or because of its own investigations. 276 Connecticut Yankee documents suggest that the utility also tried different chemistries, and that these choices could have been a causative factor in its steam generator tube problems and its heater tube, condenser tube and turbine replacement projects (see HAER No. CT-185-C-Turbine Building).277 The power companies were evidently getting insufficient help with these problems from their suppliers, and in 1977 operators formed the Steam Generators Owners Group (SGOG) in conjunction with the Electric Power Research Institute (EPRI) to address the problems.278 SGOG research showed that the chemistry parameters set by the manufacturers were too loose. As an example, the original specification on chlorides allowed 150 parts per billion (ppb) while the EPRI/SGOG guideline limited it to just 20 ppb. 279 While leak problems were usually manageable, more serious issues began in 1975 with a steam generator tube rupture at the Point Beach # 1 Plant of the Wisconsin Michigan Power Co., followed by ruptures in other plants. Tube ruptures allowed much more primary coolant to escape, and could lead to other system failures resulting in serious accidents. 28o In 1978, the NRC designated steam generator tube integrity as an unresolved safety issue.281 The solutions to these problems took years to develop. Tube plugging was augmented \vith sleeving in which a section of smaller diameter tube was inserted into the leaking one and mechanically sealed allowing coolant flow. 282 Sleeving did not occur at Connecticut Yankee until after 1987.283 The AEC/NRC instituted inspection programs in which the entire length of every tube had to be examined by eddy-current testing devices. The costs for inspection could reach $500,000 per day (including replacement power and provision for worker rem exposure) in the Westinghouse stress corrosion cracking steam generators - which had fewer tubes than those of the other makers. The problems described above were not limited to American PWR plants. In the 1970s, French stations using Westinghouse-licensed units experienced tube and tube plate corrosion. Their engineers suggested that the Inconel 600 material had to be improved with heat treatment, that the tube support plates be made of stainless steel, or that the tube/plate interfaces had to be upgraded. They felt it imperative that there be no condenser tube leakage. 284 After some problems in German stations, the plants controlled the problems with improved tube metallurgy

r Roentgen Equivalent Man=the quantity of radiation having the same effect on human tissue as one roentgen of X-rays. (Oxford English Dictionary 1989: vA, p. 576.)



(lncoloy 800), largely leak-free and corrosion-resistant condensers, and case-by-case use of either low-phosphate or all-volatile treatment. 285 In addition, European operators were more willing than American utilities to shut down their plants when chemistry upsets or condenser leaks were detected. 286

F or generator problems caused by secondary system impurities, manufacturers recommended installation of "full flow" condensate "polishing" systems which used resin-filled filter beds to purifY continuously all the condensate after it left condenser hot wells. 287 This would have provided much greater control than the more common intermittent treatment, but some engineers opposed it because of high capital and operating costs, and because of concerns that the resins used in the system would cause their own problems. Connecticut Yankee did not add a condensate polishing system, probably for these reasons?88 Another solution - which Connecticut Yankee instituted around 1977 - was to "blow down" the steam generators frequently to clear out deposits. S It is undocumented whether this change in operation required additional plant infrastructure or NRC approval. By the 1980s, nuclear engineers realized that slightly brackish cooling water supplies demanded more advanced metallurgy to prevent bio fouling and stress corrosion cracking.289 To reduce those conditions at Connecticut Yankee, all the tubes were replaced in 1986 with a proprietary stainless alloy - Trent Sea-Cure, which came on the market in 1979 - in an attempt to prevent damage in those sections, contaminant particle carryover, and subsequent denting in the steam generator tubes. 290

While most of the PWR steam generator problems were caused by the secondary system water, the chemistry of the primary system could also cause damage. Cases of primary-water-induced cracking of steam generator tubes at the stressed U-bends began to appear. 291 Damage to the reactor could also occur and was a factor in the shutdown of Yankee Rowe station. 292 A more serious development was the discovery in France that the boric acid moderator added to the coolant could attack the reactor head. The Toledo Edison-Cleveland Illuminating Co. Davis-Besse plant in Ohio was shut down in 2002 as a result of severe corrosion around the control rod penetrations?93

S Blowing down (also known as blowing off) a boiler was a water purification technique from the earliest days of steam technology (Rankine 1859: 453). Allowing some of the pressurized water inventory to escape from the boiler removed oil, salt and other contaminants which could settle on heating surfaces and restrict heat transfer leading to premature failure. Bottom blow-off and surface blow-off valves cleared out the two regions where substances generally accumulated. While blowing offwasted heat in conventional boilers, in nuclear boilers it also allowed irradiated water outside ofthe reactor coolant system boundary requiring storage tanks and filters.



During the era in which PWR owners and manufacturers were working on steam generator problems, operators ofGE's BWRs were having their own set of troubles. While the successors to the dual-cycle plants were not saddled with PWR steam generators, they began to have a form of stress corrosion cracking in the recirculation and other stainless-steel reactor piping. Following the PWR owners' lead, the utilities and EPRI formed the Boiling Water Reactor Owners Group which successfully addressed the problems. 294

Summary of Connecticut Yankee History 1974-1984 and Reactor Cavity Seal Failure

The ABC, evidently satisfied with the measures taken by Connecticut Yankee to repair the failed low- pressure turbine rotors and steam generator tubes while operating under the provisional license, authorized full-power operation at 1825 mwt on December 27, 1974 with Facility Operating License #DPR_61. 295 In 1975 the AEC was replaced by the Nuclear Regulatory Commission and two years later the Department of Energy (DOE) was created. Reflecting the ongoing national failure to accommodate spent fuel, the NRC amended Connecticut Yankee's license to allow an increase in the spent fuel pool capacity from 336 to 1172 assemblies in 1976?96

-Until the mid-1980s, most major changes in plant facilities or operations were driven by national issues in nuclear plant safety. A 1975 fire in the TVA's unfinished Browns Ferry Unit 3 in Alabama led the NRC to require upgrades offire protection in all American plants. At Connecticut Yankee, resulting improvements included barriers, detection equipment, and fire-fighting capabilities. Work and materials storage methods were changed, with great emphasis on controlling combustibles and ignition sources. On March 28, 1979, the most serious accident in the history of American commercial nuclear power plant operations occurred at the Jersey Central Power & Light Co. Three Mile Island (TMI) unit 2 facility near Middletown, P A. There were no radiation-exposure consequences, but the reactor overheated and fuel melted. Causes included personnel error, design deficiencies and component failures. As a result of this accident, significant changes were made in the industry. The NRC issued amendment No. 42 "TMI Lessons Learned Category "A" Items for Connecticut Yankee. 297 Changes to the plant from this amendment included new accident monitoring systems, new control room instrumentation, seismic improvements to the Service Building housing the control room, and the construction of an Emergency Operations Facility Building in 1980.298 The largest fire-protection modification at the plant, a new electric switchgear building proposed in 1986 and completed in 1990, was also an outgrowth of increased accident protection measures. 299

In 1982-83, Connecticut Yankee modified the reactor cavity seal ring, a vital component of the refueling system, prior to the 1983 refueling.30o During the 1984 refueling that seal ring failed, leading to the most serious accident in the plant's history.



Refueling and the Reactor Cavity Seal Design, Failure, and Reconstruction

The immersed refueling system relied on the principal of water seeking its own level between connected containers, and included the reactor cavity and its adjacent refueling (transfer) canal in containment, the fuel pool in the spent fuel building, and a transfer tube connecting them. Before refueling, water treated with boron to kill any nuclear reactions was pumped from the Refueling Water Storage Tank into the reactor cavity (surrounding the top of the vessel) to an elevation of 46.5 feet above mean sea level (equal to the elevation of the spent fuel pool) to ensure complete coverage of fuel bundles during the refueling process. The head of the reactor was lifted offby the polar crane as the water level rose, and set down in a circular 47-foot-deep concrete pit within containment. With the water levels equalized, the valves and sluice gate that sealed off the transfer tube were opened, providing a continuous water path to convey spent and new fuel rods between the two structures. The refueling water height 24.5 feet above the open top of the reactor vessel provided enough clearance to fully protect the fuel rod bundles as they were pulled out with a manipulator crane on the refueling floor above the cavity. The crane operator then placed the bundles vertically in an upender machine in the transfer canal next to and below the mouth of the reactor. The upender set them in a horizontal position on a wheeled car that carried them through the transfer tube to the fuel pool. 301 Another upender and crane handled the bundles in the pool. When it was time to bring in new bundles from the spent fuel building the process was reversed.

It was necessary to prevent the water in the reactor cavity from pouring down between the shell of the reactor and the surrounding concrete wall, past the neutron shield tank and then into the floor of the containment building. If that occurred when the canal was open during a transfer, in a worstcase-scenario, the water level in both buildings could drop, possibly enough to expose the entire length of bundles being carried by the cranes or a portion of the bundles in the up enders, and the stored fuel bundles in the poo1. 302 The subsequent heating of the rods would have produced high doses of radiation to personnel, fuel cladding failure, and possible release of radiation to the atmosphere. 303 The original reactor cavity seal was a circular steel plate bolted in place between a flange around the top of the reactor vessel and the adjacent concrete, covering the annulus (opening) between the two. During refuelings prior to 1983, the seal had small leaks which led to contamination of the lower portion of the vesse1. 304 Connecticut Yankee engineers proposed anew seal device which consisted of a plate surrounding the opening with continuous inflatable rubber boots on the inside and outside diameters. On inflation the boots would pull down T -shaped wedges of rubber to plug the openings between the flange of the reactor and the inner edge of the plate, and between the outside edge of the plate and the surrounding structure. The modification was made and it followed the recommended Plant Design Change Record (PDCR) procedures as outlined in the CFR. 305

On the morning of August 21, 1984, after the cavity had been filled and the head of the reactor removed prior to refueling, the seal failed. In less than half an hour, all 200, 000 gallons of water in the cavity drained down through the seal. The water elevation after the accident was at 22 feet,


HAERNo. CT-i85 (Page 36)

level with the open top of the vessel. Because all the fuel rods were still under water in the reactor, they were not damaged. The transfer tube had not been opened at the time of the accident so there was no loss of pool water and exposure of stored fuel bundles. Operators initiated the correct actions to begin pumping out water from the floor of containment. There was a small filtered release from the ventilation stack. Connecticut Yankee personnel followed procedures to notify the NRC, the state, and declared an "unusual event" in compliance with the Emergency Plan.

306

Refueling was terminated and the event was declared over when the water in the lower portion of containment was pumped out. On dewatering it was found that the corrosive borated water had penetrated insulation on the bottom of the reactor and piping, requiring removal and repairs.

An investigation by Connecticut Yankee engineers found that the seal had been incorrectly designed and tested, allowing a critical part to deform after inflation under the full "head" of water leading to gross failure. 307 The previous design, though not completely watertight, was more failure-proof Northeast Utilities (NU) notified other licensee's - twenty-seven reactors had a similar seal -through the Institute of Nuclear Power Operations (INPO) network. 308 INPO was created in 1979 to share information between the utilities and the DOE. NU also evaluated the possible impact on upcoming refueling operations at its three Millstone reactors in Connecticut. 309 As a result of the inquiry, hidden flaws in the refuel system design were revealed prompting the NRC to issue a bulletin to almost every operating or planned reactor in the U.S. about the danger of this type of accident. 310 Several corrective measures were taken by Connecticut Yankee to prevent a recurrence. The seal was redesigned with steel rods to prevent the top portion from deforming and a backup seal was added above the main seal. A fixed wall (cofferdam) was added in front of the canal so that even if the seal failed with the transfer tube open there would still be enough water to cover the stored bundles in the spent fuel pool. The additional height of water would also give operators time to activate pool cooling mechanisms. Operators of the manipulator cranes and upenders were trained to quickly place bundles in transit in a safe position during unplanned cavity drainage. 311 The sluice gate in the transfer tube was redesigned to close against a flow of water pouring out of the pool, an event that was not contemplated in the original design. 312 While there would still be water in the reactor after a failure, it was required that the Residual Heat Removal pump would always be activated to provide additional circulation to prevent heating of the rods still in the core. 313 At some point before 1985 the revised temporary cavity seal ring was replaced with a permanent stainless steel ring. It further reduced the chance of failure, allowed for reactor movement, saved refueling time, and eliminated worker radiation exposure. 314

On December 12, 1984, the NRC issued a "Notice of Violation and Proposed Imposition of Civil Penalty and Order Modifying License" which instituted an $80,000 fine to the Connecticut Yankee Atomic Power Company. Management elected not to contest the fine but also cited a number of compliance actions on their part which they felt should have reduced the fine. These included the prompt notification to NRC and other utilities, in-depth investigation of the event and other potential causes, an extensive redesign process and co-hosting an INPO workshop on seal failure. 315



Prior to the accident, NRC inspections revealed two other earlier modifications to the plant that the commission felt were not properly instituted. The changes involved radiation monitors and a control valve in the Post Accident Sampling System. The NRC held an enforcement conference in November 1983 to determine if there was a pattern of inadequate design modification processes.

316

The NRC did not find that to be the case, but Connecticut Yankee instituted an improved PDCR process and sent out a letter to all personnel in Nuclear Engineering and Operations asking for indepth questioning about all possible circumstances (described as "what ifs?") of future design

] 317

C langes.

Summary of Connecticut Yankee Operations 1986-1996

During 1986, the NRC issued an amendment for Connecticut Yankee regarding specifications for three-loop operation. 318 The four-loop design of Westinghouse reactors allowed one loop (coolant pump, steam generator and associated piping) to be shut down for repair while the station operated at reduced output. Connecticut Yankee rarely operated in that fashion. Construction on the new switchgear building, completed in 1990, was begun to meet updated fire protection criteria, and provide enhanced instrumentation and controls for safe plant shutdown.

By the 1980s, nuclear engineers realized that slightly brackish cooling water supplies demanded more advanced metallurgy to prevent bio fouling and stress corrosion cracking. 319 To reduce those conditions at Connecticut Yankee, all the tubes were replaced in 1986 with a proprietary stainless alloy, Trent Sea-Cure, which came on the market in 1979.320

During the fourteenth refueling outage in 1987, additional steam generator tubes were plugged and the low pressure turbines were replaced (see HAER No. CT-185-0-Turbine Building). Containment leak integrity was tested by pressurization. Repairs were made to the attachment devices on the thermal shield surrounding the lower core barrel, probably to reduce flow vibration of the shield. 321 Two NRC resident inspectors put in over 4000 hours during the assessment period before and after the shutdown. 322

Early in 1989, there was a release of radioactive liquid from the Spent Fuel Building into drainage structures at the nearby 115 kilovolt switchyard, which delivered from other power stations in the system almost all the station service power for start-up and shutdown and power production operation. Clean-up after this event required considerable soil removal. During the fifteenth refueling outage in 1989-90, the thermal shield around the lower core barrel was removed, and the entire core was transferred to the spent fuel pool. 323 Generally only one third of the rods were replaced in each refueling operation, so this may have been an unusual incident, the reasons for which are as yet undocumented but may relate to a fuel reconstitution project. 324



In May 1990 specimens of Asiatic clams were found in the service water system of the plant. The species (Cohicu/ajluminea) spread rapidly in North American fresh waters. Since fouling by these bivalves could comprise important safety systems, Connecticut Yankee was allowed by the Connecticut DEP to continuously chlorinate the system.325 From 1989 though 1994 there were no alnendments on tube plugging so it must be assumed that either plant chemistry was under control or leak rates were under limits specified in the CFR. Tube plugging was resumed during the 18th refueling in 1995 along with roll expansion repairs of tubes. This was an older, more labor intensive process in which plugs were rolled into both ends of the tubes. 326 During 1996, the last year of generation, additional trip mechanisms were added to control containment high pressure and steam generator blowdowns. The Union of Concerned Scientists (a nuclear watchdog group) claimed that the NRC had found that a critical coolant pipe was undersized and had gone undiscovered for 30 years. 327 In October 1996 a gas bubble formed in the Connecticut Yankee reactor, and unnoticed by operators, had flushed out cooling water. 328 It is undocumented whether the 19th refueling cycle was completed in advance of plant shutdown in December 1996 .

Shutdown of Connecticut Yankee

The decision not to seek license renewal and commence decommissioning was based on a study which showed that due to changing market conditions, Connecticut Yankee's customers would save money if the plant was shut down. 329 A review of the physical state of the plant as shown in CY / AECINRC documents combined with the economics of light water reactors from the "commercial" generation of plants can give a picture of what might have caused the Connecticut Yankee directors to decline to ask the NRC for an extension past 2007. License extension was an option that the DOEINRC encouraged since it gave more time to amortize the costs of upgrading the older plants.33o The ongoing steam generator problems might have been a factor in the decision. In its 1988 report on steam generator failures, the NRC worked out "value-impact" models to show what utilities could expect to spend during the remaining life of their plants under the stepped-up inspection plans being implemented because of the unresolved steam generator safety issue. The modeling estimated inspection times, plugging man-hours, occupational radiological exposure (ORE), and replacement power to give a picture of downstream costs. Included in the NRC report was the possibility of partially or completely replacing a steam generator, some of which had been operating for only 10 years. From 1981 to 1993, nine Westinghouse plants and two CE plants had Steam Generator Replacement Projects (SGRP's) 331

The first of these replacements, at Virginia Electric Power's Surry #2, a three-loop Westinghouse plant cost over 200 million dollars and involved 2141 man-rems of radiological exposure. NU spent ten years planning the replacement of two CE generators in its Millstone Unit #2 which had over 3,000 plugged or sleeved tubes. 332 In that repair operation, only the bottom portion of the generator containing the tubes was replaced. Improved methods including use of robotics and a full size mock-up building resulted in greatly reduced costs and ORE. To facilitate future replacement



projects, the NRC in 1989 allowed utilities to undertake them without prior NRC review or approval. 333

With the completion of the Millstone SGRP in 1992, NU certainly had the skill sets to assist Connecticut Yankee in a steam generator replacement program. However, it may have been difficult to find lower tube sections that were compatible with the early model Westinghouse generators. If Connecticut Yankee had elected to replace all or part of the generators it would have had to protect them from further corrosive attack. This could have included installing a condensate polishing system, additional air removal devices in the feed water supply and condensate system and further upgrades of the condensers. 334 The shutdown and decision to decommission the NUatIiliated Yankee Rowe station in 1992 and Millstone # 1 in 1996 provided the company with experience in an alternative to upgrading. 335 Another factor in Connecticut Yankee's decisionmaking process may have been the ongoing problem of keeping an older plant up to the contemporary NRC safety codes. Years of "back fitting" before and after TMI had left older plants over-complicated and crowded. 336 In the final analysis though, national economics alone could have influenced the decision. While the nuclear fuel component of Operating and Maintenance (O&M) costs in 1993 was generally lower than fossil fuel costs, overall O&M for nuclear plants had risen higher than their competition by 1987.337 Only operating efficiencies were going to improve that ratio, and they were going to be hard to come by at a 28-year-old plant that was nearing the end of its operating license.

Connecticut Yankee in Retrospect

The power industry expected Connecticut Yankee and its contemporary full-scale light-water plants to pave the way for nuclear power to be on par with advanced coal-and oil-burning power stations. In the years during Connecticut Yankee's first refueling cycles, over forty power reactors were on

order. Even before the Three Mile Island accident, the number of orders was falling sharply, however. While fossil-fueled plant orders also dropped off due to a recession in 1974-75, 2/3 of the cancellations were in nuclear plants due to their much higher construction costS?38 Cancellations rose after TMI with the numbers of operating reactors peaking in 1990.339 It is doubtful that a combination of cheap coal, oil and organized protesters could have been the only factors that lilnited PWRlBWR power production as a percentage of overall U. S. megawatt hours. They were hurt by their own weaknesses and national nuclear policies: poor siting decisions, inefficient heat cycles, technological flaws, their perceived hazard, and a spent fuel liability instead of a credit. There was clearly a gap between the somewhat messianic pronouncements from theoreticians on the necessity of nuclear power no matter what the cost,340 and the more level-headed analysis of utility executives who were hoping that eventually the technology would be profitable for their stockholders.341 While the availability of enriched fuel allowed U. S. firms to build plants with lower capital costs,342 designers had to cut corners in non-safety areas to compete with fossil-fueled plants. Even a critical safety element, the emergency core cooling system pioneered by Rickover may have been shortchanged, as considerable controversy developed as to whether it would



function properly in the event of a major loss of coolant. 343 The technology of these early reactors was probably flawed from the outset (for private utilities) because of its reliance on military designs and AEC-sponsored forced development.344 The result may have been an over-reaction by the AECINRC to the operational problems leading to over-regulation (qualifying every weld) and long 1·· 1 345 lcenslng entang ements.

It is surprising the utilities did not avail themselves of other reactor systems that the AECINRC had supported. Fort St. Vrain, the now-decommissioned AEC demonstration gas-cooled plant in Colorado, showed that there were alternatives to the water reactors. The system was 5% more etllcient, produced high-pressure and high-temperature steam, had low worker rem exposure, and was resistant to loss-of-coolant accidents.346 This nuclear technology (and well-designed and -managed water reactors) is part of the reason why Europe and Japan derive such a high percentage of their electricity from nuclear power while in the United States it is now only 20%.347 A 1999 article in the New York Times compared reactors to Apollo moon rockets- a technology that slipped into history. 348 N one of the "demonstration" phase plants and few of the "commercial" phase ones operated for more than 35 years of their 40 year licenses.349 Many of 800-1000 mw plants that closely followed Connecticut Yankee, and those built under the 1973 AEC standardization program,350 have had or will need steam generator replacements or other upgrades to reach that point. NRC- mandated inspections of all reactor heads after the 2002 Davis-Besse incident undoubtedly further diminished the profitability of 1970s-era plants. 351 In spite of these costs, these reactors got a second lease on life because of deregulation in the 1990s. New power entities bought the plants from old-line utilities and applied for license extensions and up-ratings of output. 352 Undoubtedly the lessons learned from the problems of the earlier plants will enable their successors to reach and even exceed the 40-year license milestone.

In 2005, after a 30-year hiatus, four power companies have applied for site approvals for new reactors, all of which are PWR or BWR designs?53 Perhaps reflecting the damage done to the industry by the Three Mile Island accident, the main innovations in these designs appear to be that they will have simplified passive safety systems that do not require backup generators, pumps or operator actions to contain accidents, echoing one of Rickover' s goals at Shippingport. 354 The next generation is possibly just over the border in Canada, in Europe or Japan, and perhaps on digital drawing boards in designs combining heavy water or carbon moderators, fuel breeding, gas or liquid metal cooling, continuous fueling, even directly-driven gas turbines355 and intrinsic safety.

Despite built-in flaws in some of the primary and secondary systems components, Connecticut Yankee engineers and operators achieved some record performances starting with a 1977 "W orId Light Water Reactor Record Run" of 344 days. In 1984 a record 417 day run was achieved followed by a 461 day run in 1989 becoming the first plant to have twice exceeded 400 days.356 In addition it was the first internationally to produce 50 billion and later 60 billion kwh of power. In total, Connecticut Yankee generated over 110 billion kwh, saving over 67 million tons of coae57 or over 260 million barrels of oie58 during twenty-eight years of operation.

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1. Hogerton 1968: 30, Fortner 2001: 6.

2. Leverett 1955: 23 ..



NOTES

3. Engineering 1959: 336; Koch 1963: 183; Nuclear Management Company 2002:1.

4. Argonne National Laboratory 1998: 4of8. Website.

5. Nuclear Engineering 1960: 426.

6. Macintyre and Bathe 1969: 252.

7. Engineering 1947: 307; Macintyre and Bathe 1974: 252. Hewlett and Duncan 1974: 27.

8. Hewlett and Duncan 1974: 23.

9. Ibid: 27.

1 (). Duncan and Moll 1981: 5.

1l. Hewlett and Duncan 1974: 30.

12. General Electric Review 1955: 15.


14. Stephenson 1954: 80.

1 S. Ibid: 83.

16. Ibid: 170.

17. Simpson 1 961: 81.

l~. Nuclear Engineering 1958b: 435.

19. Ibid: 72.

20. Ibid: 15.

2l. Janes Fighting Ships 1967-68: 364.


23. Friend 1964: 613.

24. Ford 1955: 490.

25. Weidenbaum, Sherman and Pashos 1964: 273



2(;. Oxford English Dictionary 1989: v.17, p. 915 (Thermal. 2b. Nucl. Physics.)

27. Ford 1955: 499, Nero and Dennis 1984: 397 ..

2K Ibid: 500.


30. Haga and Stevenson 1970: 145.

31. Hogerton 1968: 23.

32. Nero and Dennis: 1984: 388.

33. Hewlett and Duncan 1974: 226; Zinn 1957: 20; Ford 1956: 728.

34. Ford 1956: 728, Zinn, Pittman, and Hogerton 1964:32.

35. Babcock and Wilcox 1960: 27-6; Klehm 1963: 47, Hoffman 1958: 78.

3(;. Agnew 1981: 61; Power 1982a: 225.

37. Ford 1956: 730.

3 K Engineering 1956a: 418.

39. Zinn 1957: 19; Nero 1979: 109.

40. Lewis and Tsui 1960: 24.

41. Duncan and Moll 1981: 6; Engineering 1958a:682; Howe 1976: 25.


43. Engineering 1956b:761, 1958b: 683.

44. Engineering 1958a: 682.

45. Hewlett and Duncan 1974: 241,257.


47. Ibid: 682.

48. Duncan and Moll 1981: 11.


50. Niland 1959: 55.


52. Stolz 1958: 10.


HAERNo. CT-I85 (Page 45)

5J. Hewlett and Duncan 1974: 251. Nuclear Engineering 1960: 426.

54. Howe 1976: 42.

55. General Electric Review 1958: 30. Nuclear Engineering 1957: 121.

5(). General Electric Review 1955: 21.


58. Ackennan 1961: 46.

59. U.S. Nuclear Regulatory Commission 1986.

60. Hogerton 1968: 23.

61. Love, Darrow and Randolph 1958: 7.

62. Brower 1958: 73~ Yankee Atomic Energy Company 2005~ Connecticut Light and Power Company 2005.

63. Haueter 1974: 200.

64. U.S. Department of Energy 2005~ Ratical.org 2005: 6.

65. Witzke and Voysey 1958: 106.

66. Smith 1960: 476

67. Kilpatrick 1960: 466.


69. Setlur 1975: 137.

70. Love, Darrow, and Randolph 1958: 5.

71. Seymore 1992: 7.

72. Babcock and Wilcox 1960: 27 -8.

7'3. Bergstrom 1959: 106.

74. Ibid: 102.

75. Stem 1964: 248; Klehm 1963: 49.

76. Office of the Federal Register 1962: 3509.

77. Stem 1964: 247.

7K Office of the Federal Register 1962: 3510.

79. Stolz 1958: 10,

80. Jackson 1971: 724.

8l. Downs, Deegan and Boggus 1959: 2.



82. Hadjian 1973: 251. Geologic and seismic siting criteria take up 9 pages in the 2004 CFR (CFR 2004: 552).


84. Lloyd 1958: 2.

85. Ackerman 1961: 48.

86. Lloyd 1958: 7.

87. Ackerman 1961: 48, Ford 1982: 45.

8K Ibid: 50.

89. Fortner 2001: 60f8.

90. Rice and Wallin 1964: 191.


92. Morabito 1964: 258.

93. Seaborg 1965: 25, Golan 1965: 196, Proceedings of the American Power Conference 1971: 48.

94. Van Nort and Copeland 1975: 213.


96. Nuclear Management Company 2002: 4. Website.

97 . Nero and Dennis 1984: 397.

98. United Electric Light and Power Company 1926: 40.

99. Engineering 1926: 285~ Baumann 1921: 504.

100. Berry and Moreton 1922: 500.

101. Power 1922: 762.

102. Hopping 1922: 483.



103. E.g., Connecticut Yankee Atomic Power Company 1966-1974: 8.2-1--8.2-5.

104. Engineering 1955b: 198.

105. Huntley 1955: 767 -5.


107. General Electric Review 1954: 5

108. Graham 1965: 210.

109. Ibid: 217.

110. Roddis 1964: 22., Edlund 1963: 128. Haley 1971:239

111. Ritchey 1964: 298.

112. Ibid: 297.

113. Agnew 1981: 63.

114. Hi1berry 1957: 27, Lischer 1965: 2, Vennard 1965: 11.

115. Simpson 1961: 82.

116. Gaines 1965: 292.

117. Reichle 1958: 140, Hogerton 1968:29.

118. Risser 1964: 584, 586.

1 19. Dillard and Baldwin 1964: 167 ~ Dunn 1964: 17.

120. Snouffer and Hanson 1964: 589.

121. Ibid: 589.

122. Bergstrom and Hamming 1965: 87.

123. Jensen 1964: 380.

124. Reichle 1958: 140.

125. Roddis 1964: 22.

126. Seaborg 1965: 24, Roddis 1965: 284.

127. Engineering 1967: 92.



128. Dillard and Baldwin 1964: 169; Hogerton 1968: 27~ Spencer & Miller 1973: 24.

129. Johnson, Fax, and Townsend 1957: 642.

130. Rickert 1964: 268.

13 1. McIntyre 1975: 21.

132. Dillard and Baldwin 1964: 169~ Hogerton 1968: 27; Spencer and Miller 1973:24.

133. Hogerton 1968: 30.

1 ]4. Connecticut Yankee Atomic Power Company 1988: 1 ~ Southern California Edison 2002: 2.

1]5. Haueter 1974: 200.

1 ]6. Connecticut Yankee Atomic Power Company 1988; Northeast Utilities 2005.

1]7. Stem 1964: 246.

138. Ibid 252.

D9. Ibid: 247

140. Connecticut Yankee Atomic Power Company UFSAR 1998: 1.2-1.

141. U.S. Atomic Energy Commission 1974.


143. Connecticut Yankee Atomic Power Company 1975.

144. Coe 1958: 75.



145. Connecticut Yankee Atomic Power Company 1987 -1994: Chapter 7, page 4.

146. Ibid: 18.

147. Brower 1958: 74, Coe 1958:75.

148. FDSA 10170: 4.1-2,4.2-13.

149. VanNoordennen 2005: Personal correspondence.

150. Connecticut Yankee Atomic Power Company 1998: 1.2-11 ~ Gray 1917: 14.

151. Morgan 1950: 9.

152. Connecticut Yankee Atomic Power Company 1987-1994: Chapter 22, page 1.

153. Church 1935: 9.

154. Morgan 1950: 8.

155. Church 1953:10.

156. Sinton 1966: 110., Spencer & Miller 1973: 24.

157. Hogerton 1968: 30.

158. Nero 1979: 21; Power 1982: 212.

159. Ibid: 210.

1 (,0. Sinton 1966: 110.

1 () 1. Babcock & Wilcox 1972: 23-1.

1 (,2. Johnson 1971: 93

1 (,3. Power 1971: 94; MacNaughton 1967: 513.

164. Dillard and Baldwin 1964: 169, Hogerton 1968: 27., Spencer & Miller 1973: 24.

165. MacNaughton 1967: 580.

1(,6. Morgan 1950: 7.

167. Johnson 1919.

168. Johnson 1919: 1100.

169. Sinton 1966:112.

170. Bauman 1921: 630.

17 1. Richardson 1911: 23 1.

172. Morgan 19S0: 11.

173. Ibid.

174. Hossli 1969: 106.,

17S. Swann 1966: 73.

17 6. Swann 1966: 73.

177. Whitaker 1981: 9.

178. Morgan 19S0: 11.

179. Hossli 1969: 110.

U~O. Morgan 19S0: IS.

1 X 1. Ibid.

182. Brown and Donahue 1964: IS.

1 X3. Sinton 1966: 113.

184. Ibid: 112~

lXS. Spencer & Miller 1973: 2S.

1 X6. Lardner 1836: lOS.

1 X7. Nagel 19S2: 46.

lX8. Lardner 1836: lOS.

lX9. Morgan 19S0: 12.

190. Johnson 1919: 1116.





191. Connecticut Yankee Atomic Power Company 1987-1993: Ch. 23, p. 12.

192. Oxford English Dictionary (1989 corrected to 1991): XV p. 44; Brown & Campbell 1952: 46.

193. Main 1893: 224.


195. Connecticut Yankee Atomic Power Company 1966-1974: 8.4-1.

196. Ibid.

197. Power 1982: 225.

198. Ibid: 226; Penner, ed., 1976 : 7l.

199. Morgan 1950: 14.

200. Koester 1908: 250.

201. Connecticut Yankee Atomic Power Company 1987-1993: Ch. 41, p. 26

202. Power 1982: 228.

203. Kinsman 2001: l.

204. Ibid: 1.

205. Merriman 1970.

206. Passer 1953: 136.

207. Hedges 1892: 165.

208. Ibid 1892: 174.

209. Dawes 1928: 104.

210. Hedges 1892: 166.

2 1 l. Parsons 1 91 1: 190.

2l2. Ibid: 22.

213. Lafoon 1950: 21.

214. Ibid: 21.

215. Power 1982: 364.

216. Lafoon 1950: 25.

217. Ibid: 25.

2 18 . Passer 1 953: 132.

219. Ibid: 135.

220. Dawes 1902, Vol. 2: 209.

221. Hedges 1892: 174.

222. Snyder 1950: 51.

223. Bowers 1982: 183.

224. Snyder 1950: 53.

225. Ibid: 57.

226. Ibid: 57.

227. Williamson 1973: 15.

228. Spencer & Miller 1973: 25

229. Scott 1973: 508.

230. Connecticut Yankee Atomic Power c1975.

231. Ibid.



232. Connecticut Yankee Atomic Power Company 1987: PDCR 886: page 7.2-1 of 13.

233. Lunn and Harvey 1970: 189.

234. Kalderon 1972: 376., Gray 1972: 389.

2~~5. Swann 1966: 81.

236. Jonas 1985: 9:

237. Chetwin 2002

2] 8. Wi day 2002

239. Gruber, H. and Reihard. K., 1979: 11.

240. Ibid: 12

241. Spencer & Miller 1973: 25.

242. Ibid: 10.

243. Ibid: 9.

244. Power 1982: 348.

245. Gray 1972: 379.

246. Ibid: 389

247. MelTiman 1970~ Clark 2003.




249. U. S. Nuclear Regulatory Commission 1988a: 1-1, 2-37.


251. Coughlin, Stark, Brown, and Johnson 1974: 226.

252. Singley, Welinsky, Whirl, and Klein 1959: 752.



255. Randazza 1975: 761; U.S. Nuclear Regulatory Commission 1995: 10.

256. Babcock and Wilcox 1972: 29-16

257. Sopocy 1984: 1013.

258. U.S. Nuclear Regulatory Commission 1988a: 1-1.

259. Hopkinson 1982: 24~ Randazza 1975: 760.

2()0. Randazza 1975: 762; Whitaker 1981: 8.

2() 1. Whitaker 1981: 9; Berge and Vignes 1981: 763; Schuecktanz, Riess, and Stieding 1981:757.

2(j2. Randazza 1975: 762; U.S. Nuclear Regulatory Commission 1988a: 2-31~ Whitaker 1981: 9~ Hopkinson 1982: 24.

263. Noll 1964: 753~ Whitaker 1981: 10.

264. Lorenzi 1952: 21.16~ Baker 1957: 707.

265. Riedel 1964: 791.

2()6. Ibid: 791.

2(;7. Mundis and Noble 1981: 7 51 ~ van Noordennen 2005.



268. Connecticut Yankee Atomic Power Company 1966-1974: 8.7-2~ Britt, Millard and DiFilippo 1975: 753.

269. Sopocy and Kovach 1984: 1014.

270. U.S. Nuclear Regulatory Commission 1988a: 2-31 ~ Randazza 1975: 762.

271. Necci 1993: 590~ Sopocy 1984: 1016.

272. Mundis and Noble 1981: 749.

273. Singley et al 1959: 755, Huffman and Malinowki 1981: 768.

274. Sopocy and Kovach 1984: 1016


276. Britt, Millard, and DiFillippo 1975: 753.

277. Connecticut Yankee Atomic Power Company 1966-1974: 8.7-2~ 1987-1994: Chapter 18, page 9.



2g0. U.S. Nuclear Regulatory Commission 1988a: 3-9.

2g1. Ibid: 1-2

2g3. U.S. Nuclear Regulatory Commission 1987.

2g4. Berge and Vignes 1981: 763.

2~5. Schuecktanz, Riess, and Stieding 1981:757.

2X6. Dvorin and Schlesinger 1984: 1005.

2'K7. Huffman and Malinowski 1981: 773.

2'K8. Mundis and Noble 1981: 751; van Noordennen 2005.

289. Power 1982: 228.



290. Trent Tube 2004; Northeast Utilities 1985: 2; Northeast Utilities 1986; Connecticut Yankee Atomic Power Company 1987-1994: Chapter 18, pages 11-12, 19.


292. U.S. Nuclear Regulatory Commission 1995: 1.

293. Wald 2002: AI.

294. Danko and Stahlkopf 1984: 654.


296. Connecticut Yankee 1975-1992: Amendment7, June 8, 1976.

297. Connecticut Yankee 1975-1992: Amendment #42. October 8, 1981.

298. Van Noordennen 2005: Personal cOlTespondence.

299. van Noordennen 2005; Office of the Federal Register 1987.

300. U.S. Nuclear Regulatory Commission 1984b: Inspection Report page 2 of8.


302. U. S. Nuclear Regulatory Commission 1984a: cover letter signed by Richard C. DeYoung.

303. U.S. Nuclear Regulatory Commission 1984e: 2.

304. U.S. Nuclear Regulatory Commission 1984b: Inspection Report page 2 of 8.

305. U. S. Nuclear Regulatory Commission 1984c: cover letter signed by Walter A. Paulson.

306. U.S. Nuclear Regulatory Commission 1984b: Inspection Report page 4 of8.

307. U.S. Nuclear Regulatory Commission 1984a: cover letter signed by Richard C. DeYoung.

308. U.S. Nuclear Regulatory Commission 1984d.

309. Northeast Utilities n.d.

3 10. U. S. Nuclear Regulatory Commission 1984a: cover letter signed by Richard C. DeYoung.



3 11. Connecticut Yankee Atomic Power Company 1985: Attachment 1, page 3.

3 12. Connecticut Yankee Atomic Power Company 1984: Attachment 4, p. 1.

313. Ibid: Attachment 2, p. 10.

314. Connecticut Yankee Atomic Power Company 1987-1994: Chapter 9: page 6.

3 15. Connecticut Yankee Atomic Power Company 1985: cover letter.

316. U.S. Nuclear Regulatory Commission 1984a: letter, Richard W. Starostecki to W.G. Counsil.

3 17. Connecticut Yankee Atomic Power Company 1985: Program Plan, p. 4.

3 18. Connecticut Yankee 1975-1992: Amendment No. 91. 12/31/86.

319. Power 1982: 228.

320. Trent Tube 2004; Northeast Utilities 1985: 2; Northeast Utilities 1986; Connecticut Yankee Atomic Power Company 1987 -1994: Chapter 18, page 11.

321. Cf. Connecticut Yankee Atomic Power Company 1987 -1994, Chapter 3: page 5.

322. U.S. Nuclear Regulatory Commission 1988: Docket No. 50-213. 1-2. Oct 4.

323. Connecticut Yankee Atomic Power Company 1975-1992; 1987-1994, Chapter 3: page 5; 1989.

324. Connecticut Yankee Atomic Power Company 1989 and 1987-1994: Chapter 9, page 1.

325. Van Noordennen 2005: Personal Correspondence.

326. Coughlin, Stark, Brown, and Johnson. 1974: 226; Connecticut Yankee Atomic Power Company 1996.

327. Union of Concerned Scientists 2005: Website 3 of7.

328. Suburbanchicagonews 2005: 3 of 4.

329. Connecticut Yankee Atomic Power Company 2005: 1.

3]0. Clauss, Harrison, and Pickens 1993: 234.

33l. Davis, Jacobson and Colburn 1993: 605.

3]2. Necci 1993: 590.

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335. U.S. Nuclear Regulatory Commission 1995: 1; Dominion 2005: 1.

336. Martel and Stahlkopf 1981: 741.

337. Slade 1993: 212.

338. SCientific American 1976: 60A.

319. Wald1999:16.

340. Hilberry 1957: 27-28.

341. Vennard 1957: 100; Simpson 1961: 81.

342. Rhoddis 1964: 26.

343. Nero 1979: 89, Ford 1982: 101.

344. Nuclear Engineering 1960: 425, Phung 1984: 889.

345. Ibid: 889; Levenson 1981: 40.

346. Agnew 1981.

347. Nuclear Management Company 2002: 5.

348. Wald 1999: 16.

349. Fortner 2001: 6.

350. Gilbert 1975: 157.

351. Wald 2002: AI.

352. Wald 2000: Cl.

353. Wald 2005: A16.

354. Ibid 2005: F3

355. Nero 1979: 11.

356. Connecticut Yankee Atomic Power Company 1988: 2

357. Palo, Emmons, and Gardner 1963: 337. 7,000 tons per day for 600 mw x 300 average power days x 28 years = 67,200,000 tons.

358. Babcock and Wilcox 1960: 2-6. 3.9 bbl equivalent of 1 ton of bituminous coal.



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1965 New Horizons in Energy. Proceedings of The American Power Conference. 27: 10-13. Chicago: Illinois Institute of Technology, Technology Center.

WaId, Matthew L. 1999 Reactors: Healthy but Dying. New York Times. March 7. Section 4. The Nation: 16

2000 Nuclear Power's Second Act. New York Times. December 20. Cl.

2002 US, Alarmed by Corrosion, Orders Checking of Reactors. New York Times. March 25. AI.

2005a Nuclear Power Hopes to Find a Welcome Mat Again. New York Times: A16. January 27.

2005b Power Producers Window-Shop for the Latest Models of Reactors. New York Times. March 15. F3.

Weidenbaum, Bernhardt, Shennan Naymark, and Thomas J. Pashos 1964 Uranium Dioxide Fuels. Proceedings of the American Power Conference 26: 273-282.


Weinberg, Alvin M. 1994 The First Nuclear Era. New York: American Institute of Physics.

Whitaker, Ralph 1981 Corrosion. EPRI Journal. 6, 7: 6-13. September.



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Zinn, Walter H. 1957 Similarities in the Technical Features of Nuclear Power Development in the United States

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Engineered Safety Systems


HAERCT-185 (Page 76)

APPENDIX A

The substitution of engineered safeguards in place of remote siting was common by the time the fIrst large commercial phase nuclear plants were constructed. While distance was an effective mediator against direct radiation streaming from a plant, it was bound to be less effective against wind driven gaseous clouds of radioactive material. The concrete in the containment shell was able to reduce radiation intensity by over a factor of five. 1 In combination with a gas-tight pressure-resisting steel vapor shell, containment design provided a passive final barrier to release of radioactive materials from the primary system during normal operations or in an accident. 2 That system alone would not have allowed construction at the Haddam site based on the model exclusion distance calculations in the Atomic Energy Commissions (AEC) 1962 report Calculation of Distance Factors for Power and Test Reactor Sites, TID-14844 which followed the main document Reactor Site Criteria (10 CFR Part 100) of April 1962.3 Reactors designed to operate at over 300 mw could reduce their exclusion distance if any releases were directed up a vent stack, since leakage was considered more dangerous at ground level. The Connecticut Yankee concrete containment building with its attached steel vapor sphere and 175-foot-high vent stack partially fulfilled the "consequence limiting" requirements. Complete compliance with AEC standards required the addition of active engineered safeguards which were specified in further AEC documents of 1963 and 1964.a

The systems installed at the plant to insure compliance included an Emergency Core Cooling System (ECCS), a Containment Spray System, In addition, many plant operating systems including the Air RecirculatIOn and Filtration System and the Residual Heat Removal System used in normal operation could be marshaled to aid the primary safety systems in the event of an accident.

Postulated Accidents

To design against accidents, engineers of nuclear plants had to postulate worst-case scenarios and then minimize ifnot eliminate the hazard to the public from a release of radiation. In a pressurized water reactor, the worst type considered was a loss of coolant accident (LOCA). This could be a double-ended rupture or complete sheering of one of the legs of the primary reactor cooling system, known as a design basis accident or a DBA.4 While fIssion would end immediately due to the presumably reliable activation of control rods, heat from decaying fIssion products could still produce 10% operating power with the rate dropping off to 5 % in under a minute. In that short period of time, however, massive damage to the core with subsequent fIssion release could occur. If the normal flow of cooling water slowed down, the fuel rods would overheat and proceed to boil off the remaining water. The steam produced by the overheating rods, and from cooling water flashing due to the break in the pressure boundary, could raise containment pressure and temperature past design limits. 5

If emergency cooling water did not reach the core soon enough, melting fuel rods could reform into a critical

aAtomic Energy Commission, "Connecticut Yankee Atomic Power Company Nuclear Plant-Unit Number One Preliminary Hazards Summary Report", Docket No. 50-213. Washington, D.C. The Commission, September 1963, and ABC, "Hazards Analysis by the Research and Power Reactor Safety Branch Division of Licensing and Regulation for Connecticut Yankee Atomic Power Company Nuclear Power Plant -Unit No.1 ", Docket No. 50-213, Washington, D. C: The Commission, March 1964. (Stem 1964: 255).



mass triggering a nuclear catastrophe.6 Tests on rods heated by electricity showed that cooling water had to be reintroduced into the core area within 20 seconds.7 The ECCS could not mitigate a catastrophic rupture of the

reactor vessel, which was considered incredible, as in that case the cooling water would disperse. The ECCS and most other engineered safety systems and summarized below.

Emergency Core Cooling System

The Emergency Core Cooling System (ECCS) was originally (1974) comprised of three separate systems: the High-Pressure Safety Injection System (HPSI), the Low-Pressure Safety Injection (Core Deluge) System (LPSI), and the Charging System. 8 Later, the Residual Heat Removal System described below was classified as part of the ECCS.9 The systems automatically delivered water to the reactor vessel to cool the core after a loss of coolant accident. In post accident recovery operations the system provided long term core cooling. In the event of a steam line break, the ECCS provided reactivity control. The system was derived from the safety injection systems at Shippingport and Yankee Rowe. 10

Five minimum requirements for ECCS performance were set by the AEC in 10CFR50.46, although some sections of the system could not meet the passive requirements because the plant was designed and built previous to these rulings: 11

• Peak Cladding temperature of Zircaloy fuel rods was limited to 2,200 0 F to limit chemical reactions between the water or steam. The stainless cladding at Connecticut Yankee was allowed a higher limit.

• Maximum Cladding Oxidation was limited to less than 2% of the clad thickness to insure integrity.

• Maximum hydrogen generated from a reaction between the cladding and steam or water was limited to prevent an explosive concentration in containment after an accident.

• Long term cooling had to have sufficient capability to provide decay heat removal from the reactor.

• Extensive redundancy of components, interconnections, leak detection, and power supplies was required, and described as a single failure criterion: every active component (all valves and pumps) had to be duplicated. 12 To achieve that level of protection, the system was configured as two separate trains of equal capacity each consisting of a charging pump, a high pressure injection pump, a low pressure safety injection pump and a residual heat removal pump. The system could be operated locally if the control room became uninhabitable. 13 All the pumps were located in the Primary Auxiliary Building ([P AB], see HAER No. CT -185-G) The system had to be operable whenever the reactor coolant temperature was greater than 350 0 F.

The automatic feature of the ECCS required the system to be tied in with sensors to components that would be reliably and measurably impacted by the postulated accident conditions. Low pressurizer pressure and high containment pressure were the initiation parameters. 14 Dual safety injection relays with manual or solenoid tripping started the water flow when the pressurizer signaled a drop below 1700 psig resulting from coolant leaving the system. Electrical control functions were powered from 125-volt DC buses with battery backup and provision for powering from AC buses via AC-to-DC-changing motor generators. Immediately after accident



initiation, the charging pumps, high- and low-pressure safety injection pumps pulled borated water from the Refueling Water Storage Tank and injected it into the Reactor Coolant System (the injection phase). Water leaving the system was collected in the containment sump and recirculated by the Residual Heat Removal

Pumps and heat exchangers back into the RCS (recirculation phase). 15 Safety injection accumulator tanks which had been used in other PWR plants were not installed at Connecticut Yankee. 16

Safety Injection Initiation

Redundancy of the critical automatic signal came from two independent circuits with tripping relays on the control board (see HAER No. CT-185-F) Multiple pressurizer channels (sensors) de-energized relays on a pressure drop to 1,700 psig, energizing another relayb which in turn energized a solenoid which tripped another relay which then actuated numerous contacts in the system. 17 Automatic actions which followed included: Reactor Trip, Diesel Generator Start, Fans start, trip pressurizer heaters (to insure readiness for cool down), open Safety Injection Loop stop valves and core deluge valves, and signaled "Core Cooling Actuated" alarms and annunciators. Numerous relays, blocking circuits, and switchboard interlocks prevented operators from taking the wrong actions during core cooling and depressurization. 18 The second major initiator of safety injection was an indication of high containment pressure which would directly result from the loss of heated pressurized inventory in the RCS. At 5 psig overpressure (or a safety injection signal) circuits automatically closed many paths through the containment boundary to limit fission product release to the atmosphere. 19

Trains of relays and solenoids initiated automatic events through the plant to control the accident.

The two High-Pressure Safety Injection Pumps provided the entire motive force for the high head safety injection to the four coolant 100ps.2o They were horizontal, 6-stage C l,250 hp, 1750 gpm, 1,400 psi discharge centrifugal pumps operating against a head (pressure) equal to a column of water over 2,000 feet high. The stages were arranged with a single entry stage and the 5 following stages separated by the pump motor. Power came from the 4000 vac emergency system with each pump on a different bus. Control circuit breakers were operated by 125 volt dc for redundancy. Water flow was directed to the cold legs of the RCS (between the coolant pumps and the reactor vessel entrance nozzles) to avoid flow into the steam generator u-tubes with subsequent reverse steam generation (from the hot feed water) which could actually lead to higher core temperature. 21 The four safety injection loop motor operated stop valves were normally closed and would open automatically on the safety injection signal. Injection flow was maintained until manually switched to recirculation by operators or until the R WS T inventory fell below a set point.

Low-Pressure Safety Injection Pumps were single-stage 1,000 hp, 5550 gpm, 350 psi discharge, centrifugal types operating against a 590 foot head. Power was as for the HPSI pumps. They supplied through 6-inch-

b Relays were battery powered electro-magnetic devices used in early telegraph systems to extend transmission length. In 20th century control systems, components were activated by trains of energizing and deenergizing relays (Oxford English Dictionary 1989: 556)

C Multi -stage pumps had two or more impellers with each stage boosting the pressure in succession to achieve higher output pressure.



diameter motorized/manual valves the core deluge piping that entered the reactor head through spare control rod penetrations. d The cool borated water entered the core directly, aiding the high pressure injection flow coming

in through the cold legs. The LPSI pumps also supplied the Containment Spray system22, ftre ftghting spray for

the charcoal filters and #2 loop of the RCS for normal heat removal. Once recirculation was begun, the lowpressure pumps were shut down.

Charging System and Chemical and Volume Control System

lvlany plant systems had multiple functions and crossed into different buildings from their base location. The Charging System was a portion of the Chemical and Volume Control System (CVCS) which also acted as part of the ECCS by serving as an additional source of high injection. 23 The two charging pumps were horizontal, 13-stage, 2200-gpm, I50-psi discharge centrifugal pumps. Starting was automatic, with power and control from the same group of buses as the HPSI and LPSI pumps. When working in the CVCS system, the charging pumps pumped from and into the Volume Control Tank in the P AB, the reactor coolant loop #2 cold leg, and reactor coolant pump seals. On safety injection initiation they automatically started to inject cool borated water from the Refueling Cavity Water Storage Tank into loop #2. The CVCS then was utilized as part of the ECCS during the recirculation phase.24

Residual Heat Removal System (RHRS)

After shutdown the ftssion products continued to decay producing relatively large amOlll1ts of heat. The RHRS provided decay heat removal and circulation through the coolant loops when the reactor coolant pumps were not operating. The recirculation ensured that the boron injected into the RCS for shutdown was evenly concentrated through the system. 25 After being removed from the reactor at 300 F. the coolant was passed through heat exchangers for cold shutdown. The RHRS was not used during normal operations but was lined up for standby ECCS operation. In a Loss of Coolant Accident (LOCA) the system pumped spilled water from the containment sump and recirculated it through the heat exchangers and back into the reactor for long term decay heat removal.

The system had no role in the injection phase, but was manually activated to provide subsequent circulation. 26

The RHRS pumps pulled from the containment sump, collecting the spilled water and sending it through the Residual Heat Removal (RHR) heat exchangers to be cooled by the service water. It was then pumped into two cold- leg motor-operated valves. Other engineered safeguard components which could receive cooled water were the purification system, charging pump suction, Core deluge, containment spray and charcoal ftlter spray and High Pressure Injection pump suction. 27 In addition, during refueling, the RHRS was used to transfer water between the Refueling Cavity Water Storage Tank and the Refueling Cavity.

d The core deluge valves would fail to function (open) on loss of power. They were provided with hand wheels but it was accepted that during a serious accident, the radiological conditions would prevent access (Connecticut Yankee Atomic Power Company 1987-1995: Chapter 5, page 4l.)



Like many plant systems the RHRS was duplicated with two pumps, and two heat exchangers. The pumps were horizontal single stage centrifugal 2,200-gpm, ISO-psi discharge units located in the RHR pit in the Primary Auxiliary Building. Each unit could carry half of the maximum pumping load allowing reasonable flow even if one unit was out of service.28 Power came from the 480v AC buses with 12Sv DC control. The heat exchangers were shell and U-tube, SOO-psig pressure vessels with cooling water in the shell supplied by the

component cooling water system. 29 During accident conditions, supply was transferred to the Service Water System which was classed as a safety system with higher reliability. As a result of the cavity seal failure in 1984, the RHRS was required to be operable or operating during the entire refueling process to provide circulation.

Water Supply to ECCS

The 2S0,000-gallon Refueling Cavity Water Storage Tank was the reservoir of borated water for the ECCS and for filling the reactor cavity during refueling?O The high-pressure injection, charging, and RHR pumps pulled from a 16-inch-diameter supply line, while the low-pressure injection was fed from a separate 18-inch-diameter line. The tank could supply 100,000 gallons to the RHR for recirculation at which point it was stopped manually at around the 130,000 gallon leve1.31

Containment Sump

The Containment sump and pumps, were designed for use in a serious Loss of Coolant Accident (LOCA), since the 2,000-gallon capacity of the in-core instrumentation ICI sump would quickly be reached leaving the entire ground floor of Containment to become the sump for up to 32,000 to 3S,000 gallons ofwater. 32 Overflow water was automatically treated with tri-sodium phosphate from flooded baskets in the area of the sump to control the pH, reducing iodine release into containment atmosphere. In the Primary Auxiliary Building, the Residual Heat Removal pumps in the recirculation mode pulled water from the sump to the High-Pressure Safety Injection, Low-Pressure Safety Injection, and Charging pumps - components of the Emergency Core Cooling System (ECCS) - for addition to the coolant loops.

Auxiliary Feed water and Other ECCS Systems

As noted above under Injection Initiation, relays automatically closed off penetrations in the containment boundary to prevent releases. Sometime after 1990 the various shut-offs were designated the Containment Isolation System. Control of the heat and pressure produced by a LOCA required numerous active "heat sinks," as well as the passive heat sink effects expected from the concrete mass of the containment building, shield walls, and equipment. 33

Possibly in response to regulatory changes, the Auxiliary Feed water System (AFW) was also included as an engineered safeguard. e It functioned on shutdown of the main feed water system to provided feed water flow to

e The AFW system was a causative factor in the TMI accident. Operators there had isolated the pumps for testing and failed to restore them, leading to a complete loss of all feed water flow for several critical minutes. (Connecticut Yankee Atomic Power Company 1987-1995: Chapter 21, page 60).



the steam generators so they could continue to absorb heat from the RCS.34 In addition to their role in steam

production for the steam turbine, the steam generators also served as a heat sink for reactor decay heat after shutdown and in accident conditions. To ensure a continuous un-interruptible supply of feed water if the main Feed water system was not functional, Connecticut Yankee had an Auxiliary Feed water System supplied from the Terry Turbine building named after two Terry Turbine driven auxiliary feed pumpS.35 The Terry turbine was developed by the Terry Steam Turbine Company of Hartford, CT early in the 20th centwy, 36 and became a favored prime mover in the power industry for driving fans and boiler feed pumps due to its ruggedness, relatively high efficiency and high speed.37 The single forged turbine wheel had multiple semi-circular "buckets" machined directly in the forging and reversing chambers in the surrounding casing. Steam was admitted directly into the buckets moving them from the impact. Steam was then turned 180 degrees and readmitted to the blades several times until most of the energy was gone. The unit had very large clearances between the turbine wheel buckets and the reversing chambers for reliability and could even continue to operate if the steam supply turned to water. 38

The pumps were 450-gpm multi-stage centrifugal types which supplied their own lubrication, shaft sealing, and cooling, independent of plant systems. The Terry turbines were the only rotative steam powered auxiliaries in the plant. Even if all plant electric power and backup diesel generators were lost, the Terry turbines could continue to provide pumping power from steam produced by decay heat from the reactor. 39 Steam supply was from the #3 and #4 steam generators through the atmospheric steam dump valve supply header.40 Special throttling control valves with greater reliability than other types were operated by the plant control air system and would automatically open if the system failed. The system automatically activated if the circuit breakers on the main pumps tripped, or low water level was sensed in two of the steam generators. In addition to the two Terry turbines, a third component of the Auxiliary Feed water System was a manually-operated electric-motordriven pump located in a separate enclosure south of the Terry Turbine Building. The 725-gpm motoroperated pump could supply feed water when the Condensate and Feed water systems were manually shut down, and provided a back up source if the main pumps and Terry Turbine pumps were involved in a loss of feed incident.41 On account of its important role in Reactor Coolant System heat removal, the Auxiliary Feed water System was maintained as part of the plant Engineered Safety Systems.42

Supply water for the Auxiliary Feed water System came from the Demineralized Storage Water that was located within a protective concrete shield wall south of the Terry Turbine Building. Protection from freezing was insured by a plant heat trace circuit system.

The Reactor Protection System had dual functions. During normal operations, the system maintained the integrity of the fuel and the RCS loops during severe load change and component failure transients resulting from loss of feed water, loss of coolant flow and other conditions which would be expected to occur during the life of the plant. 43 The RPS also worked with the Engineered Safety Systems to limit the amount of fission products released into the atmosphere during an accident.

Operating conditions requiring protection from the RPS were categorized by the estimated frequency or probability of occurrence. Condition 1 Events occurring daily or yearly included normal steady state operation



with normal startup, ramped power changes, shutdown and refueling. Condition 2 Events resulted in reactor trip without damage to fuel or RCS overpressure. These events, which included loss of feed water, loss ofload and improper dilution of the boric acid, were not expected to prevent the plant from returning to operation and were expected to occur only once a year. During Condition 1 and 2 events, the RPS automatically tripped the reactor to maintain integrity of the fuel and RCS. Safety limits were set by a document called the Technical Specifications which covered all aspects of the plant. Core temperature had to be limited to prevent release of fission products into the RCS loops. Pressure had to limited in the loops or fission products could be released into containment.

Condition 3 Events included small LOCAs, fuel assembly miss loading, and minor steam pipe breaks. Consequences were possible damage to only a few fuel rods, could preclude return to operation and were expected only once in the life of the plant. Condition 4 Events were the worst case scenarios of postulated accidents, including major LOCAs, steam pipe breaks, steam generator tube ruptures, seizure of reactor coolant pumps, ejected fuel rods, and major fuel handling accidents. Though designers never expected them to occur, they were analyzed to determine what if any remedial measures could be taken and what the effects of the postulated accidents would have on the plant, operators, and immediate or surrounding population centers. During Condition 3 and 4 Events, the likely release of fission products into Containment required the RPS to assist the engineered safeguards in restricting release from Containment into the atmosphere around the plant.

The ultimate criteria basis for RPS function was a level of radiation exposure in rems that a person within the exclusion area radius would be subjected to for a period of two hours after the release. That level of 25 rem over their whole body or 300 rem to the thyroid was set by the Code of Federal Regulations in Chapter 10, Part 100. A similar level was set for anyone in the outer boundary of the low population zone.

Since the ECCS, the most important component of the engineered safeguards could not remove all the heat from a reactor at full power, the Reactor Protection System was an essential element of the shutdown system.44 The critical role was to prevent the failure of the first level of fission product protection, the fuel cladding. It was not possible to actually measure the departure from nucleate boiling (DNB) point during operation, but thermal power and reactor coolant temperatures and pressures could be used to correlate a parameter. A ratio of heat flux to DNB was set up and if it was exceeded the reactor protection system described below would trip the reactor. 45

The RPS protection system had two separate trains that received inputs from instmmentationcircuits at critical components. The inputs could activate protection functions if "coincidences" occurred twice in each train for reliability. On the basis of those inputs the system produced control signals to actuate protective interlocks and reactor trips. The interlocks prevented plant components from operating in a way that could interfere with the accident control process. The various reactor trip signals opened circuit breakers and de-energized the CDRMs allowing all the rods to drop in the core. The entire system was designed with the military dictum "defense in depth" in mind.46 In addition to redundant and cross checked outputs, the system was designed to be fail-safe: a reactor trip would occur if power was lost to either a protection signal channel or the reactor trip breakers.

Main Steam Line Break (MSLB)


HAER CT-185 (Page 83)

Many types of malfunctions of the reactor and reactor coolant system components were controlled by the Reactor Protection System without activation of the ECCS. A break in any of the main steam lines coming out of the steam generators could be serious enough to require a reactor trip because of resulting cooling of the RCS loops and drop in pressurizer level. A rupture of the 36-inch-diameter main steam header in the Turbine Building Auxiliary Bay could be quickly isolated from the steam generators and reactor by the isolation valves without cool down. A rupture up-stream of the valves in the 24-inch-diameter steam lines exiting the generators was more serious. In that event the Safety Injection would be actuated from the pressurizer signals to insure a flow of borated cooling water to prevent the reactor from returning to power after the trip.47

Containment Air Recirculation System (CAR)

The CAR system provided cooling and recirculation of the Containment atmosphere during normal operations with four air recirculation units. Air flow was bypassed only through the cooling coils and fans. During a LOCA or Main Steam Line Break, steam would be released into Containment, raising the pressure and temperature above the design limits. The CAR system was designed to keep conditions within specification while other emergency systems were in operation. On a safety injection signal or rise in containment pressure, the CAR system would automatically start to cool, and re-aligned itself to provide cooling and depressurization of the atmosphere by activating pre-filters and charcoal filters for post-LOCA iodine removal. After an accident the units filtered the air to reduce particulate and iodine concentrations.

The face dampers automatically opened if the containment air pressure rose above a set point. The air-powered dampers used DC power for activation with battery back-up. In addition, they were spring loaded to open if all supplies failed. In conjunction with that, the bypass dampers would fail shut to insure the "safeguard condition" accident flow path. Since moisture in the post accident air flow could reduce the effectiveness of the filters, the air was first passed through two stages of removal: a chevron separator and fiberglass pad mist eliminators. The pleated glass asbestos particulate filter removed solid matter that could foul the critical charcoal filter during post-accident recirculation. The charcoal filters were arranged in banks of 120 two inch thick cartridges in each unit. They were expected to be 99% efficient at removing radioactive organic iodines within two hours after an accident via isotropic exchange.48 Temperature sensors sent alarms to the control room if the filters heated beyond 325 F during post accident filtration. The Residual Heat Removal System provided fire protection water sprays to the filters. 49 The last stage of air handling was the cooling coil section for both normal and post accident operation. The transverse flow finned-coil banks were supplied by the Service Water System.

The calculations used to determine the design parameters for reduction of iodine (95-99%) in containment post accident were derived from testing done at Oak Ridge National Laboratory (ORNL).50 While the range of factors considered was wide, an undocumented party questioned whether they actually covered conditions in a severe LOCA with temperatures over 250 F, pressures in the 30-40 psig range, and 100% relative humidity. During 1966 Connecticut Yankee arranged to do full scale tests under incident conditions. It is undocumented how the conditions would be replicated and what the results were. 51 At the same time ORNL noted that if filters were wet their ability to remove methyl iodide was only about 13 -54%. Plant management resolved to do full scale testing with accident condition air stream mixtures. 52 Expecting verification from testing, the



Connecticut Yankee FDSA stated" ... conclusively that the filtration system is a reliable and efficient means of protection against any incident which might release fission products to the containment atmosphere. 53

During the life of the plant, some operational deficiencies were discovered in the CAR units. In 1974 a number of the charcoal filter modules were found to be fouled with boric acid deposits (from the RHRS) due to partially closed spray valves. It was determined that the filter flow in an accident may have been compromised requiring better attention to valve alignments. 54 In 1984 an Integrated Plant Safety Assessment found that Service Water flow to the coolers may have been over estimated due to changes in valve settings. Also in 1984, it was found that three of the fans could not meet the 50,000 cfm flow required in an accident due to out of adjustment fan controls and leakage. It was discovered that internal specifications for normal operation had not actually required that amount, while specifications for LOCA conditions did. 55

Containment Spray System f

The Containment Spray System was installed as a backup to the CAR system for depressurization of containment after a Loss of Coolant Accident. 56 If during an accident pressure exceeded the 40 psig design limit by 10 pounds, containment spray valves were opened by the personnel in the Control Room. In the Primary Auxiliary Building, Low-Pressure Safety Injection pumps or the Residual Heat Removal pumps forced water into a spray ring in the containment dome which spread out and fell down over 60 feet, absorbing heat and iodine from the atmosphere. Spray water would also provide an additional liquid film barrier over the steel liner. Water was collected in the Containment sump for recirculation by the residual heat removal pumps which supplied the system. 57 Initiation in a LOCA was not automatic as the priority was to have the RHR pumps lined up to supply the reactor first.gWater could also be taken from the Refueling Cavity Water Storage Tank. If flow from the ECCS was limited, a separate inexhaustible supply was available from the Connecticut River through the Fire Water System powered by diesel driven pump.58 At the time the system was designed, it was acknowledged that the experimental work done was not sufficient for accurate prediction of the efficacy of the spray system in actual accident conditions. 59 The valves would fail on power loss in the closed position60 and it is unclear how and if operators would be allowed to access the hand wheels during accident conditions if diesel powered backup power failed.

ECCS Controversy

The accuracy of some of the Engineered Safety System design bases were questioned in the 1966-1974 Facility Description and Safety Analysis, and some operational deficiencies were noted in the 1987-1995 Plant

f In the Plant Information Book, both the Charcoal Filter Spay System and the Containment Spray System were considered part of the Emergency Core Cooling System (Connecticut Yankee Atomic Power Station 1987-1995: Chapter 5, page 15).

g Use of the Residual Heat Removal pumps for spray header supply was not a "proceduralized" acceptable source of spray water. (Connecticut Yankee Atomic Power Company 1987 -1995: Chapter 5, page 78.)

HADDAM NECK NUCLEAR POWER PJ-"ANT (Connecticut Yankee Nuclear Power Plant)


Infonnation Book. While no such problems in the ECCS were noted in either document, the efficacy of that system was questioned by some in the industry and by nuclear watchdog groups. There were doubts about a number of assumptions made by the designers of several manufactures' plants regarding the projected way the ECCS would function: 61

• Flow blockage. Critics doubted that the design process considered the possibility that fuel rods left exposed by coolant draining out at high pressure might heat up rapidly and expand before the core deluge and core cooling water could reach them. The rods were so closely positioned that there was concern that swollen rods might prevent arriving emergency coolant to reach the overheating sections, leaving them to fmally rupture and release the encased radioactivity.

• Chemical Reactions. There was concern that the Zircaloy rods used in some stations would undergo physical changes (weakening) at well below the melting point (2,300° F) that the system was designed to control. 62 Whether the stainless steel clad rods in the Connecticut Yankee core were subject to the same problem is un-documented. As noted above, there was concern about the production of hydrogen gas from a stainless steel/coolant reaction.

• ECCS Bypass. There were concerns that the emergency flow could bypass the core and blowout through the leaking section of reactor coolant piping. 63

• Conflict of Interest. Tests on the ability of core cooling flow to remove heat were done by the manufacturers, not by outside, independent laboratories. 64

As a result of these issues, the AEC held hearings from January 1972 to December 1973 which determined that the issues had been resolved.65 The report produced was met with further criticism resulting in a year-long follow-up study that generally supported the earlier methodology.66 The 1979 Three Mile Island accident was not an effective indicator since operators shut off the ECCS before it could be effective. As of 1980 no full scale tests on the system had been conducted. 67

1. Stem 1964: 244.



NOTES


3. Stem 1964: 253.

4. Howe 1976: 39.

5. Connecticut Yankee Atomic Power Company 1966-1974: 10.3.1-7.


7. Ibid: 41

8. Connecticut Yankee Atomic Power Company 1966-1974: 5.2.7-1.



1 1. Connecticut Yankee Atomic Power Company 1987-1995: Chapter 5, page 4.

12. Ibid: 6.

I:L Ibid: 115.

14. Ibid: 86

15. Ibid: 3.

16. Connecticut Yankee Atomic Power Company cl97 4: 2-2.


18. Ibid: 90.

19. Ibid: 94.

20. Ibid: 23.

21. Ibid: 8.

22. Ibid: 10.

23. Ibid: 11, 41.

24. Ibid: 12.



25. Connecticut Yankee Atomic Power Company 1987 -1995: Chapter 6, page 1.

2(). Ibid: 14, Houff 2006: personal correspondence.


2g. Ibid: page 8.

29. Ibid: Chapter 5, page 66.

30. Ibid: 17.

3 1. Ibid: 23.

32. Ibid: 21, Chapter 7, page 22 ..

3]. Ibid 1987-1995: Chapter 7, page 37.

34. Connecticut Yankee Atomic Power Company 1993-1998: 1.2.4.4.3-4.

35. Connecticut Yankee Atomic Power Company 1987-1995. Chapter 21, page 1.

3(). American Society of Mechanical Engineers 1920: 40.


3g. Connecticut Yankee Atomic Power Company 1987-1995. Chapter 21, page 13.

39. Houff 2006: personal communication.


41. Ibid: page 16.

42. Ibid: 31.

4]. Ibid: Chapter 75, page 1.

44. Ibid: 5.

45. Connecticut Yankee Atomic Power Company 1966-1974: 4.3-l.


47. Ibid 1966-1974: 10.3.3-1

48. Ibid: 9.

49. Ibid: 5.




51. Ibid

52. Ibid: 3.6-5.

53. Ibid: 3.6-6

54. Connecticut Yankee Atomic Power Company 1987-1995: Chapter 56, page: 63.

55. Ibid: 67.



58. Connecticut Yankee Atomic Power Company 1966-1974: 3.6-6A.

59. Ibid: 3.6-6.


61. Howe 1976: 41, Nero 1979: 89.

62. Weinberg 1994: 197.

63. Connecticut Yankee Atomic Power Company 1987-1995: 87

64. Ford 1982: 104.

65. Ford 1982: 127, Howe 1976: 41.

6(). Lewis 1980: 59.

67. Ibid: 58.

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ER

PL

AN

T

(Con

nect

icut

Yan

kee

Nuc

lear

Pow

er P

lant

) H

AE

RN

o. C

T-1

85

(Pag

e 89

)

AP

PE

ND

IX B

-SU

MM

AR

Y O

F ST

RU

CT

UR

ES,

PR

IMA

RY

FU

NC

TIO

NS,

AN

D M

AJO

R E

QU

IPM

EN

T

NA

ME

AN

D D

AT

ES

SC

RE

EN

WE

LL

HO

US

E

1964

-196

6

RE

AC

TO

R C

ON

TA

INM

EN

T

1964

-196

6

SU

MM

AR

Y D

ES

CR

IPT

ION

Scr

eenw

ell:

con

cret

e st

ruct

ure

±6S

'xI9

.S' w

ith 4

tr

ash

rack

s o

n c

hain

ass

embl

ies,

eac

h ±

12'

wid

e x

39' h

igh;

fis

h fr

ight

ener

and

de-

icin

g fa

cili

ties

P

umpw

ell:

st

eel-

fram

ed

supe

rstr

uctu

re

wit

h in

sula

ted

Gal

best

os s

idin

g &

alu

min

um lo

uver

s,

31.5

'x78

.8',

±IS

' hi

gh a

bove

gro

und

el.

21.5

'; co

ncre

te f

ound

atio

ns t

o te

l. -

18'

circ

ular

do

med

re

info

rced

co

ncre

te

stru

ctur

e w

ith

wal

ls u

p to

4.5

' thi

ck,

144'

ou

tsid

e di

amet

er,

170'

hig

h ab

oveg

roun

d w

ith

70'-h

igh

dom

e se

ctio

n; 2

18'

abov

e bo

ttom

of s

ump

at e

l. -2

3';

maj

or fl

oors

at

els.

-1

9.6

',1.5

',16

',22

',48

.5'.

PL

AN

T S

YS

TE

M

SE

RV

ICE

WA

TE

R

CIR

CU

LA

TIN

G

WA

TE

R

RE

AC

TO

R

VE

SS

EL

RE

AC

TO

R

CO

OL

AN

T S

YS

TE

M

MA

IN S

TE

AM

SE

RV

ICE

WA

TE

R

EM

ER

GE

NC

Y

CO

RE

CO

OL

ING

LO

CA

TIO

N

Pum

pwel

l

Pum

pwel

l

el.

16'

el.

22'

el.

16'

el.

1.5'

el.

16'

on to

p o

f rea

ctor

ve

ssel

MA

JOR

EQ

UIP

ME

NT

(us

uall

y in

seq

uenc

e o

f use

)

4 Se

rvic

e W

ater

Pum

ps (

P3

7-1

A1

1B

/1C

/1D

)

4 C

ircu

lati

ng W

ater

Pum

ps (

P3

4-1

A1

lB/I

CIl

D)

Rea

ctor

Ves

sel

(E-1

), C

ontr

ol R

od D

rive

Mec

hani

sms

Rea

ctor

4 S

team

Gen

erat

ors

(E-6

-1 1

2/3 1

4)

4 R

eact

or C

oola

nt P

umps

(P

-17-

112/

3/4)

Pre

ssur

izer

(E-8

-1 )

Pre

ssur

izer

Rel

ief T

ank

(TK

-8-1

)

8 m

otor

-ope

rate

d L

oop

Isol

atio

n V

alve

s, 4

mot

or -o

pera

ted

Loo

p B

ypas

s V

alve

s, 4

Loo

p O

ver

Pre

ssur

e C

heck

Val

ves

4 S

team

Gen

erat

ors

(E-6

-1/2

/3/4

)

4 C

onta

inm

ent A

ir R

ecir

cula

tion

Fan

Mot

or C

oole

rs

(E-7

7-1!

2!3!

4)

4 C

onta

inm

ent A

ir R

ecir

cula

tion

coo

ling

coi

ls

(E-3

7-1I

2!3!

4)

2 m

otor

-ope

rate

d va

lves

sup

plie

d bo

rate

d w

ater

from

Low

P

ress

ure

Saf

ety

Inje

ctio

n P

umps

(P

-92-

1A1l

B)

in P

rim

ary

Aux

. Bld

g.

4 he

ader

s to

eac

h re

acto

r co

olan

t sys

tem

loop

char

ging

hea

der

to L

oop

2 co

ld le

g, f

or b

orat

ed w

ater

fro

m

HA

DD

AM

NE

CK

NU

CL

EA

R P

OW

ER

PL

AN

T

(Con

nect

icut

Yan

kee

Nuc

lear

Pow

er P

lant

) H

AE

RN

o. C

T-1

85

(Pag

e 90

)

AP

PE

ND

IX B

-SU

MM

AR

Y O

F ST

RU

CT

UR

ES,

PR

IMA

RY

FU

NC

TIO

NS,

AN

D M

AJO

R E

QU

IPM

EN

T

NA

ME

AN

D D

AT

ES

RE

AC

TO

R C

ON

TA

INM

EN

T

(con

t.)

TU

RB

INE

BU

ILD

ING

19

64-1

966

SU

MM

AR

Y D

ES

CR

IPT

ION

Mai

n B

uild

ing

exte

rior

: ga

ble-

end

stee

lfr

amed

str

uctu

re 2

68'x

l10'

, 12

2' h

igh

with

m

etal

roo

f;

enam

eled

flu

ted

alum

inum

si

ding

exc

ept s

outh

wal

l of c

oncr

ete

bloc

k &

gl

azed

bri

ck;

4.25

'-hig

h br

ick

base

on

nort

h &

wes

t sid

es;

on w

est s

ide,

4 a

rche

d lo

uver

ed a

lum

inum

pan

els

abov

e br

ick

PL

AN

T S

YS

TE

M

CH

EM

ICA

L &

V

OL

UM

E C

ON

TR

OL

CH

EM

ICA

L &

V

OL

UM

E C

ON

TR

OL

(c

ont.)

CO

NT

AIN

ME

NT

S

PR

AY

CO

NT

AIN

. A

IR R

E

CIR

CU

LA

TIO

N

CO

NT

AIN

ME

NT

A

IR

CIR

CU

LA

TIN

G

WA

TE

R

LO

CA

TIO

N

Reg

ener

ativ

e H

eat

Exc

hang

er C

ubic

le

(el.

22')

spra

y he

ader

rin

g in

side

Rea

ctor

C

onta

inm

ent l

iner

at

el.

110'

el.

22'

el.

22'

MA

JOR

EQ

UIP

ME

NT

(us

uall

y in

seq

uenc

e o

f use

)

Cen

trif

ugal

Cha

rgin

g P

umps

(P

-l8

-lA

/lB

) in

Pri

m.

Aux

. B

ldg.

Rea

ctor

Coo

lant

Sys

tem

Let

dow

n: m

otor

-ope

rate

d is

olat

ion

valv

e (L

D-M

OV

-200

)

Rea

ctor

Coo

lant

Sys

tem

Let

dow

n: 3

Reg

ener

ativ

e H

eat

Exc

hang

ers

(E-7

-1A

1lB

IlC

)

Rea

ctor

Coo

lant

Pum

p Se

al W

ater

Inj

ectio

n: R

eact

or

Coo

lant

Pum

p Se

al P

acka

ge

nozz

les

for

bora

ted

wat

er p

umpe

d fr

om 4

sou

rces

in

Pri

mar

y A

uxil

iary

Bui

ldin

g: 2

Res

idua

l Hea

t Rem

oval

Pu

mps

, P

-14-

1A11

B a

nd L

ow P

ress

ure

Safe

ty I

njec

tion

Pu

mps

, P

-92-

1A11

B

4 fa

n/fi

lter

unit

s (F

-17 -

112/

3/4)

2 C

onta

inm

ent A

ir S

yste

m c

ompr

esso

rs (

C--

IA1

1B

)

el.

22'

Air

Rec

eive

r (T

K-9

2-1A

)

el.

22'

Pre

-Fil

ter

(FL

-96-

1A)

el.

22'

Air

Dry

er (

FL

-98-

1A1l

B)

el.

22'

Aft

er-F

ilter

Mai

n G

roun

d F

loor

2

Mai

n C

onde

nser

s (E

-23-

1A, E

-23-

1B),

eac

h w

ith

2 in

let

& 2

out

let w

ater

boxe

s, a

nd 1

pri

min

g ta

nk

Mai

n M

ezza

nine

V

acuu

m P

rim

ing

Tan

k (T

K-2

7-1A

)

Mai

n G

roun

d F

loor

2

Vac

uum

Pri

min

g P

umps

(P

-36-

1A, P

-36-

1B)

HA

DD

AM

NE

CK

NU

CL

EA

R P

O\V

ER

PL

AN

T

(Con

nect

icut

Yan

kee

Nuc

lear

Pow

er P

lant

) H

AE

RN

o. C

T-1

85

(Pag

e 91

)

APP

EN

DIX

B -

SUM

MA

RY

OF

STR

UC

TU

RE

S, P

RIM

AR

Y F

UN

CT

ION

S, A

ND

MA

JOR

EQ

UIP

ME

NT

NA

ME

AN

D D

AT

ES

TU

RB

INE

BU

ILD

ING

(c

ont.

)

ca.

1978

1979

-198

2

SU

MM

AR

Y D

ES

CR

IPT

ION

base

(2

pane

ls r

emov

able

), &

pai

red

2.5'

w

ide

plas

tic

win

dow

sec

tion

s at

alt

erna

te

colu

mns

; o

n e

ast s

ide,

107

' of c

onti

nuou

s al

umin

um-s

ash

win

dow

s at

el.

62.5

'; pl

asti

c w

indo

w p

anel

s in

nor

th a

nd s

outh

pe

dim

ents

A

uxil

iary

Bay

ext

erio

r: f

lat-

roof

ed s

tee1

-fr

amed

str

uctu

re 2

40-2

68'x

27',

38' h

igh

at

sout

heas

t co

mer

of

mai

n b

uild

ing,

wit

h en

amel

ed fl

uted

alu

min

um s

idin

g In

teri

ors:

ci

rcul

atin

g w

ater

int

akes

and

di

scha

rges

to

el.

-6, w

ith

disc

harg

e tu

nnel

in

clud

ing

de-i

cing

line

inta

ke; i

n m

ain

and

auxi

liar

y se

ctio

ns,

grou

nd fl

oor

el.

21.5

', m

ezza

nine

lev

el e

l. 37

.5;

mai

n s

ecti

on

rehe

ater

leve

l el.

47.5

', op

erat

ing

floo

r el

. 59

.5' w

ith

cran

e ra

il to

p el

. 98

.25'

. F

rom

gr

ound

floo

r, r

einf

orce

d co

ncre

te p

edes

tals

su

ppor

ted

turb

ines

&

ge

nera

tor

on

op

erat

ing

floo

r; s

teel

-fra

med

mez

zani

ne &

re

heat

er le

vels

ind

epen

dent

fro

m tu

rbin

ege

nera

tor

foun

dati

ons.

3 co

ncre

te

sum

p ar

eas

crea

ted

in

Aux

ilia

ry B

ay G

roun

d F

loor

Was

tew

ater

Tre

atm

ent F

acil

itie

s ad

ded

to

sout

h en

d o

f Mai

n G

roun

d F

loor

PL

AN

T S

YS

TE

M

MA

lNS

TE

AM

MA

IN G

EN

ER

AT

OR

CO

ND

EN

SA

TE

&

FE

ED

WA

TE

R

CO

ND

EN

SA

TE

&

FE

ED

W A

TE

R (

cant

.)

LO

CA

TIO

N

MA

JOR

EQ

UIP

ME

NT

(us

uall

y in

seq

uenc

e o

f use

)

Aux

. M

ezza

nine

4

24"-

dia.

Mai

n S

team

Lin

es &

36"

-dia

. M

ain

Ste

am

Man

ifol

d [s

uspe

nded

at e

l. 47

.5'];

2 3

0"-d

ia.

Mai

n S

team

P

ipes

Ope

rati

ng F

loor

H

igh-

Pre

ssur

e T

urbi

ne (

TG

-l)

wit

h 2

turb

ine

mai

n st

op

trip

val

ves,

& 2

gov

erno

r co

ntro

l val

ves

on e

ach

mai

n

stop

trip

val

ve

Mai

n R

ehea

ter

Fl.

4 M

oist

ure

Sep

arat

or R

ehea

ters

(E

-28-

1A th

roug

h E

-28-

ID)

Ope

rati

ng F

loor

2

Low

-Pre

ssur

e T

urbi

nes

(TG

-IA

, T

G-I

B)

Mai

n M

ezza

nine

H

igh-

Pre

ssur

e T

urbi

ne S

team

Dum

p &

Val

ves

Mai

n R

ehea

ter

Fl.

Low

-Pre

ssur

e T

urbi

ne S

team

Du

mp

& V

alve

s

Ope

rati

ng F

loor

M

ain

Gen

erat

or

Mai

n G

roun

d F

loor

2

Mai

n C

onde

nser

s (E

-23-

1A,

E-2

3-1B

)

Mai

n G

roun

d F

loor

2

Con

dens

ate

Fee

dwat

er P

umps

(P

-35-

1A,

P-3

5-1B

) [s

outh

of c

onde

nser

s]

Mai

n M

ezza

nine

Mai

n M

ezza

nine

Mai

n M

ezza

nine

Mai

n M

ezza

nine

Mai

n M

ezza

nine

2 P

rim

ing

Air

Eje

ctor

s (E

J-2

-1A

llB

) &

2 M

ain

Air

E

ject

ors

(EJ-

I-A

, E

J-I-

B)

Gla

nd S

team

Con

dens

er (

E-6

4-1

A)

#6

A L

ow-P

ress

ure

Fee

dwat

er H

eate

r (E

-22-

1A)

[thr

ough

M

ain

Con

dens

er A

]

#6

B L

ow-P

ress

ure

Fee

dwat

er H

eate

r (E

-22-

1B)

[thr

ough

M

ain

Con

dens

er B

]

#5

A L

ow-P

ress

ure

Fee

dwat

er H

eate

r (E

-21-

1A)

[thr

ough

M

ain

Con

dens

er A

]

HADDA~l N

EC

K N

UC

LE

AR

PO

\tV

ER

PL

AN

T

(Con

nect

icut

Yan

kee

Nuc

lear

Pow

er P

lant

) H

AE

RN

o. C

T-1

85

(Pag

e 92

)

APP

EN

DIX

B -

SUM

MA

RY

OF

STR

UC

TU

RE

S, P

RIM

AR

Y F

UN

CT

ION

S, A

ND

MA

JOR

EQ

UIP

ME

NT

NA

ME

AN

D D

AT

ES

S

UM

MA

RY

DE

SC

RIP

TIO

N

TU

RB

INE

BU

ILD

ING

(con

t).

PL

AN

T S

YS

TE

M

AU

XIL

IAR

Y

FE

ED

WA

TE

R

SER

VIC

E W

AT

ER

LO

CA

TIO

N

Mai

n M

ezza

nine

Ope

ratin

g Fl

oor

Ope

ratin

g Fl

oor

MA

JOR

EQ

UIP

ME

NT

(us

uall

y in

seq

uenc

e o

f use

)

#5B

Low

-Pre

ssur

e Fe

edw

ater

Hea

ter

(E-2

l-lB

) [t

hrou

gh

Mai

n C

onde

nser

B]

#4

A&

4B

Low

-Pre

ssur

e Fe

edw

ater

Hea

ters

(E

-20-

1A,

E-

20-1

B)

[sou

thea

st a

nd s

outh

wes

t com

ers]

#3a

& 3

B L

ow-P

ress

ure

Feed

wat

er H

eate

rs (

E-1

9-1A

, E-

19-1

B)

[nor

th o

fLP

feed

wat

er h

eate

rs 4

A&

4B

]

Aux

. M

ezza

nine

#2

a &

2B

Low

-Pre

ssur

e Fe

edw

ater

Hea

ters

(E

-18-

1A,

E

l8-1

B)

Mai

n G

roun

d Fl

oor

Feed

wat

er H

eate

r D

rain

Rec

eive

r T

ank

(TK

-23-

1A)

[nor

th o

f Con

dens

er A

]

Mai

n G

roun

d Fl

oor

2 Fe

edw

ater

Hea

ter

Dra

in P

umps

(P

-33-

1All

B)

[nor

th o

f T

K-2

3-1A

]

Aux

. G

roun

d Fl

oor

2 St

eam

Gen

erat

or F

eedw

ater

Pum

ps (

P-3

l-lA

llB

)

Aux

. M

ezza

nine

#

1 A &

1 B

Hig

h-Pr

ress

ure

F eed

wat

er H

eate

rs (

E-1

7-lA

llB

) [f

ed c

omm

on h

eade

r to

4 St

eam

Gen

erat

or F

eed

Lin

es]

Aux

. M

ezza

nine

4

elec

. m

otor

-ope

rate

d va

lves

, fee

d re

gula

ting

val

ves

&

man

ual

isol

atio

n va

lves

per

Ste

am G

ener

ator

Fee

d L

ine

Aux

. M

ezza

nine

A

uxili

ary

Feed

Byp

ass

Val

ves

Mai

n G

roun

d Fl

oor

2 M

ain

Tur

bine

Lub

e O

il C

oole

rs (

E-6

0-1A

1lB

)

Mai

n G

roun

d Fl

oor

2 C

lose

d C

oolin

g W

ater

Sys

tem

Hea

t Exc

hang

ers

(E-7

0-1A

lIB

)

low

er M

ain

4 M

ain

Gen

erat

or E

xcite

r H

ydro

gen

Coo

lers

(E

-62-

1 A

G

ener

ator

cas

ing

thro

ugh

E-6

2-1

D)

HADDA~1 N

EC

K N

UC

LE

AR

PO

\VE

R P

LA

NT

(C

onne

ctic

ut Y

anke

e N

ucle

ar P

ower

Pla

nt)

HA

ER

No.

CT

-185

(P

age

93)

AP

PE

ND

IX B

-SU

MM

AR

Y O

F ST

RU

CT

UR

ES,

PR

IMA

RY

FU

NC

TIO

NS,

AN

D M

AJO

R E

QU

IPM

EN

T

NA

ME

AN

D D

AT

ES

S

UM

MA

RY

DE

SC

RIP

TIO

N

.TU

RB

INE

BU

ILD

ING

(con

t)

PL

AN

T S

YS

TE

M

SER

VIC

E W

AT

ER

CO

NT

RO

L A

IR

SER

VIC

E A

IR

WA

TE

R

TR

EA

TM

EN

T

LO

CA

TIO

N

MA

JOR

EQ

UIP

ME

NT

(us

uall

y in

seq

uenc

e o

f use

)

Mai

n G

roun

d Fl

oor

2 M

ain

Gen

erat

or S

eal O

il C

oole

rs(E

-61-

1 A

ll B

)

Mai

n M

ezza

nine

2

Mai

n G

ener

ator

Iso

late

d P

hase

Bus

Coo

lers

(E

-48-

1All

B)

Mai

n M

ezza

nine

M

ain

Exc

iter

Coo

ler

(E-1

15-l

A)

Aux

. G

roun

d Fl

oor

2 C

ontr

ol A

ir C

ompr

esso

rs (

C-3

-IA

lIB

)

Aux

. G

roun

d Fl

oor

2 C

ontr

ol A

ir R

ecei

vers

(T

K-2

9-IA

llB

)

Aux

. G

roun

d Fl

oor

4 C

ontr

ol A

ir F

ilter

s (F

L-8

-IA

llB

, FL

-9-I

AlI

B)

Aux

. G

roun

d Fl

oor

2 C

ontr

ol A

ir D

ehyd

rato

r (F

L-3

5-lA

lIB

)

Mai

n G

roun

d Fl

oor

Con

trol

Air

Com

pres

sor

(C-3

-IC

)

Mai

n G

roun

d Fl

oor

Con

trol

Air

Rec

eive

r (T

K-2

9-lC

)

Mai

n G

roun

d Fl

oor

2 C

ontr

ol A

ir F

ilter

s (F

L-l

0-1

All

B)

Mai

n G

roun

d Fl

oor

Con

trol

Air

Deh

ydra

tor

(FL

-35-

lC)

Aux

. G

roun

d Fl

oor

Serv

ice

Air

Int

ake

Filte

rs

Aux

. G

roun

d Fl

oor

2 Se

rvic

e A

ir C

ompr

esso

rs (

C-2

-lA

lIB

)

Aux

. G

roun

d Fl

oor

2 A

fter

cool

ers

& M

oist

ure

Sep

arat

ors

(F-4

9-1A

lIB

)

Aux

. G

roun

d Fl

oor

Serv

ice

Air

Rec

eive

r (T

K-2

8-IA

)

Aux

. G

roun

d Fl

oor

Wel

l W

ater

Fil

ter

(FL

-44-

IA)

Aux

. G

roun

d Fl

oor

Vac

uum

Dea

erat

or (

D-I

-IA

)

Aux

. G

roun

d Fl

oor

2 V

acuu

m D

eaer

ator

Pum

ps (

P-3

0-IA

llB

)

HA

DD

AIV

I N

EC

K N

UC

LE

AR

PO

'VE

R P

LA

NT

(C

onne

ctic

ut Y

anke

e N

ucle

ar P

ower

Pla

nt)

HA

ER

No.

CT

-185

(P

age

94)

AP

PE

ND

IX B

-S

UM

MA

RY

OF

ST

RU

CT

UR

ES

, PR

IMA

RY

FU

NC

TIO

NS

, AN

D M

AJO

R E

QU

IPM

EN

T

NA

ME

AN

D D

AT

ES

DIS

CH

AR

GE

CA

NA

L

1965

-196

6

12

R S

WIT

CH

Y A

RD

14B

SW

ITC

HY

AR

D

TE

RR

Y T

UR

BIN

E

BU

ILD

ING

& N

ON

-RE

TU

RN

V

AL

VE

ST

AT

ION

19

64-1

966

AU

XIL

IAR

Y F

EE

DW

AT

ER

P

UM

P S

KID

EN

CL

OS

UR

ES

A

&B

SU

MM

AR

Y D

ES

CR

IPT

ION

± 7

000'

fro

m T

urbi

ne B

uild

ing

disc

harg

e tu

nnel

to

Con

nect

icut

Riv

er;

upst

ream

16

2' i

s st

eel-

brac

ed,

tim

ber-

shee

t-pi

le

side

d fl

ume

endi

ng a

t ro

ck w

eir,

dro

ps

from

el.

-5.5

'to -1

0' in

to e

arth

en c

anal

65'

-80

' w

ide

at

botto

m,

130'

-160

'-wid

e at

ou

ter e

nds

of b

erm

s w

ith

grav

el o

r rip

-rap

in

side

slo

pes;

700

' wid

e at

rive

r

Ste

el-f

ram

ed

stru

ctur

e w

ith

Gal

best

os

sidi

ng,

40'

wid

e, 4

3.5'

hig

h &

ext

endi

ng

11'

wes

t o

f th

e R

eact

or C

onta

inm

ent:

G

roun

d F

loor

el

. 21

.5'

cont

aine

d 2

auxi

liar

y fe

ed w

ater

pum

ps.

Rem

aini

ng

elev

atio

ns (e

l. 3

1',

41

',4

9',

57'

) co

ntai

ned

stru

ctur

al s

teel

use

d to

sup

port

the

mai

n st

eam

and

fee

d w

ater

sys

tem

pip

ing

and

valv

es;

non-

retu

rn v

alve

sta

tion

at

uppe

r le

vel.

PL

AN

T S

YS

TE

M

SE

RV

ICE

WA

TE

R;

CIR

CU

LA

TIN

G

WA

TE

R

345

KV

34

5K

V

MA

IN S

TE

AM

MA

.1N

ST

EA

M

LO

CA

TIO

N

MA

JOR

EQ

UIP

ME

NT

(us

uall

y in

seq

uenc

e o

f use

)

Aux

. G

roun

d F

loor

2

Fon

vard

ing

Pum

ps (

P-3

5-!a

/!b)

Aux

. G

roun

d F

loor

D

emin

eral

ized

Wat

er F

ilte

r (F

L-4

5-1a

)

Ter

ry T

urbi

ne B

ldg

Non

-Ret

urn

Val

ve

Sta

tion

319

Mai

n T

rans

form

er, 3

09 R

eact

or C

oola

nt B

us

Tra

nsfo

rmer

, mai

n tr

ansf

orm

er s

econ

dary

sid

e di

scon

nect

s, m

ain

tran

sfor

mer

out

put m

otor

-ope

rate

d di

scon

nect

s, 3

20 L

inet

o 14

B S

wit

chya

rd

grou

nd d

isco

nnec

t, m

anua

l di

scon

nect

s, p

ower

cir

cuit

br

eake

rs, m

otor

-ope

rate

d di

scon

nect

s, b

lock

hous

e

2 A

uxil

iary

Ste

am p

ower

ed S

team

Gen

erat

or P

umps

(P

-32-

1A11

B)

Non

-Ret

urn

Val

ves

(2 e

ach

on 4

ste

am li

nes)

Ele

ctri

cal

Aux

ilia

ry S

team

Gen

erat

or F

eed

Pem

p (P

-32-

Ie)

HA

DD

Al\

1 N

EC

K N

UC

LE

AR

PO

'VE

R P

LA

NT

(C

onne

ctic

ut Y

anke

e N

ucle

ar P

ower

Pla

nt)

HA

ER

No.

CT

-185

(P

age

95)

APP

EN

DIX

B -

SUM

MA

RY

OF

STR

UC

TU

RE

S, P

RIM

AR

Y F

UN

CT

ION

S, A

ND

MA

JOR

EQ

UIP

ME

NT

NA

ME

AN

D D

AT

ES

SE

RV

ICE

BU

ILD

ING

19

64-1

966

1981

-198

4

PR

IMA

RY

AU

XIL

IAR

Y

BU

ILD

ING

19

64-1

966

SU

MM

AR

Y D

ES

CR

IPT

ION

Stee

l-fr

amed

, co

ncre

te-w

alle

d st

ruct

ure

wit

h G

albe

stos

sid

ing,

304

'x43

'-87.

5', 2

0'-

55' h

igh:

Gro

und

floo

r el

. 21

.5' i

nclu

des

I-st

ory,

20

'-hig

h w

areh

ouse

&

m

aint

enan

ce

shop

17

9.5'

x43'

&62

', &

lo

cker

roo

ms!

off

ices

124

.5'x

62'&

87.5

'; M

ezza

nine

el.

41.5

' 62'

x103

' ; O

pera

ting

F

loor

at e

l. 59

.5 8

5'x1

03',

over

lyin

g pa

rt

of T

urbi

ne B

ldg

Aux

. B

ay.

reno

vatio

ns

incl

udin

g ne

w

roof

an

d co

ncre

te

wal

ls

for

seis

mic

pr

otec

tion,

ra

ised

fl

oor

in

Con

trol

R

oom

, &

C

hem

istr

y L

ab

Rei

nfor

ced-

conc

rete

str

uctu

re ±

70xI

50',

32'

high

with

2. 1 '

-thi

ck w

alls

, co

ncre

te

floo

rs.

2 m

ain

leve

ls,

each

wit

h 4-

ton

mon

orai

l sy

stem

s:

grou

nd

floo

r at

el

. 21

.5',

seco

nd f

loor

at

el.

35.5

.' Se

ctio

ns

abov

e se

cond

floo

r ge

nera

lly s

teel

fram

ed

wit

h in

sula

ted

Gal

best

os s

idin

g. A

t ea

st

end,

R

esid

ual

Hea

t R

emov

al

Pit,

±70

x35.

5'

exte

nds

to

el.

-19.

' O

ther

se

ctio

ns!l

evel

s in

clud

e:

3-1e

vel

Bor

on

Rec

over

y R

oom

, ±

29.5

' sq

uare

at

el

evat

ions

. 35

.5-3

6.5'

, 25

.5'

&

15.5

'; H

igh-

Pres

sure

/Low

-Pre

ssur

e Sa

fety

In

ject

ion

Cub

icle

at

el

. 15

.5;

Blo

wdo

wni

Sam

ple

&

Non

-Rad

ioac

tive

V

alve

R

oom

, ±

15x

25'

at

el.

22';

Rad

ioac

tive

Val

ve R

oom

± 10

x19'

at

el.

25';

Cha

rgin

g &

Met

erin

g Pu

mp

Roo

ms

at

el.

15.5

'; Se

al W

ater

Fil

ter

Cub

icle

at e

l. 13

.4.'

Pipe

ga

lleri

es

belo

w

cent

ral

long

itudi

nal a

xis

at e

leva

tions

of 1

3-14

'

PL

AN

T S

YS

TE

M

PR

IMA

RY

W

AT

ER

SER

VIC

E W

AT

ER

LO

CA

TIO

N

Gro

und

Flo

or

Mez

zani

ne

Ope

rati

ng F

loor

Ope

rati

ng F

loor

Gro

und

Flo

or

(eas

t end

)

Gro

und

Flo

or

(eas

t end

)

Seco

nd F

l. (n

ear

W.

end)

thr

ough

floo

r

Bor

on R

ecov

ery

Roo

m (

uppe

r le

vel)

abov

e Se

cond

Flo

or,

el.

40'

Gro

und

Flo

or

MA

JOR

EQ

UIP

ME

NT

(us

uall

y in

seq

uenc

e o

f use

)

Cab

le S

prea

ding

Are

a

Switc

hgea

r R

oom

A

Con

trol

Roo

m

Proc

ess

Com

pute

r R

oom

2 Pr

imar

y W

ater

Tra

nsfe

r Pu

mps

(P

-29-

1A11

B)

2 R

ecyc

led

Pri

mar

y W

ater

Tra

nsfe

r Pu

mps

(P

-118

-1A

1IB

)

2 C

ompo

nent

Coo

ling

Hea

t Exc

hang

ers

(E-4

-1A

1lB

)

Bor

on R

ecov

ery

Ove

rhea

d C

onde

nser

(E-1

4-1

A)

Stea

m G

ener

ator

Blo

wdo

wn

Tan

k V

ent C

onde

nser

(E

-78-

1A)

2 St

eam

Gen

erat

or S

ampl

e C

hill

er C

onde

nser

s (C

-16

-1A

11

B)

Res

idua

l H

eat R

emov

al

2 R

esid

ual

Hea

t Rem

oval

Hea

t Exc

hang

ers

Pit

, eL

-19

' (E

-5-I

AlI

B)

HA

DD

Al\

1 N

EC

K N

UC

LE

AR

PO

'VE

R P

LA

NT

(C

onne

ctic

ut Y

anke

e N

ucle

ar P

ower

Pla

nt)

HA

ER

No.

CT

-185

(P

age

96)

AP

PE

ND

IX B

-S

UM

MA

RY

OF

ST

RU

CT

UR

ES

, PR

IMA

RY

FU

NC

TIO

NS

, AN

D M

AJO

R E

QU

IPM

EN

T

NA

ME

AN

D D

AT

ES

PR

IMA

RY

AU

XIL

IAR

Y

BU

ILD

ING

(con

t.)

SU

MM

AR

Y D

ES

CR

IPT

ION

P

LA

NT

SY

ST

EM

CO

MPO

NE

NT

C

OO

LIN

G W

AT

ER

CO

tv1P

ON

ENT

CO

OL

ING

WA

TE

R

BO

RO

N

RE

CO

VE

RY

LO

CA

TIO

N

MA

JOR

EQ

UIP

ME

NT

(us

uall

y in

seq

uenc

e o

f use

)

Sec

ond

Flo

or

2 A

dam

s F

ilte

rs (

FL

-53-

IAl1

B)

Bor

on R

ecov

ery

Bor

on R

ecov

ery

Dis

till

ate

Coo

ler

(E-1

5-1

A)

Roo

m (

low

er le

vel)

Seco

nd F

l., n

ear

top

of

Bor

ic A

cid

Mix

Tan

k V

ent C

onde

nser

(E

-78-

1 A

) B

oric

Aci

d M

ix T

ank

Prim

ary

Dra

ins

Tan

k C

ubic

le in

Res

idua

l H

eat R

emov

al P

it

Sec

ond

Flo

or

(nor

thw

est c

omer

)

Gro

und

Flo

or

(nor

thw

est c

omer

)

Gro

und

Flo

or

(nor

thw

est c

omer

)

Bor

on R

ecov

ery

Roo

m,

2nd le

vel

Pri

mar

y D

rain

s T

ank

Ven

t Con

dens

er (

E-I

I-1

A)

2000

-gal

. C

ompo

nent

Coo

ling

Sur

ge T

ank

(T

K-5

-1A

)

3 C

ompo

nent

Coo

ling

Wat

er p

umps

(P

-13-

1 A

lIB

l1 C

) [b

elow

Com

pone

nt C

ooli

ng S

urge

Tan

k]

2 C

ompo

nent

Coo

ling

Wat

er H

eat

Exc

hang

ers

(E-4

-1A

11B

) [b

elow

Com

pone

nt C

ooli

ng S

urge

Tan

k]

2 B

oron

Rec

over

y W

aste

Liq

uid

Tra

nsfe

r P

umps

(P

-22-

1AA

11B

)

Bor

on R

ecov

ery

Roo

m,

Bor

on R

ecov

ery

Dis

till

ate

Fee

d H

eat E

xcha

nger

(E

-12-

2nd le

vel

lA)

Bor

on R

ecov

ery

Roo

m,

Bor

on R

ecov

ery

Fir

st S

tage

Eva

pora

tor

Bot

tom

s P

ump

1st le

vel

(P-2

3-1A

)

Bor

on R

ecov

ery

Roo

m,

Bor

on R

ecov

ery

Fir

st S

tage

Eva

pora

tor

Boi

ler

(E-4

3-1

A)

1st le

vel

Bor

on R

ecov

ery

Roo

m,

Bor

on R

ecov

ery

Fir

st S

tage

Eva

pora

tor

(EV

-I-A

) 1st

le

vel

Bor

on R

ecov

ery

Roo

m,

Bor

on R

ecov

ery

Sec

ond

Sta

ge E

vapo

rato

r B

otto

ms

Pum

p 1st

le

vel

(P-2

5-1A

)

HADDA~1 N

EC

K N

UC

LE

AR

PO

\VE

R P

LA

NT

(C

onne

ctic

ut Y

anke

e N

ucle

ar P

ower

Pla

nt)

HA

ER

No.

CT

-185

(P

age

97)

AP

PE

ND

IX B

-S

UM

MA

RY

OF

ST

RU

CT

UR

ES

, PR

IMA

RY

FU

NC

TIO

NS

, AN

D M

AJO

R E

QU

IPM

EN

T

NA

ME

AN

D D

AT

ES

S

UM

MA

RY

DE

SC

RIP

TIO

N

PL

AN

T S

YS

TE

M

ST

EA

M

GE

NE

RA

TO

R

BL

OW

DO

WN

ST

EA

M

GE

NE

RA

TO

R

BL

OW

DO

WN

RE

SID

UA

L H

EA

T

RE

MO

VA

L

LO

CA

TIO

N

MA

JOR

EQ

UIP

ME

NT

(us

uall

y in

seq

uenc

e o

f use

)

Bor

on R

ecov

ery

Roo

m,

Bor

on R

ecov

ery

Sec

ond

Sta

ge E

vapo

rato

r B

oile

r (E

-44-

1 st

leve

l lA

)

Bor

on R

ecov

ery

Roo

m,

Bor

on R

ecov

ery

Sec

ond

Sta

ge E

vapo

rato

r (E

V-2

-A)

2nd le

vel

Bor

on R

ecov

ery

Roo

m,

Bor

on R

ecov

ery

Eva

pora

tor

Ove

rhea

d C

onde

nser

(E

-14-

3rd le

vel

1 A)

Bor

on R

ecov

ery

Roo

m,

Bor

on R

ecov

ery

Dis

till

ate

Acc

umul

ator

(T

K-1

8-1A

) 3rd

leve

l

Bor

on R

ecov

ery

Roo

m,

2 B

oron

Rec

over

y D

isti

llat

e P

umps

(P

-26-

1AlI

B)

2nd le

vel

Bor

on R

ecov

ery

Roo

m,

Bor

ic A

cid

Rec

over

y C

oole

r (E

-16-

1 A

) 1st

le

vel

Gro

und

Flo

or

Liq

uid

Was

te C

ontr

ol B

oard

thro

ugh

Sec

ond

Fl.

Bor

ic A

cid

Mix

Tan

k (T

K-2

-1A

)

Blo

wdo

wn

Roo

m

abov

e S

econ

d F

l.

Res

idua

l Hea

t R

emov

al P

it

Res

idua

l H

eat

Rem

oval

Pit

Res

idua

l H

eat

Rem

oval

Pit

Ste

am G

ener

ator

Blo

wof

fTan

k (T

K-2

2-1A

)

2 S

team

Gen

erat

or B

low

offT

ank

Con

dens

ers

(E-9

0-lA

llB

)

Blo

wof

fTan

k C

oole

r (E

-91)

[at

el.

-8.6

']

2 R

esid

ual

Hea

t R

emov

al P

umps

(P

-14-

1AlI

B)

[at e

l.-1

9']

2 R

esid

ual

Hea

t E

xcha

nger

s (E

-5-1

AlI

B)

[bet

wee

n el

. -5

' & -

19']

HA

DD

Al\-

1 N

EC

K N

UC

LE

AR

PO

\VE

R P

LA

NT

(C

onne

ctic

ut Y

anke

e N

ucle

ar P

ower

Pla

nt)

HA

ER

No.

CT

-185

(P

age

98)

APP

EN

DIX

B -

SUM

MA

RY

OF

STR

UC

TU

RE

S, P

RIM

AR

Y F

UN

CT

ION

S, A

ND

MA

JOR

EQ

UIP

ME

NT

NA

ME

AN

D D

AT

ES

PR

IMA

RY

AU

XIL

IAR

Y

BU

ILD

ING

(co

ni.)

SU

MM

AR

Y D

ES

CR

IPT

ION

P

LA

NT

SY

ST

EM

CH

EM

ICA

L &

V

OLU

1vIE

CO

NT

RO

L

CO

NT

AIN

ME

NT

P

UR

GE

EM

ER

GE

NC

Y

LO

CA

TIO

N

MA

JOR

EQ

UIP

ME

NT

(us

uall

y in

seq

uenc

e o

f use

)

Met

erin

g P

ump

Roo

m

Rea

ctor

Coo

lant

Let

dow

n: N

on-R

egen

erat

ive

Hea

t E

xcha

nger

(E

-76-

IA)

Sec

ond

Flo

or

Gro

und

Flo

or

thro

ugh

Sec

ond

Fl.

Gro

und

Flo

or

Gro

und

Flo

or

Gro

und

Flo

or

Gro

und

Flo

or

Cha

rgin

g P

ump

Roo

m

Met

erin

g P

ump

Roo

m

Sea

l Wat

er F

ilte

r C

ubic

le, e

l. 13

.4'

Sea

l W

ater

Fil

ter

Cub

icle

, el.

13.4

'

Res

idua

l H

eat

Rem

oval

Pit

Sec

ond

Flo

or

Sec

ond

Flo

or

Hig

h-P

ress

ure/

Low

P

ress

ure

Saf

ety

Rea

ctor

Coo

lant

Let

dow

n: V

olum

e C

ontr

ol T

ank

(T

K-6

-lA

)

Pur

ific

atio

n P

ump

(P-1

2-1A

)

Bor

ic A

cid

Mix

ing

Tan

k (

TK

-2-1

A)

2 B

oric

Aci

d P

umps

(P

-9-1

A1l

B)

Bor

ic A

cid

Fil

ter

(FL

-15-

1A)

Bor

ic A

cid

Ble

nder

(M

-9-1

A)

[ato

p B

oric

Aci

d F

ilte

r]

Bor

ic A

cid

Str

aine

r (S

T -6

-1 A

)

2 C

entr

ifug

al C

harg

ing

Pum

ps (

P-1

8-1

All

B)

Cha

rgin

g M

eter

ing

Pum

p (P

-II-

IA)

Rea

ctor

Coo

lant

Pum

p S

eal W

ater

Inj

ecti

on:

2 S

eal

Wat

er

Sup

ply

Fil

ters

(F

L-5

9-1A

lIB

)

Rea

ctor

Coo

lant

Pum

p S

eal

Wat

er I

njec

tion

: 2

Sea

l W

ater

R

etur

n F

ilte

rs (

FL

-36

-IA

llB

)

Rea

ctor

Coo

lant

Pum

p S

eal W

ater

Inj

ecti

on:

Sea

l Wat

er

Hea

t Exc

hang

er (

E-4

5-1A

)

Che

mic

al A

ddit

ion

Tan

k (

TK

-7-1

A)

2 P

ure

& D

ilut

ion

Air

Fan

s (F

-50

AlI

B)

2 L

ow-P

ress

ure

Saf

ety

Inje

ctio

n P

umps

(P

-92

-1A

11

B);

HA

DD

AM

NE

CK

NU

CL

EA

R P

O\V

ER

PL

AN

T

(Con

nect

icut

Yan

kee

Nuc

lear

Pow

er P

lant

) H

AE

RN

o. C

T-1

85

(Pag

e 99

)

AP

PE

ND

IX B

-SU

MM

AR

Y O

F ST

RU

CT

UR

ES,

PR

IMA

RY

FU

NC

TIO

NS,

AN

D M

AJO

R E

QU

IPM

EN

T

NA

ME

AN

D D

AT

ES

S

UM

MA

RY

DE

SC

RIP

TIO

N

PL

AN

T S

YS

TE

M

CO

RE

CO

OL

ING

CO

NT

AIN

ME

NT

S

PR

AY

LIQ

UID

WA

ST

E

WA

ST

E G

AS

LO

CA

TIO

N

Inj e

ctio

n C

ubic

le

Cha

rgin

g Pu

mp

Roo

m

MA

JOR

EQ

UIP

ME

NT

(us

uall

y in

seq

uen

ce o

f use

)

2 H

igh-

Pre

ssur

e Sa

fety

Inj

ecti

on P

umps

(P

-15-

1AlI

B)

2 C

entr

ifug

al C

harg

ing

Pum

ps (

P-1

8-lA

llB

) [s

hift

ed s

uctio

n fr

om

Vol

. C

ontr

ol T

ank

(TK

-6-l

A)

to R

efue

ling

Cav

ity W

ater

Sto

rage

T

ank

(TK

-4-l

A)

outs

ide

Prim

. A

ux.

Bld

g.]

Bor

on R

ecov

ery

Roo

m,

2 L

ow-P

ress

ure

Safe

ty I

njec

tion

Pum

ps (

P-9

2-lA

llB

) 1st

le

vel

Res

idua

l Hea

t Rem

oval

2

Res

idua

l H

eat R

emov

al P

umps

(P

-14-

1 A

ll B

) P

it

Res

idua

l Hea

t Rem

oval

2

Aer

ated

Dra

ins

Tan

ks (

TK

-12-

1All

B)

Pit

Res

idua

l H

eat R

emov

al

2 A

erat

ed D

rain

Tan

k Pu

mps

(P

-20-

IAlI

B)

Pit

Res

idua

l H

eat R

emov

al

2 A

erat

ed L

iqui

d W

aste

Str

aine

rs (

ST

-1-I

All

B)

Pit

Gro

und

Flo

or

2 W

aste

Tes

t Tan

k Pu

mps

(P

-28-

IAll

B)

Dru

mm

ing

Roo

m,

Che

mic

al N

ucle

ar P

roce

ssin

g Sk

id

Gro

und

Flo

or

Gro

und

Flo

or

Liq

uid

Was

te C

ontr

ol B

oard

pipe

gal

lery

bel

ow

Blo

wdo

wn

Roo

m

Prim

ary

Dra

ins

Tan

k C

ubic

le i

n R

esid

ual

Hea

t Rem

oval

Pit

Prim

ary

Dra

ins

Tan

k C

ubic

le i

n R

esid

ual

H<-

at R

emov

al P

it

Val

ve S

tem

Lea

koff

Coo

ler (

E-8

5)

Pri

mar

y D

rain

s C

olle

ctin

g T

ank

(TK

-II-

IA)

Pri

mar

y D

rain

s T

ank

Ven

t Con

dens

er (

E-I

I-IA

)

Res

idua

l H

eat R

emov

al

2 Pr

imar

y D

rain

s T

ank

Pum

ps (

P-1

9-1A

llB

) P

it

HA

DD

Al\

1 N

EC

K N

UC

LE

AR

PO

\VE

R P

LA

NT

(C

onne

ctic

ut Y

anke

e N

ucle

ar P

ower

Pla

nt)

HA

ER

No.

CT

-185

(P

age

100)

APP

EN

DIX

B -

SUM

MA

RY

OF

STR

UC

TU

RE

S, P

RIM

AR

Y F

UN

CT

ION

S, A

ND

MA

JOR

EQ

UIP

ME

NT

NA

ME

AN

D D

AT

ES

DIE

SE

L G

EN

ER

AT

OR

B

UIL

DIN

G

1964

-196

6

NE

W D

IES

EL

GE

NE

RA

TO

R

BU

ILD

ING

19

69-1

970

ION

EX

CH

AN

GE

AR

EA

19

64-1

967

SU

MM

AR

Y D

ES

CR

IPT

ION

I-st

ory

rein

forc

ed-c

oncr

ete

stru

ctur

e,

2s'x

32',

14'

high

fr

om

el.

21.5

', of

f so

uthw

est

com

er o

f P

rim

ary

Aux

ilia

ry

Bui

ldin

g

I st

ory

rein

forc

ed-c

oncr

ete/

conc

rete

blo

ck

stru

ctur

e, 9

1 'x

4s'&

30'

PL

AN

T S

YS

TE

M

SE

RV

ICE

WA

TE

R,

El'v

1ER

GE

NC

Y

GE

NE

RA

TIO

N

SE

RV

ICE

WA

TE

R,

El'v

1ER

GE

NC

Y

GE

NE

RA

TIO

N

2 ad

jace

nt re

info

rced

-co

ncre

te s

truc

ture

s:

BO

RO

N R

EC

OV

ER

Y

Ion

Exc

hang

e S

truc

ture

, 21

.8'x

73.s

', 18

' hi

ghfr

omel

. 14

.6',w

ithd

ecka

tel.

22.

5'

in f

ront

of i

on e

xcha

nger

& d

emin

eral

izer

fa

cili

ties

whi

ch a

re

arra

yed

wit

hin

Ion

WA

ST

E L

IQU

ID

Exc

hang

e S

truc

ture

in 1

7'-h

igh

cham

bers

; S

pent

Res

in S

tora

ge P

it, 1

3.3'

xI7.

3',

17'

high

fr

om

el.

22'

wit

h ch

ambe

r fo

r re

mov

able

lin

er,

over

pit

ext

endi

ng to

el.

CH

EM

ICA

L &

7.5'

; w

ithi

n pi

t, Io

n E

xcha

nge

Sum

p Pu

mp

VO

LU

ME

CO

NT

RO

L

(P-6

3-1A

)toe

l.

2.5'

.

SP

EN

T F

UE

L

CO

OL

ING

LO

CA

TIO

N

Gro

und

Flo

or

Ion

Exc

hang

e S

truc

t.

Ion

Exc

hang

e S

truc

t.

Ion

Exc

hang

e St

ruct

.

Ion

Exc

hang

e S

truc

t.

Ion

Exc

hang

e S

truc

t.

Ion

Exc

hang

e S

truc

t.

Ion

Exc

hang

e S

truc

t.

Ion

Exc

hang

e S

truc

t.

Ion

Exc

hang

e St

ruct

.

Ion

Exc

hang

e S

truc

t.

MA

JOR

EQ

UIP

ME

NT

(us

uall

y in

seq

uenc

e o

f use

)

Was

te G

as C

ontr

ol B

oard

3 D

iese

l Gen

erat

ors

2 D

iese

l Gen

erat

ors

(2A

12B

), c

ontr

ol c

abin

ets,

em

erge

ncy

buse

s

Was

te L

iqui

d Io

n E

xcha

nger

(1-6

-1 A

)

Was

te L

iqui

d T

rans

fer

Fil

ter

(FL

-13-

1A)

Aer

ated

Dra

ins

Dem

iner

aliz

er (

I-3-

1a)

Aer

ate

Dra

ins/

Spe

nt F

uel P

it F

ilte

r (F

L-3

-IA

lFL

-6s)

Rea

ctor

Coo

lant

Let

dow

n: 2

Pur

ific

atio

n D

emin

eral

izer

Io

n E

xcha

nger

s (I

-I-I

A1I

B)

Rea

ctor

Coo

lant

Let

dow

n: R

eact

or C

oola

nt P

re-f

ilte

r (F

L-s

-IA

)

Rea

ctor

Coo

lant

Let

dow

n: R

eact

or C

oola

nt P

ost-

Fil

ter

(FL

-II-

IA)

Pur

ific

atio

n: D

ebor

onat

ing

Ion

Exc

hang

er (1

-2-1

A)

Spe

nt F

uel P

ool I

on E

xcha

nger

(I-

I-IC

)

Aer

ate

Dra

ins/

Spe

nt F

uel P

it F

ilte

r (F

L-3

-1A

1FL

-6s)

HA

DD

Al\

1 N

EC

K N

UC

LE

AR

PO

WE

R P

LA

NT

(C

onne

ctic

ut Y

anke

e N

ucle

ar P

ower

Pla

nt)

HA

ER

No.

CT

-185

(P

age

101)

AP

PE

ND

IX B

-S

UM

MA

RY

OF

ST

RU

CT

UR

ES

, PR

IMA

RY

FU

NC

TIO

NS

, AN

D M

AJO

R E

QU

IPM

EN

T

NA

ME

AN

D D

AT

ES

ION

EX

CH

AN

GE

S

TR

UC

TU

RE

AD

DIT

ION

19

73-1

974

ION

EX

CH

AN

GE

S

TR

UC

TU

RE

AD

DIT

ION

(c

ont.)

SP

EN

T R

ES

IN S

TO

RA

GE

F

AC

ILIT

Y

1979

-198

0

NE

W &

SP

EN

T F

UE

L

BU

ILD

ING

19

64-1

966

SU

MM

AR

Y D

ES

CR

IPT

ION

P

LA

NT

SY

ST

EM

Rei

nfor

ced

conc

rete

ad

diti

on

to

Ion

WA

ST

E G

AS

Exc

hang

e St

ruct

ure,

21'

x30'

, lO

'-20'

high

fr

om e

l. 12

.8',

wit

h 4

17'-h

igh

cham

bers

W

AS

TE

LIQ

UID

Par

t o

f or

igin

al S

pent

Res

in S

tora

ge P

it

rem

oved

, re

plac

ed

to

nort

heas

t w

ith

22 .2

'x29

. 2' l

ead-

lined

rein

forc

ed-c

oncr

ete

stru

ctur

e w

ith

fl. e

l. 19

.2',

10.8

' hig

h w

ith

23.5

'-hig

h w

alls

on

nort

h &

ea

st s

ides

, co

ntai

ning

11

5.2'

-dia

., 10

.8'-h

igh

stee

l ce

lls

48'x

1l7.

3' w

ith g

roun

d fl

oor

el.

21.5

'. S

pent

Fue

l Pi

t 48

'x49

', w

ith

top

of33

.5'

deep

, 35

'x37

' st

eel-

line

d re

info

rced

co

ncre

te p

it a

t el

. 47

' &

st

eel-

fram

ed,

Gal

best

os-s

ided

st

ruct

ure

abov

e to

el

. 75

.5'.

Pit

has

116

8 5'

-dia

., l4

'-hi

gh fu

el

cask

s &

a s

kim

mer

sys

tem

at t

op o

f poo

l; 6-

ton

brid

ge c

rane

abo

ve p

ool a

t el.

67.3

'. N

ew F

uel

Bui

ldin

g 48

'x38

.2',

54'

high

ab

ove

grou

nd

floo

r w

ith

rein

forc

ed

conc

rete

to

floo

r le

vel

at e

l. 47

', st

eel

fram

e &

Gal

best

os s

idin

g ab

ove;

gro

und

floo

r se

rves

as

Sp

.::nt

F

uel

Pit

pu

mp

cubi

cle;

flo

or a

t el

. 35

' su

ppor

ts P

VC

ra

cks

for

114

1 '-d

ia.,

I2'-h

igh

fuel

as

sem

blie

s; 3

-ton

bri

dge

cran

e at

el.

67.3

'.

WA

ST

E L

IQU

ID

SO

LID

WA

ST

E

BO

RO

N

RE

CO

VE

RY

SO

LID

WA

ST

E

WA

ST

E L

IQU

ID

SP

EN

T F

UE

L

CO

OL

ING

SE

RV

ICE

WA

TE

R

LO

CA

TIO

N

Ion

Exc

hang

e A

dd.

Ion

Exc

hang

e A

dd.

Ion

Exc

hang

e A

dd.

Spe

nt R

esin

Sto

rage

Pi

t

Ion

Exc

hang

e A

dd.

Ion

Exc

hang

e A

dd.

Spe

nt F

uel P

it p

ump

cubi

cle

Spe

nt F

uel P

it p

ump

cubi

cle

Spe

nt F

uel P

it p

ump

cubi

cle

Spe

nt F

uel P

it p

ump

cubi

cle

Spe

nt F

uel P

it pu

mp

cubi

cle

MA

JOR

EQ

UIP

ME

NT

(us

uall

y in

seq

uenc

e o

f use

)

Deg

asif

ier P

re-F

ilte

r (F

L-6

7)

Was

te L

iqui

d Po

lishi

ng D

emin

eral

izer

(1-9

)

Was

te E

vapo

rato

r Dis

till

ate

Fil

ter

(FI-

69)

Spen

t Res

in S

tora

ge T

ank

(TK

-l02

-lA

) S

pent

Res

in T

rans

fer

Pum

p (P

-156

-lA

)

Bor

on E

vapo

rato

r D

isti

llat

e F

ilte

r (F

L-6

8)

Bor

on R

ecov

ery

Pol

ishi

ng D

emin

eral

izer

(I-

8-la

)

Rem

ovab

le s

tora

ge li

ners

in

stee

l cel

ls;

3" P

VC

pip

e dr

ains

fro

m c

ell b

otto

ms

into

adj

acen

t Spe

nt

Res

in S

tora

ge P

it

2 S

pent

Fue

l C

ooli

ng P

umps

(P

-2l-

lAll

B)

1 Sp

ent F

uel P

ool T

ube-

Typ

e H

eat E

xcha

nger

(E

-lO

-lA

) 1

Spe

nt F

uel

Pool

Pla

te-T

ype

Hea

t Exc

hang

er (

E-l

O-l

B)

2 S

pent

Fue

l Po

ol S

kim

mer

Pum

ps (

P-9

0-lA

lIB

)

2 S

pent

Fue

l Poo

l Ski

mm

er F

ilter

s (F

L-6

5-1

AlI

B)

2 S

pent

Fue

l Po

ol H

eat

Exc

hang

ers

(E-i

O-l

AlI

B)

HADDA~1 N

EC

K N

UC

LE

AR

PO

'VE

R P

LA

NT

(C

onne

ctic

ut Y

anke

e N

ucle

ar P

ower

Pla

nt)

HA

ER

No.

CT

-185

(P

age

102)

AP

PE

ND

IX B

-S

UM

MA

RY

OF

ST

RU

CT

UR

ES

, PR

IMA

RY

FU

NC

TIO

NS

, AN

D M

AJO

R E

QU

IPM

EN

T

NA

ME

AN

D D

AT

ES

WA

ST

E D

ISP

OS

AL

B

UIL

DIN

G

1973

-197

4

WA

ST

E D

ISP

OS

AL

B

UIL

DIN

G (c

ont.)

SU

MM

AR

Y D

ES

CR

IPT

ION

Rei

nfor

ced-

conc

rete

R

adia

tion

Con

trol

A

rea

48'x

30',

25.5

'hig

h w

ith3

-ton

bri

dge

cran

e at

e1.

39.5

'

41 'x

42' r

einf

orce

d co

ncre

te s

truc

ture

, 55.

5'

high

wit

h fl

oors

at e

l. 0'

, 18

.5' &

21.

5', &

35

.5'

PL

AN

T S

YS

TE

M

RE

FU

EL

ING

WA

STE

GA

S

WA

STE

GA

S (c

ont.)

WA

STE

LIQ

UID

LO

CA

TIO

N

Spen

t Fue

l Pi

t el.

49.5

'

thro

ugh

mid

dle

leve

l

thro

ugh

uppe

r le

vel

mid

dle

leve

l

mid

dle

leve

l -el

. 30

'

uppe

r le

vel -

el.

50'

low

er le

vel

uppe

r le

vel

-el

. 38

.3'

low

er le

vel

mid

dle

leve

l -el

. 30

'

MA

JOR

EQ

UIP

ME

NT

(us

uall

y in

seq

uenc

e o

f use

)

Fuel

Ele

vato

r (F

U-5

); S

luic

e G

ate

(FU

-7)

Deg

asif

ier

Preh

eate

r (E

-86)

Deg

asif

ier

(TK

-58-

lA)

with

Deg

asif

ier

Ven

t Coo

ler

(E-

89)

& D

egas

ifie

r V

ent C

onde

nser

(E

-87)

2 D

egas

ifie

r L

iqui

d T

rans

fer P

umps

(P

-l0

6-l

All

B)

Deg

asif

ier

Eff

luen

t Coo

ler

(E-8

8)

Deg

asif

ier

Ven

t Coo

ler

(E-8

9)

Was

te G

as S

urge

Tan

k (T

K-5

9-lA

)

2 W

aste

Gas

Com

pres

sors

(C

-13-

lAll

B)

3 W

aste

Gas

Dec

ay T

anks

(T

K-6

0-lA

llB

/lC

)

Was

te G

as S

ampl

e &

Rel

ease

Hea

der

Was

te E

vapo

rato

r Fe

ed D

istil

late

Exc

hang

er (

E-9

6)

low

er le

vel

Was

te E

vapo

rato

r R

eboi

ler

Pum

p (P

-114

-1 A

)

low

er le

vel

Was

te E

vapo

rato

r R

eboi

ler

(E-9

2)

mid

dle/

uppe

r le

vels

W

aste

Liq

uid

Eva

pora

tor

(EV

-4)

mid

dle

leve

l W

aste

Eva

pora

tor

Dis

tilla

te T

ank

Pum

p (P

-115

-1A

)

uppe

r le

vel-

el.

47.3

' W

aste

Eva

pora

tor

Ove

rhea

d C

onde

nser

(E

-93)

uppe

r le

vel

Was

te E

vapo

rato

r D

istil

late

Tan

k (T

K-6

4)

HA

DD

AN

[ N

EC

K N

UC

LE

AR

PO

\VE

R P

LA

NT

(C

onne

ctic

ut Y

anke

e N

ucle

ar P

ower

Pla

nt)

HA

ER

No.

CT

-185

(P

age

103)

AP

PE

ND

IX B

-S

UM

MA

RY

OF

ST

RU

CT

UR

ES

, PR

IMA

RY

FU

NC

TIO

NS

, AN

D M

AJO

R E

QU

IPM

EN

T

NA

ME

AN

D D

AT

ES

WA

ST

E D

ISP

OS

AL

B

UIL

DIN

G (c

ont.)

PR

IMA

RY

WA

TE

R

ST

OR

AG

E T

AN

K

(TK

-20-

1A)

c196

4-19

66

RE

CY

CL

E P

RIM

AR

Y

WA

TE

R S

TO

RA

GE

TA

NK

SU

MM

AR

Y D

ES

CR

IPT

ION

30'-O

D,

150,

000-

gal.

stee

l tan

k

22'-O

D 1

50,O

OO

-gal

. st

eel t

ank

PL

AN

T S

YS

TE

M

CO

MP

ON

EN

T

CO

OL

ING

WA

TE

R

CO

MP

ON

EN

T

CO

OL

ING

WA

TE

R

LO

CA

TIO

N

el.

30.'

fupp

er l

evel

low

er le

vel

low

er le

vel

low

er le

vel

low

er le

vel

uppe

r le

vel

low

er le

vel

low

er le

vel

thro

ugh

uppe

r le

vel

uppe

r le

vel

-el

. 38

.3'

MA

JOR

EQ

UIP

ME

NT

(us

uall

y in

seq

uenc

e o

f use

)

Was

te E

vapo

rato

r D

isti

llat

e C

oole

r (E

-94)

Was

te E

vapo

rato

r B

otto

ms

Pum

p (P

-116

-1A

)

Was

te E

vapo

rato

r B

otto

ms

Coo

ler

Pre

heat

er (

E-9

7)

Was

te E

vapo

rato

r B

otto

ms

Coo

ler

(E-9

5)

Was

te E

vapo

rato

r B

otto

ms

Coo

ler

Cir

cula

ting

Pum

p (P

-12

0-1

A)

Floo

r &

Equ

ipm

ent D

rain

Tan

k (T

K-6

5)

2 Fl

oor

Dra

in T

ank

Pum

ps (

P-1

19-1

Al1

B)

2 E

quip

men

t Dra

in T

ank

Pum

ps (

P-1

21-1

All

B)

Deg

asif

ier

(TK

-58-

1A)

wit

h D

egas

ifie

r V

ent C

oole

r (E

-89

) &

Deg

asif

ier

Ven

t Con

dens

er (

E-8

7)

2 W

aste

Gas

Com

pres

sors

(C

-13-

1 A

ll B

)

mid

dle/

uppe

r le

vels

W

aste

Liq

uid

Eva

pora

tor

(EV

-4)

SE

RV

ICE

WA

TE

R

mid

dle/

uppe

r lev

els

Was

te L

iqui

d E

vapo

rato

r (E

V-4

)

PR

IMA

RY

W

AT

ER

PR

IMA

RY

W

AT

ER

thro

ugh

uppe

r le

vel

Deg

asif

ier

(TK

-5 8

-1 A

) w

ith

Deg

asif

ier

Ven

t Coo

ler

(E-

89)

& D

egas

ifie

r V

ent C

onde

nser

(E

-87)

east

of N

ew &

Spe

nt

Fue

l Bld

g.

& y

ard

cran

e

east

of R

adw

aste

R

educ

tion

Fac

ilit

y

HA

DD

Al\1

NE

CK

NU

CL

EA

R P

O\V

ER

PL

AN

T

(Con

nect

icut

Yan

kee

Nuc

lear

Pow

er P

lant

) H

AE

RN

o. C

T-1

85

(Pag

e 10

4)

APP

EN

DIX

B -

SUM

MA

RY

OF

STR

UC

TU

RE

S, P

RIM

AR

Y F

UN

CT

ION

S, A

ND

MA

JOR

EQ

UIP

ME

NT

NA

ME

AN

D D

AT

ES

(TK

-62-

1A)

1973

SU

MM

AR

Y D

ES

CR

IPT

ION

DE

MIN

ER

AL

IZE

D W

AT

ER

25

'-ID

lO

O,O

OO

-gaL

ste

el ta

nk

ST

OR

AG

E T

AN

K

(TK

-25-

1A)

c196

4-19

66

CO

ND

EN

SA

TE

ST

OR

AG

E

25'-I

D 1

00,O

OO

-gal

. ste

el ta

nk

TA

NK

(TK

-25-

lB)

c199

2

2 R

EC

YC

LE

TE

ST

TA

NK

S

13.6

'-OD

, 33'

-hig

h, 1

6,00

0-ga

L

stee

l (T

K-6

3-1A

1lB

) c1

973

tank

s

2 B

OR

ON

WA

ST

E

26'-O

D 7

5,O

OO

-gaL

st

eel t

anks

S

TO

RA

GE

TA

NK

S (

TK

-14-

lAlI

B)

c196

4-19

66

AE

RA

TE

D D

RA

INS

26

.3'-h

igh,

24'

-ID

, 99,

280-

gaL

st

eel

HO

LD

UP

TA

NK

(TK

-61-

IA)

tank

c1

973

2 W

AS

TE

WA

TE

R

TR

EA

TM

EN

T T

AN

KS

(T

K-1

7-1A

1l1B

) c1

964-

1966

RE

FU

EL

ING

CA

VIT

Y

WA

TE

R S

TO

RA

GE

TA

NK

(T

K-4

-1A

) c1

964-

1966

RA

DW

AS

TE

RE

DU

CT

ION

F

AC

ILIT

Y

SW

ITC

HG

EA

R B

UIL

DIN

G

1987

-199

0

14'-O

D,1

6,O

OO

-gal

. st

eel t

anks

37'-O

D,

50S

-hig

h, 2

50,O

OO

-gaL

st

eel

tank

I-st

ory,

met

al-f

ram

ed,

met

al-s

ided

st

ruct

ure,

± 8

8'x6

1 '&

37'

pile

-sup

port

ed,

rein

forc

ed-c

oncr

ete

stru

ctur

e 34

'x64

', 37

.5'h

igh

abov

e gr

ound

fl.

el.

21.5

';

PL

AN

T S

YS

TE

M

AU

XIL

IAR

Y

FE

ED

WA

TE

R

AU

XIL

IAR

Y

FE

ED

WA

TE

R

BO

RO

N

RE

CO

VE

RY

BO

RO

N

RE

CO

VE

RY

LIQ

UID

WA

STE

LIQ

UID

WA

STE

1. C

HE

MIC

AL

&

VO

L.

CO

NT

RO

L:

PU

RIF

ICA

TIO

N

2. E

NlE

RG

EN

CY

C

OR

E C

OO

LIN

G

3. C

ON

TA

INN

lEN

T

SP

RA

Y

4. R

EF

UE

LIN

G

SO

LID

WA

ST

E

LO

CA

TIO

N

in d

iked

enc

losu

re

SW o

f Rea

ctor

C

onta

inm

ent

SW o

f Rea

ctor

C

onta

inm

ent

conc

rete

-dik

ed

encl

osur

e N

E o

f Pr

imar

y A

ux.

Bld

g

conc

rete

-dik

ed

encl

osur

e ea

st o

f P

rim

ary

Aux

. B

ldg

MA

JOR

EQ

UIP

ME

NT

(us

uall

y in

seq

uenc

e o

f use

)

conc

rete

-dik

ed

2 W

aste

Eva

pora

tor

Feed

Pum

ps (

P-I

13-1

AlI

B)

encl

osur

e N

E o

f P

rim

ary

Aux

. B

ldg

nort

heas

t of R

eact

or

Con

tain

men

t

nort

heas

t of R

eact

or

Con

tain

men

t

sout

h o

f Spe

nt F

uel

com

pact

ors,

cle

anin

g fa

cilit

y, s

tora

ge a

rea

for

solid

B

uild

ing

radi

oact

ive

and

mix

ed w

aste

nort

h o

f Ser

vice

B

uild

ing

inst

rum

enta

tion

and

cont

rol f

or s

afe

plan

t shu

tdow

n

HA

DD

Al\

1 N

EC

K N

UC

LE

AR

PO

\VE

R P

LA

NT

(C

onne

ctic

ut Y

anke

e N

ucle

ar P

ower

Pla

nt)

HA

ER

No.

CT

-185

(P

age

105)

AP

PE

ND

IX B

-S

UM

MA

RY

OF

ST

RU

CT

UR

ES

, PR

IMA

RY

FU

NC

TIO

NS

, AN

D M

AJO

R E

QU

IPM

EN

T

NA

ME

AN

D D

AT

ES

115

KV

SW

ITC

HY

AR

D

1964

-196

6

SE

RV

ICE

BO

ILE

R R

OO

M

1964

-196

6

AD

MIN

IST

RA

TIO

N

BU

ILD

ING

19

64-1

966

c198

0-19

85

INF

OR

MA

TIO

N C

EN

TE

R

1964

-196

7 19

77 r

enov

atio

ns

HE

AL

TH

PH

YS

ICS

F

AC

ILIT

Y 1

964-

1966

HE

AL

TH

PH

YS

ICS

P

RO

JEC

T T

RA

ILE

R

HE

AL

TH

PH

YS

ICS

CO

UN

T

MO

DU

LE

WA

RE

HO

US

E #

11

97

6

SU

MM

AR

Y D

ES

CR

IPT

ION

base

men

t lau

ndry

are

a el

. 13

.5';

air

hand

ling

eq

uipm

ent

at e

l. 27

.5;

swit

chge

ar e

quip

men

t at

el.

41.5

'

I-st

ory

24'-h

igh,

45'

x36.

5' s

teel

-fra

med

, co

ncre

te-w

alle

d st

ruct

ure

with

Gal

best

os

sidi

ng;

2 se

rvic

e do

ors

2-st

ory,

3

0S

-hig

h

bric

k &

co

ncre

te-b

lock

st

ruct

ure

100'

x24'

(lst

fl.)

&

36'

(2nd

fl

), w

ith

45'x

15'

rear

sec

tion.

N

,S,&

W f

acad

es

have

tr

ipar

tite

fa

cade

s o

f gl

azed

br

ick,

fl

uted

en

amel

ed

alum

inum

si

ding

, &

en

amel

ed

alum

inum

sash

; W

. si

de h

as 2

I2'

-hig

h, 3

0'-w

ide

arch

ed p

anel

s w

ith

glas

s an

d gl

azed

bri

ck in

fill

&

doo

rs i

n no

rth

arch

ed p

anel

I-st

ory

±30

'-squ

are

Rec

ords

Bui

ldin

g ad

ded

in

sim

ilar

st

yle

to

sout

h en

d,

wit

h w

est

face

ex

tend

ed n

orth

pas

t sou

th a

rche

d pa

nel

I-st

ory

conc

rete

-blo

ck s

truc

ture

I2

0'x3

8'-5

0', e

xter

ior s

tone

bas

e an

d m

etal

pan

els

2-st

ory

conc

rete

str

uctu

re,

±35

'x51

.5',

±3

0S

hig

h

±20

'xl7

'

2-st

ory

±25

'x19

' tem

pora

ry

I-st

ory

stee

l-fr

amed

, m

etal

-sid

ed,

gabl

e-ro

of

stru

ctur

e I2

0'x1

22.S

', 20

' hig

h

PL

AN

T S

YS

TE

M

115

KV

HE

AT

ING

LO

CA

TIO

N

MA

JOR

EQ

UIP

ME

NT

(us

uall

y in

seq

uenc

e o

f use

)

sout

heas

t of R

eact

or

115-

kv o

il ci

rcui

t bre

aker

, 2

tran

sfon

ners

wit

h lig

htni

ng

Con

tain

men

t ar

rest

ors,

man

uall

y-an

d D

C-m

otor

-ope

rate

d di

scon

nect

sw

itche

s

sout

hwes

t cor

ner

of

2 oi

l-fi

red

boile

rs

Serv

ice

Bui

ldin

g

wes

t of T

urbi

ne

Bui

ldin

g

wes

t of m

ain

plan

t

coun

t roo

m,

proc

essi

ng, o

ffic

e

offic

e

coun

t roo

m

stor

age

HA

DD

A.l\

-1 N

EC

K N

UC

LE

AR

PO

WE

R P

LA

NT

(C

onne

ctic

ut Y

anke

e N

ucle

ar P

ower

Pla

nt)

HA

ER

No.

CT

-185

(P

age

106)

APP

EN

DIX

B -

SUM

MA

RY

OF

STR

UC

TU

RE

S, P

RIM

AR

Y F

UN

CT

ION

S, A

ND

MA

JOR

EQ

UIP

ME

NT

NA

ME

AN

D D

AT

ES

WA

RE

HO

US

E #

2 19

94

ST

EA

M G

EN

ER

AT

OR

M

OC

K-U

P B

UIL

DIN

G 1

982

EN

GIN

EE

RIN

G O

FF

ICE

/ M

OD

UL

E c

1985

INS

TR

UM

EN

TA

TIO

N &

C

ON

TR

OL

OP

ER

AT

ION

S

BU

ILD

ING

c19

85

UN

CO

ND

ITIO

NA

L

RE

LE

AS

E F

AC

ILIT

Y

TR

AIN

ING

& S

TO

RE

S

OF

FIC

E c

1978

-197

9

EM

ER

GE

NC

Y

OP

ER

AT

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WRITTEN HISTORICAL AND DESCRIPTIVE DATA

HAER CT-185HAER CT-185

ADDENDUM TO:HADDAM NECK NUCLEAR POWER PLANT(Connecticut Yankee Nuclear Power Plant)362 Injun Hollow RoadHaddamMiddlesex CountyConnecticut

HISTORIC AMERICAN ENGINEERING RECORDNational Park Service

U.S. Department of the Interior1849 C Street NW

Washington, DC 20240-0001

ADDENDUM TO HADDAM NECK NUCLEAR POWER PLANT

(Connecticut Yankee Nuclear Power Plant) HAER No. CT-185

(Page 107)

HISTORIC AMERICAN ENGINEERING RECORD


This report is an addendum to a 106-page report previously transmitted to the Library of Congress in 2010.

Location: 362 Injun Hollow Road Haddam Middlesex County Connecticut

U.S. Geological Survey Deep River Quadrangle UTM Coordinates 18.708837.80 E - 4595337.91 N +41o 28' 56.866" latitude, -72o 29' 54.983" longitude i

Dates of Construction: 1964-1966

Engineers: Westinghouse Electric Company Stone & Webster Engineering Corporation

Present Owners: Connecticut Yankee Atomic Power Company (CYAPCO) 362 Injun Hollow Road Haddam Neck CT 06424-3022

Present Use: Demolished with some foundations left in place.

Significance: The Haddam Neck Nuclear Power Plant was one of the earliest commercial-scale nuclear power stations in the United States and the first completed on the East Coast. During its operating history from 1967 to 1996, this plant established several records in electrical production. The plant was eligible for the National Register of Historic Places.

Project Information: CYAPCO ceased electrical generation at the Haddam Neck plant in 1996. Decommissioning operations started in 1998, subject to authority of the Nuclear Regulatory Commission (NRC). NRC authority brought the project under the purview of federal acts and regulation protecting significant cultural resources from adverse project effects.ii This documentation was requested by the Connecticut State Historic Preservation Office to preclude the possibility of any adverse project effects.

i The Haddam Neck Nuclear Power Plant was located at latitude +41o 28' 56.866"; longitude -72o 29' 54.983". The coordinate represents the center point of the former reactor containment building. This coordinate was obtained on November 4, 2009 using a GPS unit accurate to +/- 5 meters. The coordinates were compared to values obtained on the Google Earth website and USGS Deep River Quadrangle and the accuracy of the coordinates is +/- 15 meters..

iiNational Historic Preservation Act of 1966 (PL 89-655), the National Environmental Policy Act of 1969 (PL 91-190), the Archaeological and Historical Preservation Act (PL 93-291), Executive Order 11593, Procedures for the Protection of Historic and Cultural Properties (36 CFR Part 800).



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Note: The material in this report is in large part based on Connecticut Yankee Atomic Power Company records that are archived at the University of Connecticut, Dodd Library. The records consist of plant design drawings, plant historical records, employee newsletters, environmental reports, regulatory correspondence, scrapbooks, plaques, historic photographs, and other audiovisual materials. The records are available to the public. For information contact the librarian at: University of Connecticut Thomas J. Dodd Research Center 405 Babbidge Road, Unit 1205 Storrs, Connecticut 06269-1205 860.486.4500 / 860.486.4521 (Fax) http://doddcenter.edu Project Manager and Historian Michael S. Raber Raber Associates 81 Dayton Road, P.O. Box 46 South Glastonbury, CT 06073 860/633-9026

Nuclear, Steam and Electric Power Historian Gerald Weinstein 40 West 77th Street, Apt 17B New York, NY 10024 212/431-6100

Industrial Archaeologist Robert C. Stewart Historical Technologies 1230 Copper Hill Road West Suffield, CT 06093 860/668-2928

Nuclear Power Consultant Graphics Consultant Connecticut Yankee Liaison Gerald van Noordennen Gerald Loftus John Arnold Dutchman Consulting 5 Roberts Road 362 Injun Hollow Road 30 Miller Road Marlborough, CT 06447 Haddam Neck, CT 06424 Burlington, CT 06013

Connecticut Yankee Operations Specialist Shea Hemingway 362 Injun Hollow Road Haddam Neck, CT 06424



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Letter designations of Haddam Neck Power Plant structures and facilities. Letter designations for structures A through W were assigned when the project started. Research and detailed knowledge of the plant acquired during the recordation process indicated that some facilities could logically be combined with others. In addition, some ancillary facilities were described in the overview of CT-185, or briefly in Appendix A. The following shows which letter designated facilities were combined or included in another section. CT-185-E Terry Turbine Building is included in CT-185-C Turbine Building CT-185-K Ion Exchange Area is described in Appendix A - Summary of Structures, Primary Functions and

Major Equipment and included in CT-185-L Waste Disposal Building CT-185-M Spent Resin Storage Facility Area is described in Appendix A - Summary of Structures, Primary Functions and Major Equipment and included in CT-185-L Waste Disposal Building CT-185-N Radwaste Reduction Facility is included in CT-185-L Waste Disposal Building CT-185-Q Service Boiler Room is included in CT-185-F Service Building CT-185-S Information Center is described in Appendix A - Summary of Structures, Primary

Functions and Major Equipment CT-185-T Health Physics Facility is described in Appendix A - Summary of Structures, Primary Functions

and Major Equipment CT-185-U Health Physics Count Module is described in Appendix A - Summary of Structures, Primary

Functions and Major Equipment CT-185-V Steam Generator Mock-up is described in Appendix A - Summary of Structures, Primary

Functions and Major Equipment CT-185-W Emergency Operations Facility is described in Appendix A - Summary of Structures, Primary

Functions and Major Equipment. A reference to the facility will also be found in the CT-185 Overview.



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Historical Information

The Haddam Neck Nuclear Power Plant had an important place in the history of American nuclear power generation as one of the first two large commercial plants in the United States, and the first completed on the East Coast.1 The plant centered on a pressurized water reactor (PWR). In this design water, kept under pressure to prevent boiling, removed heat from the chain reaction in the reactor core. It also moderated or slowed neutrons from the uranium "fuel" to energies at which the fission process was initiated and could continue. Reaction control was provided by control rods which could be inserted or removed from the core to absorb excess neutrons. Boron added to the water also controlled the reaction. The cooling water passing through the core became radioactive and was cycled through a heat exchanger, called a steam generator. Water passing through a secondary system in the steam generator absorbed heat from the pressurized system. It turned to steam which drove the turbine and generator, was condensed, then cycled back to the steam generator. During its operating history from 1967 to 1996, this plant established several records in electrical production. It typified the steam powered electric plants which combined a primary nuclear steam supply system based on two decades of post-World-War-II development, and secondary systems based on 19th and early 20th century technology for converting steam to electrical energy. The design of the plant, and issues arising from that design, are significant examples of the limitations inherent in the first generation of American nuclear plants.

The Nuclear Furnace: Principals and Construction

In a nuclear power station, the traditional role of a firebox making heat from a chemical reaction between either wood or a fossil fuel and oxygen is taken over by a nuclear fuel core producing heat from a nuclear-fission chain reaction that is initiated, maintained and controlled to give an even power output. The nuclear age can be dated to 1895, when Wilhelm Roentgen discovered that a cathode ray tube invented by William Crookes in England emitted rays that passed though solids.2 The existence of nuclear energy from Uranium was recognized soon after through the work of Antoine-Henri Becquerel and the team of Pierre and Marie Curie in his lab. The Curies discovered Radium, the heat of which while low, continued for thousands of years and hinted at the power in a nucleus.3 Early in the 20th century, Earnest Rutherford in England and Niels Bohr in Denmark constructed a model of the atom with a central nucleus surrounded by electrons and posited their respective electrical charges.4 Parallel with these concepts was Max Planck’s 1900 quantum theory that fundamentally reordered the way scientists viewed the nature of energy. Creation of new elements by particle bombardment (artificial transmutation) was demonstrated by Rutherford in 19195 followed by the actual disintegration of a nucleus with accelerated particles in the early 1930s.6 The construction of atom smashers in the 1930s confirmed Einstein’s 1905 theory of relativity dealing with the conversion of mass into energy and ultimately explained the immense power produced by splitting the atom.7

The discovery of the neutron by James Chadwick in 1932 was central to the development of nuclear power since it explained isotopes or variations of elements also known as nuclides and gave scientists a good a good missile to project at nuclei.8 Unlike the proton, the other basic component of matter, the neutron carries no electrical charge enabling it to penetrate nuclei easily.9 Enrico Fermi in Italy showed that bombarding uranium with neutrons produced new radioactive isotopes when nuclei captured a neutron, and he believed that new elements above uranium in the periodic table had been produced. More importantly, Fermi showed that neutron capture occurred more frequently if the beamed neutrons were slowed by passing them through a



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hydrogen rich moderator such as water or paraffin.10 However, it was Frederic Joliot, Irene Curie and Pascal Savitch in France, and Otto Hahn and Fritz Strassman (with the help of Lise Meitner) in Germanyiii who were determining via chemical separation that in fact the nucleus under bombardment had absorbed an extra neutron and then broken up or split into two fragments of unequal mass and two or more neutrons.11 Meitner and her nephew Otto Frisch, basing their research on Bohr’s model of the atom, called the process fission12 and its discovery led to worldwide research into the fissioning of other elements. Most significant for military and civilian programs was the fact that the fission fragments split apart at high speed and produced prodigious amounts of energy. That energy could be calculated in theoretical terms in millions of electron volts (Mev.) and was largely kinetic from fission and neutron fragments augmented by gamma ray production and fragment decay.13iv The chemical energy of fossil fuel heat production involving the outer electronic particles of atoms was on the order of a few electron volts while nuclear reactions caused by a splitting of the neutrons and protons in the nucleus of an atom produced millions of electron volts.14 This disparity was demonstrated by Einstein as mass and energy being interchangeable and related in the equation E = mc², where one gram of any matter was the equivalent of 25 million kilowatt hours of energy.15 The immensely smaller change in mass from burning coal vs. the large loss of mass from the atoms participating in the fission process explained why fissioning one pound of uranium was equal to that produced by burning 1,000 tons of coal or 6,000 barrels of oil.16 The chemical proofs of fissioning were duplicated by physicists in the early months of 1939 and showed that a portion of the energy release was in the form of heat and that more neutrons were produced.17 Leo Szilard conceptualized a nuclear chain reaction while standing on a London street corner in 1933 and received a British patent for the idea. Later Enrico Fermi (at Columbia University) and Eugene Wigner also confirmed that a sustainable divergent chain reaction could be made to occur.18 That reaction would produce two or three more neutrons per fissioned nucleus, with at least one of these neutrons colliding with and splitting another nucleus with the same result,19 and could be a controlled event or an uncontrolled event such as the Atomic Bomb.20 Shortly after the discovery of fission, further work revealed the existence of delayed neutrons which were to be essential in the control of nuclear reactors.21

As a reliable, controllable heat source, a reactor had to establish a balance between neutron production, absorption and losses.22 The delayed neutrons enabled mechanical devices to control the fission process. At the same time, researchers were comparing the fission cross sections (probability23) of energetic fast neutrons verses slowed thermal neutrons. Fermi predicted that slowed neutrons would be required to make a chain reaction in natural uranium possible.24 The large cross section of thermal neutron fission, coupled with confirmation of Bohr’s theory that it was produced by the uranium isotope U235 (discovered by Arthur Dempster at University of Chicago in 193525) led to the intense effort to isolate enough of the isotope to study and ultimately the creation of a huge industry to produce enriched uranium. As Fermi and Szilard had theorized, the first chain-reacting pile at Chicago started in late 1942 could only achieve a chain reaction with

iii Chemical evidence of fission was discovered by Hahn and Strassman in December 1938 and announced in early 1939. In May 1968 the State and City of New York declared Nuclear Week in honor of the 30th anniversary of the discovery. Connecticut Yankee Atomic Power Company was one of the industry patrons for events held at the Union Carbide building (Booth 1969).

iv Unstable (radioactive) nuclei disintegrated or decayed over a fixed period of time, measured as the time required for ½ of the active material to present to decay. While theoretically infinite, ten half lives rendered the activity negligible. Radon gas with a half life of 382 days would be largely decayed in 30 days (Oxford English Dictionary 1989: 4: 322; Stephenson 1954: 23).



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theorized, the first chain-reacting pile at Chicago started in late 1942 could only achieve a chain reaction with natural uranium U235 if the fuel in the form of lumps was embedded in a lattice of carbon moderator blocks.26 Until its construction the work on fission had been that of theoretical chemistry and physicsv but the successful controlled reaction there and the subsequent build up for the Manhattan project brought the concepts into the realm of nuclear engineering.

The Development of American Nuclear Steam Supply Systems, c1945-1960

Nuclear Submarines and the Beginnings of Commercial Nuclear Power

Perhaps the first electricity produced by a nuclear chain reaction occurred when one of the World-War-II-era reactors at Oak Ridge National Laboratory in Tennessee was fitted in the early postwar years with a boiler supplying steam to a small external turbine generator which lit up a light bulb.27 Most sources credit the Experimental Breeder Reactor (EBR-1), built by the Argonne National Laboratory at the National Reactor Testing Station (NRTS) in Idaho, with being the first nuclear reactor to produce electrical power in December 1951.28 Less than five years, later one of the NRTS reactors was used to provide all electric power for 1200 people in Arco, Idaho.29

It was, however, Cold War military planning for submarine propulsion which ultimately drove civilian American power reactor development.30 The idea of a submarine that could travel at high speed underwater was proven by the German Type XXI U-Boat of World War II. The XXI boats had super battery plants and streamlined hulls designed to give greater submerged speed.31 Taking that a step further, the German navy came up with a design that eliminated the air-breathing diesel-charged battery plants with their limited endurance. The Walter closed-cycle propulsion system in the Type XXVI submarines utilized hydrogen peroxide as an oxidizer to make steam which powered a turbine to drive the propeller, providing extended underwater operation. Both designs were evaluated by the U.S. and British navies after the war, but the peroxide system was unreliable and hazardous.32 The end of the war and the arrival of nuclear power doomed that technology.

Less than six months after the close of the war, nuclear physicists were suggesting using nuclear power to drive ships.33 In March 1946, the Naval Research Laboratory recommended that nuclear submarines be constructed on high speed hulls based on the German XXI and XXII types.34 Captain (later Rear Admiral) Hyman Rickover, who had observed operations at Oak Ridge since 1946, took a lead role in the development of the nuclear submarine and American nuclear power plants.35 As head of the Naval Reactors Branch (Division of Reactor Development) he started working with the General Electric Company (GE) to develop a submarine reactor plant.36 GE scientists were involved with the isolation of Uranium-235 before the war, and the company took over management of the plutonium production operations at the U.S. Atomic Energy Commission’s (AEC) Hanford Laboratories in Richland, Washington in 1946.37 GE was an advocate of a Submarine Intermediate Reactor (SIR) that utilized relatively fast neutron velocity and was cooled by liquid metal sodium. Concerned about the viability of that technology, Rickover asked the Westinghouse Electric

vEven at the dawn of nuclear power, scientists ability to calculate with precision nuclear events was remarkable. In 1939 John Dunning who was working at isolating U235 at Columbia University accurately predicted that an atomic bomb yield would equal 20,000 tons of TNT. (Laurence 1969: 12.)



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Corporation to develop an alternative design. Westinghouse had also been involved at an early stage of nuclear development as a supplier to the Manhattan Project.38 They favored a reactor cooled with ordinary (light) water with slowed velocity (thermal type) neutrons. The genesis of both companies' designs was probably the CP-3 Heavy-Watervi reactor built at Lemont, I by Argonne in 1944 which had a core of uranium rods submerged in a water tank cooled by a heat exchanger system.39 GE’s commitments to production of weapons materials at Hanford and difficulties with the sodium system led Rickover to push Westinghouse to provide the first working plant.40 Their Submarine Thermal Reactor (STR Mark I) propulsion system was designed at the Westinghouse Bettis Laboratory near Pittsburgh, and tested in a mock-up submarine hull at the NRTS.41 The developed power plant (STR Mark II) utilized a Pressurized Water Reactor (PWR) powered by highly-enriched U-235. The fuel in metallic form was clad with zirconium alloy forming tubes which were arranged in bundles. The nuclear chain reaction occurring in the fuel rods produced heat, and was controlled by hafnium neutron absorber rods inserted into the bundles. The coolant that took the heat out of the reactor core and moderated the reactions was ordinary (light) water, pressurized to prevent boiling as it was pumped back and forth between the reactor and tubing in separate external heat exchanger vessels called steam generators. Feed water pumped around the heated tubes turned to steam which powered turbines geared to the propellers (Figure 2).

The PWR (also known as a closed-cycle water reactor)42 utilizing highly-enriched uranium was a good choice because its compact size allowed it to fit in the confines of a submarine hull. The light water was easy to handle, and the relatively low fluid and steam pressures outside of the reactor made the plant very reliable, a requirement for a vessel designed for undersea warfare. In comparison to other reactor types under development, the PWR was not very efficient,43 but it did not have to be commercially profitable in a naval vessel and was far better than the diesel-electric and peroxide drives that preceded it. Fears of the Soviet Union fielding a large fleet of captured German advanced U-Boats44 spurred intense, well-organized, and rapid design development among Rickover, Westinghouse and the Electric Boat Division of General Dynamics at Groton, Connecticut. President Truman laid the keel of the first nuclear submarine, USS Nautilus in 1952. The shakedown cruises in 1955 set records for underwater distances and the success of its plant was a tremendous impetus for use of that type at sea. Westinghouse subsequently supplied the reactors for the first U.S. Navy surface vessels and many of the second-generation submarines. Despite Westinghouse’s success, Rickover still wanted another system as an alternative, and GE continued to work on their design. Its Submarine Intermediate Reactor (SIR)vii was tested on land at Knolls Atomic Power Laboratory in Milton, New York before installation in USS Seawolf, launched in 1956 as the second nuclear submarine. Surplus power from that plant near Schenectady was distributed locally and may have been the first commercial electricity to be produced by nuclear energy.45 The Seawolf was commissioned in 1957, but the SIR proved unreliable and was replaced by 1960 with a conventional Westinghouse PWR.46 GE then switched to water-cooled reactors with its twin high-pressure Submarine Advanced Reactors for the USS Triton.47 Their later-model reactors powered a majority of the navy’s aircraft carriers, cruisers and submarines.

viHeavy-water (D2O), discovered in the 1930s, has the hydrogen atoms replaced with deuterium. It makes a good moderator because it does not capture or waste as many neutrons as light water allowing the use of fuel without enrichment (Oxford English Dictionary 1989: v.4, p. 559 (Deuterium); Nero & Dennis 1984:391).

viiThe Nautilus reactor utilized thermal neutrons of reduced energy. The SIR used neutrons of intermediate energy. Breeder reactors that create more fuel use high energy fast neutrons (Weinberg and Wigner 1958: 12).



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Atoms for Peace and Early Commercial Reactor Designs

The 1946 Atomic Energy Act encouraged civilian uses of nuclear power without specifying the means. President Eisenhower’s Energy for Peace program and the 1954 Atomic Energy Act opened the way for military research and fissionable materials to reach civilian programs.48 In August of 1955 the International Conference on the Peaceful Uses of Atomic Energy opened in Geneva. Papers were presented on three broad categories of reactors being considered by seven nations: water-cooled, gas-cooled, and liquid-metal-cooled. Water-cooled reactors were divided into pressurized and boiling types.49 Types of coolants were light water vs. heavy water in the water reactors, air vs. carbon dioxide in the gas reactors and sodium vs. bismuth in the liquid metal reactors. Moderators which slowed neutrons and aided the reaction included light water and heavy water. Carbon in the form of graphite was also proposed as a moderator in all but the boiling water types. The classes were further divided by types of nuclear fuel (natural vs. enriched)viii and by fuel configuration (heterogeneous vs. homogeneous).ix

Heterogeneous loading with uranium dioxide pellets stacked in stainless steel or zirconium alloy (Zircaloy) tubes was the most common arrangement.50 Most of the reactors were thermal types that had neutrons slowed to thermal velocities.51 An additional type, the Fast Fission Breeder that had no moderator and produced additional fuel, was also described.52

Considering all the variables of cooling types, moderators, fuels, and fuel configurations there were at least 100 feasible arrangements.53 The United States reported on about ten reactor types completed or in construction under the Atomic Energy Commission (AEC) prototype program at the national laboratories or

viiiCivilian reactors using light water require uranium fuel in which the natural percentage of U-235 (about 0.7 percent) has been slightly enriched to typical concentrations of 2-5 percent, less than the concentrations needed for naval submarine reactors. American civilian reactors have relied on government-owned gaseous diffusion enrichment plants; centrifuge-enrichment plants have been built in Europe, Japan, and South Africa. Reactors using heavy water, notably the Canadian Candu models discussed below, can use natural uranium without enrichment (McIntyre 1975; Power 1982b).

ixIn the homogeneous fuel arrangement, the uranium or other fuel was suspended in liquid or formed into a slurry which could be pumped in and out of the reactor and replaced at will without shutting down the reactor. The design was adaptable to fuel breeding and recycling was an integral part of the process. The AEC decided not to pursue the concept, or its successor the molten salt breeder and instead advanced the fast breeder reactor to be built at Clinch River. (Weinberg 1994: 117-129).



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production plants at Argonne, IL, Brookhaven, NY, Hanford, WA, Oak Ridge, TN, and the NRTS to assess the various characteristics. A movable package reactor for the Army and an aircraft power plant for the Air Force were also being developed. GE and Westinghouse worked closely with some of these facilities and the results of their experimentation would profoundly affect American reactor economics for years.54 Containment building criteria developed in parallel with reactor types. The dangers from the release of fission materials were well recognized and each reactor design seemed to favor one type of construction over another. While remote siting had originally been the main safety feature, the Seawolf test plant near a population center was housed in one of the first vapor-tight steel containment shells.55 Clearly, Cold War military competition drove American choices for power station reactors. The AEC wanted nuclear power to advance and the naval program offered the fastest and best chance for that to happen. Private industry may have been enticed by financial assistance from the commission’s Power Demonstration Reactor Program (PDRP),56 and the exponentially greater power of fission: the energy of one pound of fissioned U-235 was equal to the energy in 1000 tons of high quality coal.57 The government’s willingness to provide enriched fuel at nominal prices from its gaseous diffusion and centrifuge plants (built during the war to provide uranium and plutonium for bombs) allowed American designers to push the relatively less-efficient light-water reactor that required enriched fuel as the ideal power producer. Based on queries sent by the AEC to private industry in 1951, plans for four private, commercial electric-power generating plants were announced at the 1955 conference.58 The stations, all with light-water reactors were at Shippingport, Pennsylvania; Dresden, Illinois; Rowe, Massachusetts (Yankee); and Buchanan, New York (Indian Point). American and British engineers rightly noted that light-water reactors were going to be limited in their steam pressure and temperature abilities. The parameters of the light-water coolant and moderator were limited by two factors: the difficulty of making large reactor shells pressure proof, and the requirement that the water temperature around the hottest fuel rod areas be kept low enough to prevent film boiling which could cause inadequate cooling.59 Manufacturing limitations resulted in a maximum working pressure of about 2,000 pounds per square inch (psi) corresponding to a temperature of about 636 degrees Fahrenheit (F). A lower temperature had to be maintained around the fuel rods, resulting in the reactor producing steam at a temperature well below that in contemporary fossil-fueled boilers. It was expected that the poor quality steam would cause moisture problems with steam turbines. In addition, a technology available in fossil-fueled boilers, adding heat to the steam after generation (superheating) was not possible with the water reactors. In some cases plant designers added oil or gas-fired superheaters to raise the steam temperature.60 The relative inefficiency of the water reactors had another downside that would cause problems later: they produced large amounts of waste heat that had to be dumped into bodies of water unless the utilities opted for large and expensive cooling towers.61 The engineers were concerned about the safety of PWRs because the large quantities of pressurized water in the reactor coolant loops increased the potential damage from accidents.62 The gas-cooled system favored by the British had its own share of safety concerns, however. Without plants to enrich uranium cheaply, or the immediate need for naval reactors, Britain chose graphite-moderated carbon-dioxide-cooled reactors fueled with natural uranium, and moved faster than the Americans in opening Calder Hall, the world’s first commercial nuclear power station, in late 1956 at 65 megawatts (mw).63 France also started off with gas-cooled reactors but later went over to PWRs. Canada was an early nuclear power advocate, based on wartime research work at the Chalk River Nuclear Laboratories in Ontario. Assigned to work on natural-fuel, graphite-moderated, heavy-water cooled reactors, the Canadians stayed with that technology in their Nuclear Power Demonstration reactors. This technology led to the successful Candu



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(Canadian deuterium-uranium) type which uses heavy-water for both the moderator and coolant.64 The USSR inaugurated their power program with graphite-moderated water cooled PWRs.

Steam generators were a critical component of PWRs. Unlike the water-filled steam generating tubes of fossil fueled boilers which were directly impinged by the hot gases of combustion, steam generators were fluid heat exchangers. Heated water pumped out of the reactor was forced through steam generator tubes (primary side) without boiling and then returned to cool the reactor. Feed water was continuously pumped around the tubes and was heated to boiling by contact with their surfaces (secondary side). The generated steam was passed through moisture removal devices and sent to the turbines. Because of the relatively low temperature of the reactor coolant in a PWR, it was necessary to have a large heat-transfer surface to insure reasonable efficiency, calling for almost 4,000 tubes in the Westinghouse units and even more in those of other manufacturers. The cooling water had to flow evenly through all tubes and the feed water around them to insure full heat transfer, requiring complex perforated tube plates and baffles to channel the flow and insure that the primary and secondary water never merged. The tube bundles had to be supported to resist the flows on both sides with a network of braces and connections. The designs proved vulnerable to damage by various foreign substances.

Shippingport

In its prototype reactor program, the AEC supported in whole or in part the construction of small experimental reactor plants that included gas, polyphenol or sodium cooling, fast breeding, homogeneous fuels, etc.65 It was no coincidence, however, that the first large nuclear electric power utility station in the United States was a PWR built by Westinghouse and supervised by Rickover, a team with a proven track record.66 That plant, in Shippingport, was essentially a land-based version of a projected naval aircraft carrier reactor and went on line in late 1957. A group of manufacturers got together to build this pioneer plant. Westinghouse was the designer and main supervising contractor of the primary (reactor) systems and fabricated the intricate core assembly consisting of almost 100,000 fuel elements and the critical reactor coolant pumps.67 Three established fossil-fuel boiler manufacturers supplied other hardware. The reactor vessel was built by Combustion Engineering Inc. (CE). Foster-Wheeler Corporation (F-W) supplied two straight- tube steam generators, and The Babcock and Wilcox Company (B&W) provided two u-tube generators, each of which was part of a coolant loop out of the reactor.68 Stone and Webster were the architect-engineers with construction shared by Dravo Corporation and Burns and Roe, Inc. Duquesne Light Company supplied the secondary systems (turbine generator, condenser and auxiliaries) and guaranteed to purchase a block of 60 mw power.

The economic efficiency was not expected to compete with conventional plants, even with the AEC supplying most of the development dollars and the enriched fuel.69 Rickover’s standards served as a model for the industry,70 including 3000-hour core life, redundant fuel rods, backup safety systems, four separate coolant loops to ensure reliability, corrosion resistant materials in the reactor (but not in the secondary systems71), and commercially available equipment.72 Placing safety in the forefront, he insisted that the reactor be partially buried below grade so that a safety injection system could immerse the core with cooling water without needing pumps.73 If one of the control rods failed, water containing boron (boric acid) to kill the nuclear reactions could be injected into the system. 74 The reactor was contained in its own gas-tight steel chamber, with pairs of steam generators, coolant pumps, and auxiliaries in separate steel chambers, the whole



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surrounded by more than 5 feet of concrete shielding. The steel shells were designed to resist internal missiles such as valves traveling at high speed.75 The compact Nautilus reactor was made possible by the use of expensive highly-enriched uranium. The larger reactor for Shippingport had a more economical arrangement with highly-enriched uranium-zircaloy strips welded in plates and clad with Zircaloy-2 surrounded by a blanket of natural uranium dioxide cylindrical pellets in Zircaloy-2 tubes.76 The removable head of the reactor was penetrated for instrumentation and multiple fail safe control rods to start, maintain, and stop the fission reaction. Fuel elements could be individually replaced through fuel ports with the head in place. The lower section of the reactor vessel containing the core was surrounded by three feet of water to reflect neutrons. Coolant water entered the bottom of the vessel, flowed up through the fuel rod bundles taking off their heat and out through nozzles above the core to the steam generators. Each reactor coolant pump was a “canned” leak-proof unit developed for the submarine reactors, without seals between the centrifugal pump and motor and cooled by the primary water. Pressure in the system was closely controlled by electric heaters or water sprays in a separate pressurizer vessel. Probably following the lead of the Experimental Boiling Water Reactor at Argonne, the spent fuel rod bundles were unloaded underwater (to contain the radiation) by remote handling devices and moved out of containment in a flooded canal to a storage pool in an adjacent fuel handling building which also handled the new fuel.77 Shippingport was very much a product of national priorities, Rickover’s driving leadership, and perhaps a desire to beat Britain and the USSR.78 The forced nature of the engineering and construction, with speed of design a significant factor, impacted a whole generation of U.S. power reactors.

General Electric’s Boiling Water Reactors While Westinghouse was partnering with the AEC on land based PWR development, GE was working to perfect a boiling water reactor (BWR) to produce electric power from concepts tried out in experimental plants at Argonne and NRTS.79 In their single cycle BWR, cooling water was allowed to boil in the reactor dome producing steam that was sent directly to the turbine. Starting early with private funding, GE built the Vallecitos boiling water plant in California (AEC license #1) which sent out a small block of power over the Pacific Gas and Electric Company (PG&E) system in 1958.80 The benefits of reduced pressure in the core (compared to PWRs) and the elimination of what were to become troublesome steam generators were offset by the fact that the power regulation was poor and that irradiated steam traveled out of containment into the turbine, complicating environmental safeguards and turbine maintenance.81 Despite these problems, the design had development potential, proved to be equally efficient per kilowatt-hour (kw-h), and became the second most common type in the United States. The honeymoon that occurred during the development of Shippingport between Westinghouse, B&W, and CE did not last as their public relations departments touted the benefits of the reactors and components each was putting on the market. With GE, Alco Products Inc., AMF Atomics, The Martin Company, Allis-Chalmers Mfg. Company, North American Aviation (Atomics International), General Nuclear Engineering Corporation, and General Atomic in the mix, there probably were too many suppliers promoting too many design variations.82 By the early 1960s, power companies were likely attracted to the relatively lower costs of light-water designs, and to the more common models for which supplier profits and experience seemed to assure better potential customer support.



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Initial Licensing The AEC licensing program, established under the Atomic Energy Act of 1954, was critical to the siting, design, construction, start-up and power production of commercial plants. The Act gave the government extensive control over research and nuclear materials, but encouraged private industry to build plants.83 Fuel and rods were leased from the AEC at rates that were designed to make nuclear power stations competitive with fossil-fueled plants. The AEC exercised its control through extensive licensing procedures. Provisional Operating Licenses, Full Term Licenses, amendments, modifications, safety evaluations, and violations or penalties were the ruling documents. The commission specified everything from the facility location to the limits of worker radiation exposure. Changes to equipment that could in any way lead to radiation releases had to be approved. Resident inspectors were constantly in the field checking start up/shutdown procedures, re-fueling and maintenance. The utilities had to respond to the documents with detailed descriptions of actions taken and also send in annual reports. Commission authority over operational issues was accorded by the Code of Federal Regulations (CFR) and license amendments were printed in the Federal Register.84

Early Commercial Plants All of the plants built into the early 1960s including Shippingport were heavily supported by the government, but some built under the PDRP were mostly financed and built by private utilities with the AEC providing research, development assistance and free fuel for five years.85 The first of this series, commenced by Commonwealth Edison Company of Chicago at Dresden, Illinois, was GE’s first large 180 mw BWR plant. The dual-cycle design utilizing secondary steam generators to supply the low-pressure turbine had better power regulation than the earlier models.86 Westinghouse began its first private venture with the Yankee Atomic Power Company plant in Rowe, Massachusetts on the Sherman Pond reservoir of the Deerfield River, the first plant built by a consortium of New England power companies including Connecticut Light & Power Company and other later affiliates of Northeast Utilities. Completed in 1960 as the third American nuclear power plant and the first in New England, the PWR plant with a net output of 167 mw set the trend for subsequent Westinghouse three- and four-loop plants including Connecticut Yankee. Consolidated Edison partnered with B&W to build the 163-mw PWR Indian Point plant on the Hudson River in Buchanan, NY which was designed to use uranium and thorium as fuels.87 These stations were followed in 1962-3 by the 65- and 67-mw BWR’s at Big Rock Point of Consumers Public Power Company in Michigan and PG&E’s Humboldt Bay in California. In 1966 two more stations were completed: Northern States Power’s 60-mw Pathfinder BWR plant in Sioux Falls, South Dakota, and Philadelphia Electric Company’s 40-mw Peach Bottom High Temperature Gas Cooled Reactor plant. The last plant completed before Connecticut Yankee was the 48-mw BWR La Cross Nuclear Generating Station built in 1967 by the Dairyland Power Company of Wisconsin. The outputs of most of these demonstration phase plants were generally less than the fossil-fueled stations already on their respective grids.88 One other plant usually not included in the history of commercial plants was the N Reactor constructed at Hanford by the government in 1963. While producing weapons grade plutonium it also put a large block of electric power onto the Washington Public Power Supply System grid.89



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The Yankee Nuclear Plant at Rowe was the direct precursor to Westinghouse’s later plants including Connecticut Yankee. While based on the technologies worked out at Shippingport, the design basis was different enough so that project engineers stated in the company magazine Westinghouse Engineer that no direct comparisons could be made.90 Among the changes at Rowe were: use of a single, above-ground steel containment sphere; modification of reactor coolant flow with entry and exit nozzles above the core to facilitate the

admission of emergency core cooling water; stainless steel cladding on fuel rods instead of Zircaloy; uniform slightly-enriched fuel loading instead of the seed and blanket arrangement; silver-indium-cadmium control rods (instead of hafnium) with supports extending below the core; no ability to load fuel with the reactor head in place. Boron injection into the coolant aided in normal shutdowns and was also used in the safety injection system (later known as the emergency core cooling system.) Westinghouse-designed vertical U-tube steam generators were used in place of the contractor-built horizontal straight- and u-tube types at Shippingport. The elevated position of the reactor required an inclined water filled chute in which a transfer car carried the spent or new fuel rod bundles and control rods to and from the fuel building on ground level.91 Overall, the goals were safety (the location near the Vermont border was considered remote92) and low first cost to make the plant economically viable.93 Construction was supervised by Stone and Webster.

Containment Structures The containment structures of these early commercial stations began to assume the features that were standardized in the 1970s.94 The predominant shapes were either spheres or cylinders with hemispherical tops/bottoms. Dresden had a steel sphere modeled on the earlier Milton shell.95 Yankee Rowe containment was an above ground spherical steel vapor container surrounding a reinforced-concrete reactor support structure.96 Indian Point had a domed concrete cylinder with a separate internal steel vapor containment sphere.97 These early containment structures were built under local building codes and American Society of Mechanical Engineers (ASME) pressure vessel codes.98 While containing radiation was relatively easy, they also had to resist a pressure build-up from a release of the stored energy from the coolants and moderators, requiring a pressure rating of around 20 to 30 pounds per square inch gagex (psig).99 Radiation release to the atmosphere was to be restricted but not necessarily prohibited.100

xSteam or gas pressure was stated as pounds per square inch gage (psig) which was the pressure over the nominal atmospheric pressure at sea level of 14.7 pounds per square inch (psi). Pressure over true 0 was known as pounds per square inch absolute (psia).



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The AEC’s first standardized requirements for siting distance and emissions control were not proposed until 1961.101 As a result of objections to aspects of that draft, actual criteria of the 1962 document loosened the definition of population center distance, and led to a trend of reliance on engineered safeguards for protection rather than remote locations.102 It suggested that meteorological conditions be considered, and that no reactor be located within a quarter mile of a surface fault. Local codes and equivalent earthquake areas were to be the guides. In fact, wording on all environmental factors of seismology, meteorology, geology, and hydrology in the CFR was limited to four paragraphs.103 Planning engineers could use conventional methodology for assessing seismic and wind loading.104 Assumptions were based on experience with non-mechanical-system-filled structures such as offices.105 Thus the earliest commercial plants were often sited in areas that later would probably have been prohibited. The 1959 Santa Susana Station was sited by the Southern California Edison Company sixteen miles from the San Gabriel fault in an area that was described by project engineers to be “...as free from seismic disturbances as any in the vicinity of Los Angeles.”106 The siting of the Indian Point plant near the Ramapo fault was another example. The first detailed criteria from the AEC occurred well after the commercial phase plants were built.107 It was also some years before effective models were devised to show how critical nuclear components would interact with structures during earthquakes. By 1970 prestressed concrete (previously instituted in French nuclear plants) had taken over from simple reinforced construction.108

Early Insurance Issues

While prevention of release via engineered containment buildings and safety systems was generally accepted by the industry, there was developing resistance to insurance and siting criteria in place during the demonstration phase of reactor construction. As early as the 1860s, private insurers in partnership with boiler makers had arrived at specifications and inspections procedures to protect the public from power boiler explosions.109 The insurance industry was understandably uncertain about extending fossil-fuel-powered boiler insurance programs to nuclear reactors.110 To spread the risk they set up insurance pools (syndicates) and instituted rating plans to assess various nuclear perils111 As a result of a 1957 AEC report noting the possibility of up to four billion dollars in costs and thousands of fatalities from a major accident, most public liability was transferred from reactor manufacturers and operators to taxpayers though an act of Congress.112 The substitution of engineered safeguards for remote siting led to challenges to the construction permit for the Fermi Station, located within 35 miles of Detroit and Toledo. A federal court of appeals’ decision to rescind the AEC’s construction permit ended up in the U.S. Supreme Court which reversed the lower court’s decision.113

Advanced Reactors While the licensing trend was leaning towards water-cooled reactors, the AEC was still sponsoring attempts at developing systems that could develop higher pressure and superheated steam in commercial plants. The 1963 prototype Carolinas Virginia Nuclear Power Associates (CVNPA) plant in South Carolina, like reactors in Canada using a heavy water moderated, pressure tube design, operated for just 4 years.114 xi Experimental

xi In the Canadian reactors, the fuel rods were contained in individual pressure tubes through which the heavy water moderator/coolant flowed, eliminating the need for a reactor vessel containing a large volume of pressurized water surrounding the core. (Mcintyre 1975: 18)



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superheating reactors were built at Argonne and Vallecitos, California (an early private grid supplier) but the technology was not adopted by the industry.115 Sodium cooling with graphite moderation and superheating was tried at the AEC-sponsored, Atomics International (AI) built, experimental Santa Susana plant and their subsequent 77-mw Hallam plant in Nebraska built in late 1963 for the Consumers Public Power District.116 Hallam was followed by the AI built 61-mw, Detroit Edison Company, Enrico Fermi Liquid Metal Fast Breeder Reactor (LMFBR) plant in Monroe, Michigan which operated for over ten years.117 There were start-up difficulties in these plants with damage and leaking. While such problems were to be expected with advanced technology, it may have soured the concepts with utilities that were experiencing reliability with their light-water thermal plants. Many in the nuclear industry thought that the LMFBR technology was the only economical route in the long term due to limited uranium reserves.118 In 1972, a major research and development project to create an advanced LMFBR was started by the AEC and two utilities at Clinch River, Tennessee.119 Three competitors, Westinghouse, GE and AI, combined forces for this project. In 1977 the Department of Energy (DOE) placed a new breeder core in the Shippingport reactor which operated until plant shutdown in 1982.120 The Fermi plant closed in 1972, Congress ended funding for the Clinch River project in 1983121, and with ample enriched uranium supplies there was little impetus for further breeder development in the United States.122

The Economics and Efficiency of Early Nuclear Generation All commercial nuclear plants were designed to meet efficiency objectives which originated over a century earlier. The overall efficiency of a steam-powered electric generating station, regardless of fuel source, was determined by a close interaction of all components in the system from the heat source through the prime mover and condenser, with inputs to and from the feed water heating and other auxiliary systems. The goal was to achieve an overall operational efficiency based on an ideal number drawn from the 19th-century theoretical works of Carnot and Rankine.xii Beginning in the early 1920’s power station operators used the extraction method of feed heating, in which exhaust steam was withdrawn from the turbines to pre-heat the feed water going back to the boiler to boost overall station economy.123 A few years later, re-heating the steam between separate high- and low-pressure turbine casings also improved efficiency while reducing erosion in turbine blades from moisture.124 With so many stages of heat utilization, the calculations required to achieve maximum efficiency were complex. A new concept called Heat Balancing, (also known as BTU auditing), treated the heat utilization as a balance sheet in which the usage in the components had to balance for maximum efficiency.125 Tables to plot the heat flows were augmented by heat balance diagrams in which all the important components producing and using heat in a plant were drawn in a simplified manner to illustrate the relationship of the components in the system.126 As the technique was perfected engineers of some plants showed all the important parameters of temperature, pressure, and quantity of flow on the diagrams to help designers plan the most efficient arrangements.127 With these factors included, plant

xiiSadie Carnot (1796-1832), a French natural philosopher, founded the science of thermodynamics in 1824. His work “Reflections on the Motive Power of Heat and on the Machines Adapted to Develop This Power” described an ideal cycle of steam through an engine and was later developed by Lord Kelvin (1824-1907) and others as the Carnot Cycle, the basis of a practical measure of the maximum possible output from a given power system (Wilson 1981:137, Engineering 1907: 847). In England, William J. M. Rankine (1820-1872) wrote a paper entitled “On the General Law of the Conservation of Energy” in 1853 followed by other writings expanding on thermodynamics. The cycle he described called the Rankine Cycle is used to measure the comparative efficiency of turbine power systems (Engineering 1873: 14; Babcock & Wilcox: 1960: 10-6).



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designers and operators could clearly project how changes of temperature, pressure and flow quantity in one component would affect other components upstream and downstream, relative to the total plant output as measured in gross heat rate of BTUs per kilowatt hour. Designers of commercial nuclear plants prepared diagrams for various heat rates, from initial startup through licensed maximum output.128 The heat balance diagrams for the Connecticut Yankee plant were prepared by Westinghouse in 1964. Separate diagrams for 266 megawatt electrical (Mwe), 491 Mwe, 600 Mwe and 646 Mwe were drawn up to show the plants heat rates at initial startup, and then as the plant was run up, followed by a calculated maximum possible output. Heat balance considerations were only one set of factors in determining the economic viability of the early commercial stations, whose owners used complex formulas based on assumptions regarding costs of construction, operation, and government-supplied fuel, measured against current and expected costs of coal, the main competitor. Nuclear fuel costs were part of a larger cycle including mining, enrichment, loading, burn up, and storage or recycling of spent fuel. In theory, it was expected that the government would take responsibility for all but the loading and burn up components of this cycle, but the unresolved issues of storage and recycling introduced costs to utilities which were not fully factored into early cost calculations. While there were several methods of measuring the costs of power stations including cents per BTU, or dollars per kw, the standard measure was the mill (one thousandth of a dollar per kilowatt hour.)129 Plant costs were broken down into fixed (structures and equipment), operation, and fuel. Conventional fossil-fuel plant costs during the critical planning period of the first commercial stations were about seven mills total.130 This was a number that nuclear plants had to approach to be viable. The fixed costs of nuclear plants were much higher due to their still-experimental nature and the requirements for remote siting or containment structures. Shippingport was projected to come in at 64 mills, but Duquesne Light Company was buying the steam at only eight mills reflecting the extent of the government support.131 Nuclear planners expected that perfected (larger) designs, cheap nuclear fuel costs, a credit for burned fuel, increasing demand, and stable or rising coal costs would change that imbalance.132 For several reasons their assumptions proved inaccurate. In 1964, the AEC amended the 1954 act to require private ownership of enriched fuel which would continue to be enriched at AEC facilities (known as toll enrichment) until at least 1970.133 Private recycling facilities were to be set up and the fuel price was stabilized.134 However, nuclear fuel cost projections failed at the tail end of the nuclear cycle because the expected credits from recycled fuel never materialized due to poor planningxiii.135 Starting in 1958 the AEC sponsored the development of rail, truck and barge shipping casks for the spent fuel rods.136 The failure early on137 to set up storage locations, and possible public objection to projected routes, resulted in the used fuel being stored in plants’ increasingly crowded spent-fuel pools.xiv This came to be known as the “stowaway cycle.”138 In projecting profitability for the nuclear industry, planners took as a given that demand for electricity in the US doubled every ten years and that coal burning plants would not be able to meet that load.139 A graph provided by the AEC in a 1959

xiiiUp to 1971, no commercial reactors had shipped any fuel for recycling (Osbourn and Larson 1971:247)

xivBetween 1974 and 1980, Connecticut Yankee shipped a total of 83 fuel assemblies to General Electric and Battelle , with the remaining 1,019 removed assemblies stored in the spent fuel pool (van Noordennen 2005). Battelle is a Columbus, Ohio based, global, non-profit scientific research and management enterprise founded in 1929 which assists the DOE in operating the national laboratories at Brookhaven, Oak Ridge, and Idaho (Battelle 2003).



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study of operating costs showed nuclear stations of increasing size (450 mw) dropping to meet a flat line for coal station costs.140 More conservative projections aimed at establishing a nuclear parity with coal plants stated that to be competitive, the plants would have to reach 1,000 mw.141 Any belief that coal costs which had been dropping from 1948 to 1958 would rise in the 1960s142 was also in error. Seeing nuclear power as a second threat after oil to their hegemony, the coal industry came up with new methods and technology that generally kept prices stable and actually lowered costs in some areas.143 Coal companies encouraged or built power plants close to coal mines that were worked with giant stripping shovels or advanced mining machines. Though these “Mine-Mouth” power plants were sited far from load centers, they were made feasible by extra high voltage (EHV) transmission lines (500 kilovolt amperes) which could send power economically hundreds of miles at five mills.144 At the same time, railroads concerned about competition from oil-or natural gas-fired stations, coal slurry pipelines, and the Mine-Mouth/EHV technology came up with the unit/integral coal train concept.145 The unit train of all coal cars had favorable rates (up to 35% cheaper than regular trains146) as it shuttled between a mine and power station near the load center. The permanently coupled integral train took the concept even farther. The unloading process was streamlined by providing the coal cars with rotary couplers allowing automatic high speed discharge from car dumpers.147 One utility, Commonwealth Edison, expected to save over four million dollars per year from these innovations.148 At the same time improved fuel-burning technologies and super-high-pressure boilers were driving down the number of BTUs required to produce a kw-h of power.149 Thanks to these advances, and the use of large marine colliers for delivery of coal to plants on navigable waterways, costs of coal-powered generation were not much more than the Jersey Central Power & Light Company Oyster Creek Plant in Toms River, New Jersey, the nuclear industry leader at four mills.150 As early as 1965 nuclear industry planners were acknowledging that improvements in coal technology had changed the equation but they claimed that it benefitted the nation as a whole.151 By 1967, the Tennessee Valley Authority (TVA) was projecting its new fossil-fuel and nuclear plants to come in at under three mills.152 Thus during the critical early phase of commercial nuclear power, there was considerable pressure on profitability. Another factor that had to be considered in nuclear station economics was the “fit” with the fossil-fueled stations on the grid. The fact that PWR/BWR nuclear stations had to be of large size to be economic153 posed a problem for the utilities since their intermittent fueling meant that during shut-down for refuel, a large block of power had to be replaced. Still committed to water reactors, Westinghouse sponsored development of a homogeneous breeder reactor system in which nuclear fuel in a slurry loop could be replaced without shut-down.154 Their Pennsylvania Advanced Reactor (PAR) concept never reached construction. Continuous on-line fueling was also a goal of Combustion Engineering’s proposed heavy-water-moderated, organic-cooled 500-mw plant.xv CE was aiming for generating costs of three to five mills in full-scale plants but their technology never came to fruition.155 The only widespread use of continuous fueled reactors in North America are those of the Canadian Candu series.156 Despite the AEC’ encouragement of (and international use of) diverse and more efficient reactor types, the pressure of getting plants on line and making a profit led American utilities to concentrate on the apparently-reliable, somewhat-efficient light-water reactors.

xvOrganics are a class of compounds (diphenyls, terphenyls, etc) derived from or containing hydrocarbon radicals. They were first produced by Faraday in 1850 through compression of oil gas (Oxford English Dictionary 1989: v.X, p. 675[Phenyl] and v.XI, p.920 [Organic]). Organics have many benefits as a reactor coolant or moderator: providing a compact core, low system pressure, lack of reactions with fuels or water, compatibility with standard metals, and production of higher temperature steam (Balent 1959: 120).



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The First Full-Scale Commercial Nuclear Plants and Construction of Connecticut Yankee By 1963, the stage was set for the first nuclear power stations that could function in multi-station grids on a nearly equal power production and operational cost footing with coal- or oil-powered units. These were important criteria because it was clear that the nuclear stations were not economical unless they were putting large blocks of power into their respective grids.157 The first “full scale” stations were the three-loop 436-mw San Onofre Nuclear Generating Station of the Southern California Edison Company and San Diego Gas and Electric Company in San Clemente, California and the four-loop 616-mw Connecticut Yankee Nuclear Generating Station in Haddam, Connecticut. A 1968 article in Scientific American titled “The Arrival of Nuclear Power” noted the importance of these plants in the maturation of commercial atomic energy.158 Both were Westinghouse-designed plants and both began commercial operation on January 1, 1968.159 They were closely followed by the 650-mw Oyster Creek station and Niagara Mohawk Power Corporations’s Nine Mile Point 610 mw-plant in New York State to round out what has been called the “commercial phase.”160 The long lead times, large size, and problems of engineering containment buildings virtually assured that only a few design and construction concerns in the United States would share the work. Ebasco Services, Inc., Sargent & Lundy, Burns and Roe, Inc., Stone & Webster Engineering Corporation, and Bechtel Power Corporation engineered or built many of the early and commercial phase plants and the latter two would figure in the history of Connecticut Yankee.

Planning and Corporate Organization for Connecticut Yankee

The early success of the Yankee Rowe station, which began commercial operations in 1961, and the increased demand for electricity in Connecticut prompted the states’ three largest utilities Connecticut Light & Power Company (CL&P), Hartford Electric Light Company (HELCO), and United Illuminating Company (UI) to consider another PWR plant in April 1962. Initially organized as the Nutmeg Electric Companies Atomic Project, this consortium soon concluded that a nuclear station could be competitive against fossil-fuel generation over the life of the plant, using the then-common assumption that coal (and to a lesser extent oil) costs would rise. As discussed above, this assumption later proved false, although in regional terms the construction of more nuclear generating capacity contributed to lower costs in conjunction with pressure on coal prices, the introduction of larger generating units, higher-voltage long-distance transmission facilities, and increased coordination among power companies. Nutmeg Electric moved quickly to option the 500-acre site of what became Connecticut Yankee in Haddam Neck, and by the end of 1962 selected Westinghouse to produce the major plant components and Stone and Webster to design, engineer, and build the plant.xvi At the same time, the considerable costs involved, plus the model of the Yankee Rowe consortium and the long history of cooperation among New England power producers,xvii led to the dissolution of Nutmeg Electric and xvi The Boston company was founded by Charles Stone and Edwin Webster in 1889 as an electrical testing lab. The company grew to provide worldwide engineering consulting with a particular emphasis on design and construction of power stations. In 2000, it became a subsidiary of the Shaw Group of Baton Rouge (European Construction Institute 2005: Website, 1; Hoovers 2005: Website, 1).

xviiThe Connecticut Valley Power Exchange, consisting of CL&P, HELCO, and Western Massachusetts Electric Company (WMECO), was the nation's first electric power pool when created in 1925 (Northeast Utilities 2005).



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the creation of a new corporation to build Connecticut Yankee in December 1962. The Connecticut Yankee Atomic Power Company (CYAPCO) expanded the consortium beyond Connecticut to include eleven utilities. As plant construction was about to begin, the project became a factor in the negotiations leading to the 1966 creation of Northeast Utilities, an affiliation of CL&P, HELCO, and Western Massachusetts Electric Company (WMECO), the latter also an owner of CYAPCO. The NU system, later expanded by the absorption of Holyoke Water Power Company and Public Service Company of New Hampshire, immediately became the largest utility in New England and one of the twenty largest in the nation.161

Connecticut Yankee Initial Licensing, Construction, and Initial Operation Provisional and final construction permits for Connecticut Yankee were issued in May and June 1964. Plant siting had to conform to the 1962 AEC Reactor Site Criteria which allowed engineered safeguards to replace remote siting. An additional document, Calculation of Distance Factors for Power or Test Reactors (TID-1844) guided the process which included the maximum allowed releases, containment capability, and environmental conditions at the proposed site to arrive at an exclusion zone.162 The efficacy of the safeguards including structures, safety injection, water sprays, and filters was balanced against the TID’s recommended distance factor. In the case of Connecticut Yankee, the plant’s engineered safeguards reduced the exclusion radius from about a mile to 1,700 feet.163 In addition to the exclusion zone, low population zones and population centers were considered. For instance, the low population zone was one where it could be expected that the residents could be protected from a hypothesized major accident while receiving only a specified radiation limit.164 Although Connecticut was considered to be a seismically stable area, borings were taken and the plant was designed to be able to shut down safely in a ”..moderately strong earthquake...”165 Concrete pouring for the containment and turbine pedestal foundations began in August 1964. The reactor vessel was installed in May 1966 and construction was completed in early 1967. The plant received its provisional license (No. DPR-14) from the AEC in June of that year and initial reactor criticality followed in August166 AEC licenses governed the power level of the reactor which was measured in megawatts thermal (abbreviated mwt).xviii For startup, the reactor was limited to 1473 mwt out of a possible 1825 mwt.167 Electricity generation began in August. Under the provisional license the AEC closely monitored start-up activities. During the period before full power operation, adjustments were made to equipment, leaks were sealed and turbine stop valves were modified on two occasions. Most of the work was done on the secondary systems outside of containment with some power production continuing. A repair to a steam generator access door a few months after start-up did require an output drop to less than 50 mw. Commercial operation began in 1968. The amendment to the provisional license for full power operation was not granted until February-March 1969 and 600-mwe generation was not achieved until January 1970.168

xviiiFrom the beginnings of the electric power industry, the power of boilers and turbines was rated in horsepower and the power of generators in watts, kilowatts (a thousand watts), and megawatts (a thousand kilowatts.) From their inception, the output of power reactors was measured in megawatts of heat (Ford 1955: 492.) and later in thermal megawatts. The electrical output in megawatts was lower than the thermal number due to loses in the nuclear steam supply system and generator. In this historical overview, the capacities of nuclear power stations are given in megawatts of electrical output as described in contemporary and later documents. Published figures for a station can vary because the AEC/NRC often allowed increases in output over the life of the plant. The abbreviation mwe came into use in the 1960s and was common in the 1970s. Dates of stations may vary between sources due to the length of time between completion and commercial generation.



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Connecticut Yankee Containment and Primary Systems Connecticut Yankee containment structure and primary systems combined some elements of the Shippingport and Yankee Rowe stations reflecting ten years of development. Shippingport’s proximity to a population center and undeveloped standards for mechanical engineered safeguards led the designers to place the steel reactor, steam generator and auxiliary system vessels largely underground. The Yankee Rowe plant had all its reactor systems surrounded by concrete in a completely aboveground steel sphere.169 Connecticut Yankee reverted to a sub-grade reactor location inside a newer above- and below-ground, industry-standard reinforced-concrete straight-walled cylinder with a hemispherical top known as a “right circular cylinder.”170 The steel vapor shell was attached to the inside surface of the outer concrete wall. The main components within containment were the reactor; four steam generators and coolant pumps; pressurizer, emergency core cooling system (ECCS), ventilation and filter equipment; refueling systems; and overhead crane. The reactor was enclosed in a separate concrete chamber by a primary shield wall which isolated the coolant pumps and steam generators from radiation, allowing access shortly after shutdown.171 A concrete floor over the reactor and pumps provided a surface for access to the reactor head for refueling. Additional protective walls separated the pumps and generators into pairs. A second concentric circular concrete wall (secondary shield) further isolated the primary system from the containment shell and provided the support for the “polar” overhead crane that rotated and traversed to cover all the equipment areas. Between the secondary shield and the outer wall of containment were auxiliary systems. The containment building was closely abutted to the spent fuel building and turbine building to ease fuel bundle transfers and keep steam pipe runs short. The Connecticut Yankee reactor was generally similar to the Rowe reactor but the increased output required a wider and higher vessel, weighing over twice as much and operating at greater pressure. The lower control rod supports used at Rowe were eliminated so the core sat lower in the reactor. Unlike Rowe, the bottom head of the reactor was penetrated for instrumentation devices. The nuclear fuel was clad with stainless steel. While the Zircaloy cladding used at Shippingport had superior nuclear properties, it was considered hard to fabricate and not worth the cost at that time.172 In later years, Zircaloy rods were tested at Rowe, and two complete assemblies of Zircaloy clad rods were included in early Connecticut Yankee cores for testing.173 As a result of these tests which showed the potential for longer core life, the fuel rods were being completely converted to Zircaloy cladding in the years before shutdown.174 Control rod materials were the same as at Rowe. The “canned” main coolant pumps used in the first two Westinghouse stations were succeeded by a new shaft-seal design with almost three times more output. They also incorporated flywheels which insured vital extra seconds of pumping power after a power failure. Since the reactor was nearly at sub-grade a horizontal fuel canal connected the reactor cavity and the spent fuel pool. Secondary Systems for Electrical Generation and Feed Water Control

Steam Turbine Design, Construction, Operations At Connecticut Yankee, the heat energy to rotative energy conversion device was a Westinghouse/Kraftwerke Union (KWU) three-casing, tandem-compound turbine direct-connected to a Westinghouse generator. The nominal turbine output was 619,328 kw with a maximum output of 648,527 kw or 869,339 hp.175 At that load, the turbine was taking 7.463 million pounds of steam per hour. The turbine included one high-pressure and two low-pressure units, which was typical of many large nuclear turbines of the era. The turbines of the



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Connecticut Yankee unit were all on a single shaft with a high-pressure element exhausting into twin low-pressure units. This design was known as a tandem-compound arrangement to distinguish it from cross-compound types which had two or three separate turbine shafts.176 Turbines were also typed according to the directional flow of steam through the casing. The Connecticut Yankee units were double-axial flow types, in which the steam entered the center of the casing and flowed outward to each end.177 One advantage of this design was that the thrust on the blades was well balanced which simplified the design of the support bearings.178 The Connecticut Yankee turbines were also categorized by their exhaust arrangements. The steam exhausted each casing from two ports at each end, called quadruple exhaust. The splitting of the exhaust path allowed a greater flow without greatly increasing the size of the casing ends. This was an additional benefit of the double-flow design.179 The steam flow volume dictated the size of the exhaust ports, which in turn dictated the length of the last row of blades in each stage. As discussed below, blade size is a critical factor in turbines because of centrifugal forces acting to pull the blades out by the roots. Casing size and blade length had to be increased as steam pressure dropped and steam volume increased during the flow of steam through the turbine.180 The constraint of relatively poor steam conditions from the pressurized water reactor generators required larger-diameter blading in the last stages than were found in fossil-fuel power stations. Developing steam turbine designs to operate with the first generation of full-scale nuclear reactors of the late 1960's proved to be to an engineering challenge for Westinghouse and General Electric.181 The pressurized water reactors favored by Westinghouse, B&W and Combustion Engineering had a design limitation: their use of ordinary water as the reactor coolant severely limited the pressure and temperature of delivered steam.182 The transfer of heat from the reactor to the steam generators by an indirect heat exchange loop contributed to this problem.183 Even the boiling water reactors of General Electric were limited in their output temperature.184 The Connecticut Yankee reactor produced steam at 690 psi and 501 F. Coal and oil fired central stations of the early 1960's generally had boilers operating at over 3000 psig and 1000 F.185 The direct impingement of combustion gases on the water filled generating tubes explained part of their higher operating conditions. In addition, fossil fuel plants utilized superheaters to add extra heat to the steam by running the steam back through the boiler before it went to the turbines. The high temperature steam was very dry which simplified the engineering of the turbines. The pressurized water reactors of the 1960s could not provide any superheat. In an attempt to achieve higher temperatures, some early plants such as Con Edison’s B&W-built Indian Point Plant of 1965 had an oil-fired superheater to improve the steam conditions.186 The nuclear power industry did not pursue that solution. B&W turned to “Once Through” steam generators in the early 1970's which gave a modest degree of superheat.187 Westinghouse and Combustion Engineering continued on with their proven U-tube generators producing saturated steam, with a temperature the same as that of the water from which it was liberated.188 Having made that decision, Westinghouse attempted to design turbogenerators that could effectively utilize huge amounts of relatively poor quality steam. These were machines that were as large or larger than existing high-pressure, high-temperature fossil-fuel designs and had to be because the economics of relatively small nuclear plants were poor.189 In designing the Connecticut Yankee turbines for relatively low-pressure, low-temperature steam conditions, Westinghouse had to build units working on parameters not seen in large power stations since the late 1920s,190 by which time a number of reliable designs were available.



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The Connecticut Yankee use of the three-casing, tandem-compound turbine direct-connected to a generator was a direct descendant of the groundbreaking reaction turbine design patented by Sir Charles Parsons in England in 1884, for which the Westinghouse Electric Corporation of East Pittsburgh was the original American licensee.191 The three-casing arrangement was an efficient, practical way of handling the huge increase in volume that occurs as steam works its way through the turbine.192 In Parsons’ early machines, a number of increasing diameter blade wheels in a single unit utilized the energy of the steam as it flowed through the blades, losing pressure and gaining in volume. As steam pressures got higher in the twentieth century, builders split the turbine blade stages into high pressure (hp) and low pressure (lp) casings (called compounding) with the steam passing out of the hp turbine via exhaust ports and into the lp turbine.193 The lp casing was larger to accommodate the increase in volume. In addition, each casing was enlarged at the exhaust end to provide free flow for the steam.194 The stationary and moving turbine blades increased in length along the steam flow path to fill the casings. By 1920 the first large central station turbines in the 15,000-60,000 kw range built by General Electric and Westinghouse had several years' operational experience which included a spate of serious accidents. In some cases blades were completely shed from their mounting discs, in others the discs burst at high speed and wrecked the casings which also cracked from temperature stress.195 It was clear that engineering had not kept up with the size of the machines. The turbine rotors of that period were of built-up design, including forged spindles with cast steel blade attachment discs bolted or pressed on.196 The bore holes of the pressed on discs were actually machined slightly smaller than their mating spindle diameter. For assembly, the discs were heated to expand the hole and then forced on the spindle with hydraulic pressure. When the components returned to normal temperature they were locked together producing considerable stress at the mating surfaces. A machined steel key was inserted into a slot cut in both the spindle and disc to prevent rotative separation. Investigators used high speed photography on test rotors which showed that the disc wheels were flexing. Metallurgical examinations showed cracks were emanating from keyways, balancing holes and any rough discontinuities in surfaces. Remedies included stiffening the discs, using forgings instead of castings and rounding off corners known as “stress raisers” in the key way areas. It was discovered that cast iron casings “grew” from the higher temperature steam produced by pulverized coal boilers requiring substitution of cast steel. There were also problems with blading clearance, oiling systems and bearings that had to be addressed. At that time a primary problem of turbine builders was blade design. Securing the rotating turbine blades from destruction by vibration and ejection required advanced metallurgy and specialized mechanical fastenings. Erosion of the blades from wet steam in the last stages also became a problem.197 The solutions to these problems emerged from cooperative engineering between boiler manufactures and the turbine makers. Westinghouse started building one-piece forged rotors in the early 1920s which reduced the risk of assembly flaws.198 Boiler designers increased the superheat so that the steam stayed dry through the turbine cycle reducing the chance of wet steam damaging the elements. They also increased pressures, which helped turbine designers tackle another high-risk area: the blades in the last rows of the low pressure sections.199 The longer blades in those areas were particularly susceptible to stress cracking at their attachment roots, wearing along their impingement surfaces (erosion), and centrifugal force working to tear them out of the discs. At the same time it was recognized that steel under stress was particularly vulnerable to corrosive media, a condition first called “season cracking” in the early twentieth century due to its occurrence during wet weather. It was later known as Stress Corrosion Cracking.200 This phenomenon was observed by jewelers in the 19th century,201 and was seen in brass cartridge cases shipped from Britain to arsenals in India



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in the early 20th century. It resulted from corrosive media attacking metal parts under mechanical stress produced by applied forces, forming operations, or expansion/contraction.202 The higher pressures generated by advanced fossil-fuel boilers of the mid-twentieth century mitigated stress corrosion cracking (SCC) and allowed turbine designers to build smaller machines with high outputs. New solid forged rotors allowed rotation at 3600 rpm and were resistant to mechanical failure. It was easier to engineer blading near the smaller exhaust ports and the blades edges were protected from erosion by attached hard metal alloy strips.203 The high speed engineering also saved money in manufacturing and foundations. 204 By 1950 there were reliable standardized designs putting out 100,000 kw (100mw) at 3600 rpm. The largest turbines of the period still required built-up rotors, and had occasional failures,205 but the excellent steam conditions produced by fossil fuel boilers of that era ensured reasonable reliability.206 Output pressure and temperature were slowly increased through the 1950s. By the time the Connecticut Yankee unit was ordered, coal-or oil-fired boiler/turbine generators were producing 600 mw.207 The trend in the first hundred years of turbine development was to produce the smallest possible machines producing the highest power at high speed with reliability. This achievement was aided by boiler designs that produced steam at very high pressures with high superheat. A 1966 article on the Connecticut Yankee turbines and their contemporaries in Westinghouse Engineer, the company’s house magazine, described the extra lengths that their engineers took to make the huge nuclear power station turbines function in the poor steam conditions.208 Inlet and exhaust ports had to be larger than those at contemporary fossil-fuel plants to do the same work. Rotor speed had to be reduced to 1800 rpm, known as “half speed” to prevent ejection and erosion of the very long last row blades resulting from the large ports.209 General Electric (Westinghouse’s chief competitor) used the same rational for its low-speed nuclear turbines.210 The lack of superheat required steam drying between the stages to protect the vulnerable low-pressure units from moisture. Live steam from the reactor was then used to bring the temperature back up. This process of reheating between stages was used in fossil fuel stations, but the 90-100 degrees of reheat obtained in the Connecticut Yankee plant and contemporary nuclear plants was very low in comparison to levels obtained in fossil fueled stations. Live steam was even sent direct to the exhaust ends in an attempt to pull out entrained water. Large amounts of steam were used in these areas, but it was not really a problem because the reactor was sized to produce much more steam than was needed for actually powering the turbines. More water removal occurred at extraction points where steam was bled off to heat the feed water going back into the steam generators.

Turbine Governing All the design calculations for maximum rotative speed depended on the turbine governing devices doing their job within specified limits. The function of the turbine governing system was to control the speed of the unit to ensure that the generator was producing even, continuous, high quality electric power. In addition the governing devices prevented over speeding which could lead to explosive destruction of critical components. The Connecticut Yankee governing system evolved from the flyball devices used by millers in the 18th and 19th centuries to control grind stones in their water and wind powered corn mills.211 This was an early feedback device: a self-regulating mechanism.212 James Watt later took that design and applied it to steam engine governing. He used it to open and shut a valve on the steam pipe.213 Later designers adapted the



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flyball governor to control the admission of steam by varying the settings of the steam admission valves with mechanical linkages. Westinghouse used the same device on its first turbines.214 By 1920, the necessity of precise speed control for electricity generation from large turbines led to the oil pressure relay type governor.215 This utilized the pressurized bearing lubricating oil as a working fluid. The Connecticut Yankee governor dispensed with the mechanical complication of a large rotating flyball governor and instead activated the relays with speed sensitive oil pressure supplied by a pump impeller on the turbine shaft.216 A key element in the Connecticut Yankee turbine governor was the servo control system developed in the 1920s in which a powerful control motion was produced from a remote and relatively weak signal via sensors, amplifiers, and servo-motors.217

Condensate and Feed water Components After the steam finished its work in the turbines, it was condensed back to water and recycled. Two surface condensers (nos. 1A and 1B) stood directly below the low-pressure turbines. The condensers were of shell-and-tube construction in which cooling water and exhaust steam were not mixed, a standard design evolved from mid-19th-century steamships which needed fresh water for feeding high pressure boilers.218 In principal the Connecticut Yankee condensers simply reversed the heat exchange of the steam generators. Water pumped from the Connecticut River flowing through tubes cooled and condensed steam flowing around them. At full load, 93,000 gallons per minute (gpm) of water was required for condensation.219 The condensed steam (condensate) was the main source of feed water for the steam generators. The collection points were the hot wells which constituted the lower two feet of the condenser shell, and normally held 33,000 gallons -- enough for 2.75 minutes of steaming.220 Two condensate feed pumps on the Turbine Building ground floor removed water from the hot wells and directed it to Reactor Containment.221 The Connecticut Yankee condensation and condensate component design was a “once through” system for condensing the used steam and returning it to the boilers. All the cooling water needed was drawn from the Connecticut River and sent back to the river in a heated condition.222 This was a common choice in the less environmentally-aware early 1960s. The other, more expensive option would have been a closed system in which the condensing water would pulled from the river or preferably a dedicated pond to be cooled in towers and sent back to the source.223 Westinghouse was an early advocate of marine-type shell-and-tube surface condensers for utility steam turbines like those supplied to Connecticut Yankee.224 Surface condensers were originally necessary for conserving fresh water boiler feed in steam ships operating in salt water. Most early land turbine installations used simpler condenser types operating on barometric or jet mixing principals. The increasing size of turbines in the twentieth century led to widespread reliance on the ability of surface types to condense large amounts of steam and provide high levels of vacuum for the lp turbines.225 Their heavy water flow required plant siting near rivers, lakes, or oceans. Because they were originally designed to operate in corrosive ocean salt environments, they had non-ferrous metal tubes to resist wastage. This technology transferred well to power plants in tidal estuaries where salt or brackish water was the rule. Their complex construction with thousands of closely spaced tubes was still vulnerable to corrosion and fouling by biological organisms.226 Tube material had to be carefully chosen to suit the particular local water chemistry. Choice of a closed cycle cooling system not reliant on river water would have reduced bio-fouling and the chance of corrosion.227 In the Connecticut Yankee units, the bulk of the original tubing was fabricated of Admiralty Brass. The brass



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tubes deteriorated due to ammonia induced stress cracking, but operation continued by plugging the affected tubes. This could only be considered a stop-gap repair since output would ultimately drop.228 Failing condenser tubes were a problem in many aging American power plants, and as discussed below eventually led to complete tube replacement at Connecticut Yankee.229 Studies done to determine that there would be no impact on fish and bird life in the river adjacent to and downstream from the heated discharge water from the plant were completed after plant design and construction.230 Generator and Transformer Designs Connecticut Yankee turbines drove an 869,339-hp generator which proved far more reliable than the low pressure turbines and condensers. Although the generator also had to be scaled up for such large output at the half speed of 1800 rpm improvements in mechanical construction, metallurgy, insulation, and cooling during the previous sixty years of development kept pace with the engineering requirements. The basic form of the Connecticut Yankee generator evolved from designs developed for European systems of alternating-current high-tension transmission. George Westinghouse recognized the superiority of this system over direct current as early as 1885 and he aggressively purchased patents and licenses from several engineers on the continent and from Tesla in this country in an attempt to lock in the technology.231 The main benefit of AC high-tension distribution was in economy of copper transmission wire, which was a major portion of the capital expense of electrification.232 A strong influence on emerging technology was the experimental 108-mile 25,000-volt polyphase transmission from Lauffen to Frankfort in Germany in 1891. This installation pioneered high tension, three-phase transmission, with a water powered revolving-field generator and step-up transformers.233 The use of a three-phase generator gave smoother power, greater capacity and saved money in conductors.234 In addition the first reliable AC motors worked better on a polyphase system.235 The main constructional feature was the use of a revolving-field magnet surrounded by stationary armature conductors. This arrangement (which reversed earlier practice in which the armature revolved inside the stationary field magnets), disposed the main elements where they could add to structural simplicity and strength. The copper conductors arrayed in the stationary armature were easier to brace against displacement by electromotive forces. This also eliminated the difficulties of taking high voltages and currents from a moving element.236 The invention of silicon steel for the conductor-supporting laminations greatly reduced stray currents allowing more output per pound of metal.237 Placing the relatively simple field magnet wiring on the rotating armature allowed this element to be strongly built to resist centrifugal forces. Increasingly efficient insulating materials for the copper conductors also played a part in making the early nuclear era generators possible. Varnished paper and cloth used in the first generators gave way to mica in the 1890s.238 In the 1920s asphalt-bonded mica was the norm. Resin-bonded mica and fiberglass came into use around 1950.239 The most important factor in making large generators like the Connecticut Yankee unit possible was hydrogen cooling. In the late 1920s, designers realized that improvements in heat removal in natural circulation air-cooled generators would enable them to get higher outputs from smaller machines.240 Water cooling the air helped, but the Swiss invention of hydrogen cooling in the 1920's paved the way to greater outputs.241 Hydrogen’s lower density and higher heat transfer boosted outputs. At first the pressure was just high enough to keep out air.242 By increasing the pressure to 60 psi and ducting gas through the



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conductors, reliable machines of 600 to 1000 mw served the Westinghouse built plants. Essential to the economics of long-distance transmission was the adoption of alternating currents and step-up transformers. One of the technologies that Westinghouse acquired was the transformer design of Gaulard and Gibbs in Britain.243 They utilized the principal of electromagnetic induction: current entering the “primary” coil of copper discs at one end of a magnetic circuit produced an electro-magnetic flux which induced a current in an opposite “secondary” coil. Westinghouse engineers rapidly improved this design using copper wire coils and stacked iron plates for the magnetic circuit.244 By increasing the number of wire turns on the second leg, a relatively low voltage/high current (amperage) incoming current produced an opposite high voltage/low amperage output.245 This had two features that aided long distance transmission: it allowed for more powerful generators which did not need hard-to-engineer high-voltage connections, and economized the use of copper transmission wire. The high voltage/low amperage output of the new transformers allowed much more electricity to be sent though a given wire size which could be economically strung for hundreds of miles.246 At the receiving end, the same type of transformers reduced the voltage to a safe level for industry or home use. The main areas of development were similar to those for generators: core, insulation, and cooling. The critical core metallurgy was a challenge early on to designers because with ordinary iron, electrical losses increased with time.247 The same silicon steel used in armature cores solved that problem.248 Insulation materials evolved along much the same lines as in generators. By the 1930s designers began to design units that could withstand lightning strikes which required a new order of testing, mechanical integrity, and surge resistance.249 Early transformers tended to be cooled by either forced air or oil, with oil becoming the predominant method for power stations. Natural convection of the heated oil gave way to water cooling of the oil and later to forced oil circulation in external tubed coolers.250 By mid-century, thermostat-activated fans were added to draw the heat off from the oil in the cooling banks.251 The Connecticut Yankee output transformer was derived from those “double- and triple-rated” units, as was a step-down transformer which produced a lower voltage to supply the reactor coolant pumps. Summary of Connecticut Yankee Operations and Repair Issues 1970-1974 The first refueling shutdown began in April 1970 and took about two months, with later refuelings scheduled roughly every year. Each new core required extensive design and engineering of the fuel arrangement, control rods and moderator chemistry to ensure the required power output. Refueling shutdowns also allowed for operational improvements and introduction of other new or re-designed facilities. During the first refueling episode, a new Diesel Generator Building (see HAER No. CT-185-I) was completed to enhance auxiliary power supply. Enhanced or enlarged facilities to process gaseous and liquid nuclear waste were completed in 1973-74, largely during the fourth refueling shutdown. Until mid-1973, plant designers and operators proclaimed satisfaction with what appeared to be trouble-free operations and the production of some 21 billion kilowatt hours, with particular satisfaction expressed about turbine performance.252 At about the same time, a GE engineer stated that the erosion rate of their nuclear turbines was no worse than their fossil fuel units,253 and Nuclear Safety, the bimonthly review of the Atomic Energy Commission, indicated no turbine erosion or corrosion problems after reviewing the performance of twenty-eight light water reactors.254 Soon after these articles appeared, however, significant design problems appeared in the steam generators and low-pressure turbines.



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Turbine Repairs 1973-1974 - Emerging Issues of Stress Corrosion and Erosion

The low-pressure turbines at Connecticut Yankee began to fail in the spring of 1973.255 The first repair on the No. 2 unit was in June 1973, and the spindles on both low-pressure units were replaced between July-December of that year, possibly using spares provided by Westinghouse. The rotors must have been severely degraded since the repair was not done during refueling.256 The station was out of service for over five months. Another month-long repair requiring shut-down began in 1974. The problems included disc cracking, blade root cracking, and erosion of the stationary and rotating blading.257 Indirect evidence from other plants suggests these repairs reflected corrosion problems which began soon after Connecticut Yankee began full-power operations in 1969, if not earlier and a re-emergence of the blade erosion problems that had occurred in the early 20th century. Beginning in 1965, while Connecticut Yankee was under construction, the steam generators and turbines in one of the pioneer nuclear power stations of the Central Electricity Generating Board (C.E.G.B.) in England began to show damage from feed water impurities.258 Just before the low-pressure turbine problems at Connecticut Yankee became evident, a groundbreaking report on a turbine failure in another C.E.G.B. station reached the engineering journals. In 1969 the Hinkley Point A Nuclear Station had a catastrophic failure of the low-pressure blade discs on one of its turbines. The unit was very similar in layout and construction to the Connecticut Yankee unit, though smaller and running at 3000 rpm. Fragments of ruptured multi-ton turbine discs were projected in 100 foot vertical and horizontal arcs and were capable of penetrating the reactor containment. A 2-year investigation by the board revealed that the discs failed due to stress corrosion cracking (SCC). The report found that the causes of the corrosion were minute impurities in the steam, attacking very tiny defects in the disc attachment points.259 By this time, SCC was well understood at the molecular level, although it was many years before the engineering caught up with the science.260 Failure from SCC takes about four years to develop - about the length of time between full-power operation and first turbine repairs at Connecticut Yankee.261 The problems afflicting the Connecticut Yankee low-pressure turbines occurred in other first generation Westinghouse units. The near-sister plant to Connecticut Yankee at San Onofre, California had problems with cracking in the keyways that locked in the blade discs to the spindles. The carbon steel of the rotors could not handle the relatively high moisture content.262 Brookwood #1 (now Ginna) of Rochester Power and Light suffered blade failures with blade ejection requiring operation with the last row blades removed while engineers tried to find solutions.263 Ten years after delivery of the Connecticut Yankee turbines, some engineers still felt that conditions in low-pressure nuclear units were not very different from fossil-fired units and did not require new engineering.264 Westinghouse’s (and GE’s) assumption that lower rotor speeds would reduce the erosion of the last row blades may have been in error as later observations found that high revolutions led to longer blade life.265 Until c1973, GE turbines operating at PWR and BWR plants had not shown signs of stress corrosion cracking.266



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While the incidents undoubtedly caused economic harm to the utilities, public notice was probably muted by the apparent lack of safety issues, the subsequent Three Mile Island accident, and later ongoing problems with the more newsworthy steam generator tubing. It is clear that the Connecticut Yankee low-pressure turbine problems were caused by unforeseen engineering decisions and manufacturing methods used by Westinghouse. Excluding the C.E.G.B stations, these types of turbine failures were very much a United States problem that also affected non-nuclear turbines with increasing frequency from 1964 to 1973.267 For various reasons, similar turbines, even American-made ones, did not fail in Germany or Japan.268 A critical component in stress corrosion control is the chemical quality of the feed water going into the system. Very slight rises in salinity or minerals could exacerbate the deterioration of components in the steam path.269 The Hinkley Point investigators even found that SCC could also be induced by certain water purity control chemicals.270 However, in reviewing all the C.E.G.B. stations that had cracking, they found that though water quality varied it was still within operating specifications. In addition they felt that it would have been impossible to expect the controls to be any better. Their recommendations were that the details of the highly vulnerable disc keyways had to be better engineered.271 The cooling water intakes for Connecticut Yankee were in theory sited upstream of observed salinity, and the operating engineers kept fairly close controls to prevent its entry into the system.272 Water quality issues were also central to problems with steam generators at Connecticut Yankee, and at many other nuclear plants.

Connecticut Yankee and Worldwide Steam Generator Problems Shortly after full power operation began in 1970, leaks in the steam generator tubes were detected.273 Tube leaks allowed irradiated primary coolant water past the Reactor Coolant System (RCS) boundary, creating a safety hazard. Consequently, federal regulations set maximum gallons per day leakage rate for all steam generators.274 Initially no repairs were needed. Between June of 1973 and April of 1974, during repairs on the low-pressure turbines, steam generator tube leaks were stopped by explosive plug welding.275 This technique was developed around 1970 in answer to the difficulties of closing tubes in an irradiated area with manual plugging or welding. On detonation of a small nitroglycerine-based charge, molten metal cleaned the inside of the tube, and weld positioned a wooden plug to block loss of coolant.276 While plugging was effective in preventing coolant loss, it could only be considered a stop-gap as plant output would drop if too many tubes were plugged. The history and causes of steam generator problems were related to the issues noted for turbine blades and rotors, and were found in nuclear plants built before Connecticut Yankee. In early 1958, after only a few months of operation, one of the B&W horizontal u-tube steam generators at the Shippingport station developed a tube leak. Testing with an eddy-current devicexix showed extensive surface stress corrosion cracking. An analysis by Westinghouse and Duquesne concluded that the chemicals then being used in Duquesne’s fossil-fueled boilers to control oxygen levels (sulfite) and the pH of the feed water (phosphate) were not as effective in the smaller nuclear (in comparison to fossil-fuel) steam generators.277 Changes in the type of phosphate, amounts of sulfite and modifications to the tube arrangements controlled the problems. xixEddy-current testing of boiler tubes is a remote, non-destructive procedure that utilizes electro-magnetic fields from a probe to find faults by a change in signal intensity caused by variances in wall thickness and cracking. (Singley et al 1959: 753)



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British investigations of feed water contaminants in the late 1960s, noted above, found that steam generator tubes were scarred by corrosion.278 The intensive investigation by the board indicated several causes including steam generator tube configurations, breakdown of the feed water heater tubing and carryover of the particles, insufficient air removal from condensed steam, and failure of resin filter beds to remove organic compounds from lake-sourced feed water. Improving the filter media, eliminating air from the feed water and adding phosphate water treatment largely resolved the problems.279 Six years after Yankee Rowe began operation in 1960, leaks were detected in its steam generators with stainless steel tubes, but operators controlled the problems.280 A few years later, utilities with Westinghouse, B&W, and CE units tubed with Inconel 600 (nickel-iron-chrome alloy281) tubes began to have an epidemic of tube and tube support failures that eventually affected forty stations. Some plants had steam generators with 20% of their tubes plugged to prevent leakage.282 The types of degradation included wastage, pitting, denting, cracking of the tubes, support plate damage, and mechanical damage from vibration and loose parts. In analyzing steam generator problems, a 1988 NRC report found a complex interaction between the mechanical design, materials, fabrication methods, water treatment chemistry and corrosion products from the plant secondary systems, mainly the condensers.283 The construction of the steam generators, with thousands of tubes, tube seal plates, and bracing, provided many areas for corrosion materials to accumulate and do damage. Designers of the first PWRs saw the water in the primary reactor/steam generator loop as a potential source of reactor or steam-generator tube damage due to the addition of acidic boron to help control the nuclear reactions. While they specified precise chemical parameters for that system, less attention was given to secondary system feed water control.284 In both fossil-and nuclear-fueled stations, the secondary system condensers that returned the steam leaving the turbines to water were designed to be the ”first line of defense” against the entry of corrosion products, yet they also could be their main source.285 The thousands of tubes providing the condensing surface area had to be sealed into tube plates at both ends and were acknowledged by American operators to be practically impossible to keep leak-free.286 The fact that the condenser shell was operating in a vacuum meant impurities and air in the cooling water drawn from oceans, rivers, or lakes would be sucked out of any leaking tubes to mix with the condensing steam. As a result, air, chlorides, hydroxides and up to ninety other chemical compounds could enter the system and induce SCC in the steam generator tubes and also the turbine blade roots and mounting discs.287 Deaerating sections of the condensers, air ejectors and chemical water treatment were necessary to mitigate these problems. Recognizing that condensers and mechanical devices alone could not bring pH and oxygen levels low enough to prevent corrosion, engineers of fossil-fueled stations had started using extensive chemical feed water treatment around 1950. Most stations used phosphate to keep water pH below corrosion thresholds.288 Phosphate treatment did not alleviate the dissolved oxygen content of the water, so some utilities added sulfites or hydrazine as oxygen scavengers in the mid 1950s.289 Later ammonia or morpholine were added to control the pH of the water.290 Initially positive results were mitigated by increasing evidence that the hydrazine or ammonia attacked the copper tubes in feed heaters and condensers, leading to suggestions that those units be entirely tubed with carbon or stainless steel.291 American nuclear plants through the generation including Connecticut Yankee used phosphate treatment until it became evident that the phosphate built up as sludge on tube sheets causing wastage and thinning of the steam generator tubes.292 For better control of oxygen content and to forestall tube deterioration, both Connecticut Yankee and San Onofre tried hydrazine injection as early as 1970.293 With the blessing of the



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manufacturers, American nuclear plants switched to an ammonia and hydrazine-based “all volatile treatment” (AVT) around 1974.294 At first AVT reduced tube plugging, but it then began to cause other problems. The ammonia injected into the feed water (or that produced by the breakdown of the hydrazine), in combination with the dissolved oxygen, attacked the copper-based tubing that had been specified for the demonstration- and commercial-phase feed water heaters (feed train) and condensers.295 The copper shed from the tubes ended up in the steam generators where it acted as both an oxidizer and catalyst for pitting of the tubes.296 At the same time, the feed train materials and the chlorides from condenser leaks would lodge in the spaces between the steam generator tubes and their drilled, carbon-steel support plates. A resulting buildup of oxide would then squeeze the tubes (denting) leading to cracking and ruptures.297 Steel and copper corrosion products would also accumulate on top of the lower tube sheet as a sludge that would lead to inter-granular attack (a form of SCC) damage of tubes.298 The stress set up when the tubes were bent to form an inverted U-shape made that area particularly vulnerable to SCC. Even rigid control of pH did not guarantee steam generator health because the right level needed to control copper corrosion would allow iron oxidation.299 The chemistry problems were exacerbated by operations of the plants at low loads and during startup. Mechanical damage was caused either by loose or foreign parts (some left in the units by assembly or repair teams) impacting the tubes or by vibration due to inadequate supports. Clearance between the tubes and their support plates and bars was necessary in some models of generators but it allowed the large volumes of water and gases traveling through the generators at high speeds to cause relative movement (fretting) leading to tube damage.300

These problems perplexed manufacturers and plant operators. From 1968 to 1975, San Onofre modified its phosphate chemistry four times, switching between AVT and phosphate before settling on AVT at the request of Westinghouse or because of its own investigations.301 Connecticut Yankee documents suggest that the utility also tried different chemistries, and that these choices could have been a causative factor in its steam generator tube problems and its heater tube, condenser tube and turbine replacement projects (see HAER No. CT-185-C-Turbine Building).302 The power companies were evidently getting insufficient help with these problems from their suppliers, and in 1977 operators formed the Steam Generators Owners Group (SGOG) in conjunction with the Electric Power Research Institute (EPRI) to address the problems.303 SGOG research showed that the chemistry parameters set by the manufacturers were too loose. As an example, the original specification on chlorides allowed 150 parts per billion (ppb) while the EPRI/SGOG guideline limited it to just 20 ppb.304 While leak problems were usually manageable, more serious issues began in 1975 with a steam generator tube rupture at the Point Beach #1 Plant of the Wisconsin Michigan Power Company, followed by ruptures in other plants. Tube ruptures allowed much more primary coolant to escape, and could lead to other system failures resulting in serious accidents.305 In 1978, the NRC designated steam generator tube integrity as an unresolved safety issue.306 The solutions to these problems took years to develop. Tube plugging was augmented with sleeving in which a section of smaller diameter tube was inserted into the leaking one and mechanically sealed allowing coolant flow.307 Sleeving did not occur at Connecticut Yankee until after 1987.308 The AEC/NRC (Nuclear Regulatory Commission) instituted inspection programs in which the entire length of every tube had to be examined by eddy-current testing devices. The costs for inspection could reach $500,000 per day (including replacement power and provision for worker rem exposurexx) in the Westinghouse type steam generators which had fewer tubes than those of the other makers. The problems

xx Roentgen Equivalent Man=the quantity of radiation having the same effect on human tissue as one roentgen of X-rays. (Oxford English Dictionary 1989: v.4, p. 576.)



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described above were not limited to American PWR plants. In the 1970s, French stations using Westinghouse-licensed units experienced tube and tube plate corrosion. Their engineers suggested that the Inconel 600 material had to be improved with heat treatment, that the tube support plates be made of stainless steel, or that the tube/plate interfaces had to be upgraded. They felt it imperative that there be no condenser tube leakage.309 After some problems in German stations, the plants controlled the problems with improved tube metallurgy (Incoloy 800, a nickel-iron-chromium alloy), largely leak-free and corrosion-resistant condensers, and case-by-case use of either low-phosphate or all-volatile treatment.310 In addition, European operators were more willing than American utilities to shut down their plants when chemistry upsets or condenser leaks were detected.311

For generator problems caused by secondary system impurities, manufacturers recommended installation of “full flow” condensate polishing systems which used resin-filled filter beds to purify continuously all the condensate after it left condenser hot wells.312 This would have provided much greater control than the more common intermittent treatment, but some engineers opposed it because of high capital and operating costs, and because of concerns that the resins used in the system would cause their own problems. Connecticut Yankee did not add a condensate polishing system, probably for these reasons.313 Another solution which Connecticut Yankee instituted around 1977 was to “blow down” the steam generators frequently to clear out deposits.xxi It is undocumented whether this change in operation required additional plant infrastructure or NRC approval. By the 1980s, power station designers realized that slightly brackish cooling water supplies demanded more advanced metallurgy in condensers to prevent bio fouling and stress corrosion cracking.314 To reduce those conditions at Connecticut Yankee, all the tubes were replaced in 1986 with a proprietary stainless alloy Trent Sea-Cure, which came on the market in 1979. This was an attempt to prevent damage in those sections, contaminant particle carryover, and subsequent denting in the steam generator tubes.315

xxi Blowing down (also known as blowing off) a boiler was a water purification technique from the earliest days of steam technology (Rankine 1859: 453). Allowing some of the pressurized water inventory to escape from the boiler removed oil, salt and other contaminants which could settle on heating surfaces and restrict heat transfer leading to premature failure. Bottom blow-off and surface blow-off valves cleared out the two regions where substances generally accumulated. While blowing off wasted heat in conventional boilers, in nuclear boilers it also allowed irradiated water outside of the reactor coolant system boundary requiring storage tanks and filters.

While most of the PWR steam generator problems were caused by the secondary system water, the chemistry of the primary system could also cause damage. Cases of primary-water-induced cracking of steam generator tubes at the stressed U-bends began to appear.316 Damage to the reactor could also occur and was a factor in the shutdown of Yankee Rowe station.317 A more serious development was the discovery in France that the boric acid moderator added to the coolant could attack the reactor head. The Toledo Edison-Cleveland



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Illuminating Co. Davis-Besse plant in Ohio was shut down in 2002 as a result of severe corrosion around the control rod penetrations.318 During the era in which PWR owners and manufacturers were working on steam generator problems, operators of GE’s BWR’s were having their own set of troubles. While the successors to the dual-cycle plants were not saddled with PWR steam generators, they began to have a form of stress corrosion cracking in the recirculation and other stainless-steel reactor piping. Following the PWR owners’ lead, the utilities and EPRI formed the Boiling Water Reactor Owners Group which successfully addressed the problems.319 Summary of Connecticut Yankee History 1974-1984 and Reactor Cavity Seal Failure The AEC, evidently satisfied with the measures taken by Connecticut Yankee to repair the failed low-pressure turbine rotors and steam generator tubes while operating under the provisional license, authorized full-power operation at 1825 mwt on December 27, 1974 with Facility Operating License #DPR-61.320 In 1975 the AEC was replaced by the Nuclear Regulatory Commission and two years later the Department of Energy (DOE) was created. Reflecting the ongoing national failure to accommodate spent fuel, the NRC amended Connecticut Yankee’s license to allow an increase in the spent fuel pool capacity from 336 to 1172 assemblies in 1976.321 Until the mid-1980s, most major changes in plant facilities or operations were driven by national issues in nuclear plant safety. A 1975 fire in the Tennessee Valley Authority’s unfinished Browns Ferry Unit 3 in Alabama led the NRC to require upgrades of fire protection in all American plants. At Connecticut Yankee, resulting improvements included barriers, detection equipment, and fire-fighting capabilities. Work and materials storage methods were changed, with great emphasis on controlling combustibles and ignition sources. On March 28, 1979, the most serious accident in the history of American commercial nuclear power plant operations occurred at the Jersey Central Power & Light Co. Three Mile Island (TMI) unit 2 facility near Middletown, PA. There were no radiation-exposure consequences, but the reactor overheated and fuel melted. Causes included personnel error, design deficiencies and component failures. As a result of this accident, significant changes were made in the industry. The NRC issued amendment No. 42 TMI Lessons Learned Category A Items for Connecticut Yankee.322 Changes to the plant from this amendment included new accident monitoring systems, new control room instrumentation, seismic improvements to the Service Building housing the control room, and the construction of an Emergency Operations Facility Building (HAER No. CT-185-W) in 1980.323 The largest fire-protection modification at the plant, a new electric switchgear building (see HAER No. CT-185-O) proposed in 1986 and completed in 1990, was also an outgrowth of increased accident protection measures.324 In 1982-83, Connecticut Yankee modified the reactor cavity seal ring, a vital component of the refueling system, prior to the 1983 refueling.325 During the 1984 refueling that seal ring failed, leading to the most serious accident in the plant’s history.



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Refueling and the Reactor Cavity Seal Design, Failure, and Reconstruction The immersed refueling system relied on the principal of water seeking its own level between connected containers, and included the reactor cavity and its adjacent refueling (transfer) canal in containment, the fuel pool in the spent fuel building, and a transfer tube connecting them. Before refueling, water treated with boron to kill any nuclear reactions was pumped from the Refueling Water Storage Tank into the reactor cavity (surrounding the top of the vessel) to an elevation of 46.5 feet above mean sea level (equal to the surface elevation of the spent fuel pool) to ensure complete coverage of fuel bundles during the refueling process and so there would be no flow between buildings. The head of the reactor was lifted off by the polar crane as the water level rose, and was set down in a circular 47-foot-deep concrete pit within containment. With the water levels equalized, the valves and sluice gate that sealed off the transfer tube were opened, providing a continuous water path to convey spent and new fuel rods between the two structures. The refueling water height 24.5 feet above the open top of the reactor vessel provided enough clearance to fully protect the spent fuel rod bundles as they were pulled out with a manipulator crane on the refueling floor above the cavity. The crane operator then placed the bundles vertically in an upender machine in the transfer canal next to and below the mouth of the reactor. The upender set them in a horizontal position on a wheeled car that carried them through the transfer tube to the fuel pool.326 Another upender and crane handled the bundles in the pool. When it was time to bring in new fuel bundles from the spent fuel building the process was reversed. It was necessary to prevent the water in the reactor cavity from pouring down between the shell of the reactor and the surrounding concrete wall, past the neutron shield tank and then into the floor of the containment building. If that occurred when the canal was open during a transfer, in a worst-case-scenario, the water level in both buildings could drop, possibly enough to expose the entire length of bundles being carried by the cranes or a portion of the bundles in the upenders, and the stored fuel bundles in the pool.327 The subsequent heating of the rods would have produced high doses of radiation to personnel, fuel cladding failure, and possible release of radiation to the atmosphere.328 The original reactor cavity seal was a circular steel plate bolted in place between a flange around the top of the reactor vessel and the adjacent concrete, covering the annulus (opening) between the two. During refuelings prior to 1983, the seal had small leaks which led to contamination of the lower portion of the vessel.329 Connecticut Yankee engineers proposed a new seal device which consisted of a plate surrounding the opening with continuous inflatable rubber boots on the inside and outside diameters. On inflation the boots would pull down T-shaped wedges of rubber to plug the openings between the flange of the reactor and the inner edge of the plate, and between the outside edge of the plate and the surrounding structure to prevent leakage. The modification was made and it followed the recommended Plant Design Change Request (PDCR) procedures as outlined in the CFR.330 During refueling operations on the morning of August 21, 1984, after the cavity had been filled and the head of the reactor removed, the seal failed. In less than half an hour, all 200,000 gallons of water in the cavity drained down through the seal. The water elevation after the accident was at 22 feet, level with the open top of the vessel. Because all the fuel rods were still under water in the reactor, they were not damaged. The transfer tube had not been opened at the time of the accident so there was no loss of pool water and exposure of stored fuel bundles. Operators initiated the correct actions to begin pumping out water from the floor of containment. There was a small filtered release from the ventilation stack. Connecticut Yankee personnel followed procedures to notify the NRC, the state, and declared an “unusual event” in compliance with the Emergency Plan.331 Refueling was terminated and the event was declared over when the water in the lower



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portion of containment was pumped out. On dewatering it was found that the corrosive borated water had penetrated insulation on the bottom of the reactor and piping, requiring removal and repairs. An investigation by Connecticut Yankee engineers found that the seal had been incorrectly designed and tested, allowing a critical part to deform after inflation under the full head of water leading to gross failure.332 The previous design, though not completely watertight, was more failure-proof. Northeast Utilities (NU) notified licensee's of twenty-seven reactors that had a similar seal of the event through the Institute of Nuclear Power Operations (INPO) network.333 INPO was created in 1979 to share information between the utilities and the DOE. NU also evaluated the possible impact on upcoming refueling operations at its three Millstone reactors in Connecticut.334 As a result of the inquiry, hidden flaws in the refuel system design were revealed prompting the NRC to issue a bulletin to almost every operating or planned reactor in the U.S. about the danger of this type of accident.335 Several corrective measures were taken by Connecticut Yankee to prevent a recurrence. The seal was redesigned with steel rods to prevent the top portion from deforming and a backup seal was added above the main seal. A fixed wall (cofferdam) was added in front of the canal so that even if the seal failed with the transfer tube open there would still be enough water to cover the stored bundles in the spent fuel pool. The additional height of water would also give operators time to activate pool cooling mechanisms. Operators of the manipulator cranes and upenders were trained to quickly place bundles in transit in a safe position during unplanned cavity drainage.336 The sluice gate in the transfer tube was redesigned to close against a flow of water pouring out of the pool, an event that was not contemplated in the original design.337 While there would still be water in the reactor after a failure, it was required that the Residual Heat Removal pump would always be activated to provide additional circulation to prevent heating of the rods still in the core.338 At some point before 1985 the revised temporary cavity seal ring was replaced with a permanent stainless steel ring. It further reduced the chance of failure, allowed for reactor movement, saved refueling time, and eliminated worker radiation exposure.339 On December 12, 1984, the NRC issued a “Notice of Violation and Proposed Imposition of Civil Penalty and Order Modifying License” which instituted an $80,000 fine to the Connecticut Yankee Atomic Power Company. Management elected not to contest the fine but also cited a number of compliance actions on their part which they felt should have reduced the fine. These included the prompt notification to NRC and other utilities, in-depth investigation of the event and other potential causes, an extensive redesign process and co-hosting an INPO workshop on seal failure.340 Prior to the accident, NRC inspections revealed two other earlier modifications to the plant that the commission felt were not properly instituted. The changes involved radiation monitors and a control valve in the Post Accident Sampling System. The NRC held an enforcement conference in November 1983 to determine if there was a pattern of inadequate design modification processes.341 The NRC did not find that to be the case, but Connecticut Yankee instituted an improved PDCR process and sent out a letter to all personnel in Nuclear Engineering and Operations asking for in-depth questioning about all possible circumstances (described as “what ifs?”) of future design changes.342



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Summary of Connecticut Yankee Operations 1986-1996 During 1986, the NRC issued an amendment for Connecticut Yankee regarding specifications for three-loop operation.343 The four-loop design of Westinghouse reactors allowed one loop (coolant pump, steam generator and associated piping) to be shut down for repair while the station operated at reduced output. Connecticut Yankee rarely operated in that fashion. Construction on the new switchgear building, completed in 1990, was begun to meet updated fire protection criteria, and provide enhanced instrumentation and controls for safe plant shutdown. During the fourteenth refueling outage in 1987, additional steam generator tubes were plugged and the low pressure turbines were replaced (see HAER No. CT-185-O-Turbine Building). Containment leak integrity was tested by pressurization. Repairs were made to the attachment devices on the thermal shield surrounding the lower core barrel, probably to reduce flow vibration of the shield.344 Two NRC resident inspectors put in over 4000 hours during the assessment period before and after the shutdown.345 Early in 1989, there was a release of radioactive liquid from the Spent Fuel Building into drainage structures at the nearby 115 kilovolt switchyard, which delivered from other power stations in the system almost all the station service power for start-up, shutdown and power production operation. Clean-up after this event required considerable soil removal. During the fifteenth refueling outage in 1989-90, the thermal shield around the lower core barrel was removed due to damage of the supports. That required the entire core to be transferred to the spent fuel pool.346 In May 1990 specimens of Asiatic clams were found in the service water system of the plant. The species (Corbicula fluminea) spread rapidly in North American fresh waters. Since fouling by these bivalves could comprise important safety systems, Connecticut Yankee was allowed by the Connecticut DEP to continuously chlorinate the system.347 From 1989 though 1994 there were no amendments on tube plugging so it must be assumed that either plant chemistry was under control or leak rates were under limits specified in the CFR. Tube plugging was resumed during the 18th refueling in 1995 along with roll expansion repairs of tubes. This was an older, more labor intensive process in which plugs were rolled into both ends of the tubes.348 During 1996, the last year of generation, additional trip mechanisms were added to control containment high pressure and steam generator blowdowns. The Union of Concerned Scientists (a nuclear watchdog group) claimed that the NRC had found that a critical coolant pipe was undersized and had gone undiscovered for 30 years.349 In October 1996 a gas bubble formed in the Connecticut Yankee reactor, and unnoticed by operators, had flushed out cooling water.350 It is undocumented whether the 19th refueling cycle was completed in advance of plant shutdown in December 1996 . Shutdown of Connecticut Yankee The decision not to seek license renewal and commence decommissioning was based on a study which showed that due to changing market conditions, Connecticut Yankee’s customers would save money if the plant was shut down.351 A review of the physical state of the plant as shown in CY/AEC/NRC documents combined with the economics of light water reactors from the commercial generation of plants can give a picture of what might have caused the Connecticut Yankee directors to decline to ask the NRC for an extension past 2007. License extension was an option that the DOE/NRC encouraged since it gave more



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time to amortize the costs of upgrading the older plants.352 The ongoing steam generator problems might have been a factor in the decision. In its 1988 report on steam generator failures, the NRC worked out “value-impact” models to show what utilities could expect to spend during the remaining life of their plants under the stepped-up inspection plans being implemented because of the unresolved steam generator safety issue. The modeling estimated inspection times, plugging man-hours, occupational radiological exposure (ORE), and replacement power to give a picture of downstream costs. Included in the NRC report was the possibility of partially or completely replacing a steam generator, some of which had been operating for only 10 years. From 1981 to 1993, nine Westinghouse plants and two Combustion Engineering (CE) plants had Steam Generator Replacement Projects (SGRP’s) 353 The first of these replacements, at Virginia Electric Power’s Surry #2, a three-loop Westinghouse plant cost over 200 million dollars and involved 2141 man-rems of radiological exposure. NU spent ten years planning the replacement of two CE generators in its Millstone Unit #2 which had over 3,000 plugged or sleeved tubes.354 In that repair operation, only the bottom portion of the generator containing the tubes was replaced. Improved methods including use of robotics and a full size mock-up building for practicing the procedures resulted in greatly reduced costs and ORE. To facilitate future replacement projects, the NRC in 1989 allowed utilities to undertake them without prior NRC review or approval.355 With the completion of the Millstone SGRP in 1992, NU certainly had the skill sets to assist Connecticut Yankee in a steam generator replacement program. However, it may have been difficult to find lower tube sections that were compatible with the early model Westinghouse generators. If Connecticut Yankee had elected to replace all or part of the generators it would have had to protect them from further corrosive attack. This could have included installing a condensate polishing system, additional air removal devices in the feed water supply and condensate system and further upgrades of the condensers.356 The shutdown and decision to decommission the NU-affiliated Yankee Rowe station in 1992 and Millstone #1 in 1996 provided the company with experience in an alternative to upgrading.357 Another factor in Connecticut Yankee’s decision-making process may have been the ongoing problem of keeping an older plant up to the contemporary NRC safety codes. Years of “back fitting” before and after TMI had left older plants over-complicated and crowded.358 In the final analysis though, national economics alone could have influenced the decision. While the nuclear fuel component of Operating and Maintenance (O&M) costs in 1993 was generally lower than fossil fuel costs, overall O&M for nuclear plants had risen higher than their competition by 1987.359 Only operating efficiencies were going to improve that ratio, and they were going to be hard to come by at a 28-year-old plant that was nearing the end of its operating license. Connecticut Yankee in Retrospect The power industry expected Connecticut Yankee and its contemporary full-scale light-water plants to pave the way for nuclear power to be on par with advanced coal-and oil-burning power stations. In the years during Connecticut Yankee’s first refueling cycles, over forty power reactors were on order. Even before the Three Mile Island accident, the number of orders was falling sharply, however. While fossil-fueled plant orders also dropped off due to a recession in 1974-75, 2/3 of the cancellations were in nuclear plants due to their much higher construction costs.360 Cancellations rose after TMI with the numbers of operating reactors peaking in 1990.361 It is doubtful that a combination of cheap coal, oil and organized protesters could have been the only factors that limited PWR/BWR power production as a percentage of overall U.S. megawatt



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hours. They were hurt by their own design weaknesses and national nuclear policies: poor siting decisions, inefficient heat cycles, technological flaws, their perceived hazard, and a spent fuel liability instead of a credit. There was clearly a gap between the somewhat messianic pronouncements from theoreticians on the necessity of nuclear power no matter what the cost,362 and the more level-headed analysis of utility executives who were hoping that eventually the technology would be profitable for their stockholders.363 While the availability of enriched fuel allowed U.S. firms to build plants with lower capital costs,364 designers had to cut corners in non-safety areas to compete with fossil-fueled plants. Even a critical safety element, the emergency core cooling system pioneered by Rickover may have been shortchanged, as considerable controversy developed as to whether it would function properly in the event of a major loss of coolant.365 The technology of these early reactors was probably flawed from the outset (for private utilities) because of its reliance on military designs and AEC-sponsored forced development.366 The result may have been an over-reaction by the AEC/NRC to the operational problems leading to over-regulation (qualifying every weld) and long licensing entanglements.367 It is surprising the utilities did not avail themselves of other reactor systems that the AEC/NRC had supported. Fort St. Vrain, the now-decommissioned AEC demonstration gas-cooled plant in Colorado, showed that there were alternatives to the water reactors. The system was 5% more efficient, produced high-pressure and high-temperature steam, had low worker rem exposure, and was resistant to loss-of-coolant accidents.368 While Europe and Japan derived a high percentage of their electricity from nuclear power until the 2011 Fukushima disaster, at that time no more than 20% of American electricity was similarly produced.369 A 1999 article in the New York Times compared reactors to Apollo moon rockets- a technology that slipped into history.370 None of the demonstration phase plants and few of the commercial phase ones operated for more than 35 years of their 40 year licenses.371 Many of 800-1000 mw plants that closely followed Connecticut Yankee, and those built under the 1973 AEC standardization program,372 have had or will need steam generator replacements or other upgrades to reach that point. NRC- mandated inspections of all reactor heads after the 2002 Davis-Besse incident undoubtedly further diminished the profitability of 1970s-era plants.373 In spite of these costs, these reactors got a second lease on life because of deregulation in the 1990s. New power entities bought the plants from old-line utilities and applied for license extensions and up-ratings of output.374 Possibly the lessons learned from the problems of the earlier plants will enable their successors to reach and even exceed the 40-year license milestone. In 2005, after a 30-year hiatus, four power companies applied for site approvals for new reactors, all of which are PWR or BWR designs.375 Perhaps reflecting the damage done to the industry by the Three Mile Island accident, the main innovations in these designs appear to be that they will have simplified passive safety systems that do not require backup generators, pumps or operator actions to contain accidents, echoing one of Rickover’s goals at Shippingport.376 The next generation may be just over the border in Canada, in Europe or Japan, and perhaps on CAD drawing boards in designs combining heavy water or carbon moderators, fuel breeding, molten salt, gas or liquid metal cooling, even directly-driven gas turbines377 and intrinsic safety. Despite built-in flaws in some of the primary and secondary systems components, Connecticut Yankee engineers and operators achieved some record performances starting with a 1977 World Light Water Reactor Record Run of 344 days. In 1984 a record 417 day run was achieved followed by a 461 day run in 1989 becoming the first plant to have twice exceeded 400 days.378 In addition it was the first internationally to produce 50 billion and later 60 billion kwh of power. In total, Connecticut Yankee generated over 110 billion



(Page 144) kwh, saving over 58 million tons of coal379 or over 260 million barrels of oil380 during twenty-eight years of operation.

NOTES 1. Hogerton 1968: 30, Fortner 2001: 6. 2. Dietz 1945: 45.

3. Kay 1956: 36.

4. Leprince-Ringuet 1956: 49.

5. Ibid: 51.

6. Stephenson 1954: 44, Leprince-Ringuet 1956: 48.

7. Dietz: 1945: 122.

8. Landis 1955: 768, Hahn: 1958: 77

9. Landis 1955: 768., Nero 1984: 388.

10. Dietz 1945: 128.

11. Stephenson 1954: 46. Hahn 1958: 82.

12. Hahn 1958: 82.


14. Ibid: 26.

15. Ibid : 3; Laurence 1969: 12.

16. Nero and Dennis 1984: 388.

17. Kay 1956: 37, Hahn 1958: 84. Dietz 1945: 134.

18. Weinberg 1994: 13.


20. Ibid: 58.

21. Booth 1969: 8.



(Page 145) 22. Landis 1955: 2.

23. Weinberg 1994: 12.

24. Zinn 1969: 12.

25. Compton 1956: 40.

26. Weinberg 1994: 14.

27. Leverett 1955: 23..

28. Engineering 1959: 336; Koch 1963: 183; Nuclear Management Company 2002:1.

29. Argonne National Laboratory 1998: 4 of 8. Website.


31. Engineering 1947: 308; Macintyre and Bathe 1969: 252.

32. Engineering 1947: 307; Macintyre and Bathe 1974: 252. Hewlett and Duncan 1974: 27.


34. Ibid: 27.




38. Hewlett and Duncan 1974: 72. 39. Stephenson 1954: 80.


41. Ibid: 170.

42. Simpson 1961: 81.

43. Nuclear Engineering 1958b: 435.



46. Janes Fighting Ships 1967-68: 364.




(Page 146) 48. Friend 1964: 613.

49. Ford 1955: 490.

50. Weidenbaum, Sherman and Pashos 1964: 273

51. Oxford English Dictionary 1989: v.17, p. 915 (Thermal. 2b. Nucl. Physics.)

52. Ford 1955: 499, Nero and Dennis 1984: 397..

53. Ford 1955: 500.

54. Nuclear Engineering 1960: 425. 55. Haga and Stevenson 1970: 145.

56. Hogerton 1968: 23.

57. Nero and Dennis: 1984: 388.

58. Hewlett and Duncan 1974: 226; Zinn 1957: 20; Ford 1956: 728.

59. Ford 1956: 728, Zinn, Pittman, and Hogerton 1964:32.

60. Babcock and Wilcox 1960: 27-6; Klehm 1963: 47, Hoffman 1958: 78.

61. Agnew 1981: 61; Power 1982a: 225.

62. Ford 1956: 730.


64. Zinn 1957: 19; Nero 1979: 109.

65. Lewis and Tsui 1960: 24.

66. Duncan and Moll 1981: 6; Engineering 1958a:682; Howe 1976: 25.


68. Engineering 1956b:761, 1958b: 683.

69. Ibid: 1958a: 682.

70. Hewlett and Duncan 1974: 241, 257.


72. Ibid: 682.




(Page 147) 74. Nuclear Engineering 1958: 239.

75. Niland 1959: 55.


77. Stolz 1958: 10.

78. Hewlett and Duncan 1974: 251. Nuclear Engineering 1960: 426.

79. Howe 1976: 42.

80. General Electric Review 1958: 30. Nuclear Engineering 1957:121.



83. Ackerman 1961: 46.


85. Hogerton 1968: 23.

86. Love, Darrow and Randolph 1958: 7.

87. Brower 1958: 73; Yankee Atomic Energy Company 2005; Connecticut Light and Power Company 2005.

88. Haueter 1974: 200.

89. U.S. Department of Energy 2005; Ratical.org 2005: 6.


91. Smith 1960: 476

92. Kilpatrick 1960: 466.

93. Witzke, R. L. and Voysey, Alfred E.- 1958 Westinghouse Engineer. 18, 4: 102-106. July. 94. Setlur 1975: 137.

95. Love, Darrow, and Randolph 1958: 5.

96. Seymore 1992: 7.

97. Babcock and Wilcox 1960: 27-8.

98. Bergstrom 1959: 106.

99. Ibid: 102.



(Page 148) 100. Stern 1964: 248; Klehm 1963: 49.


102. Stern 1964: 247.


104. Stolz 1958: 10,

105. Jackson 1971: 724.

106. Downs, Deegan and Boggus 1959: 2.

107. Hadjian 1973: 251. Geologic and seismic siting criteria take up 9 pages in the 2004 CFR (CFR 2004: 552).


109. Lloyd 1958: 2.

110. Ackerman 1961: 48.

111. Lloyd 1958: 7.

112. Ackerman 1961: 48, Ford 1982: 45.

113. Ackerman 1961: 50.

114. Fortner 2001: 6 of 8.

115. Rice and Wallin 1964: 191.


117. Morabito 1964: 258.

118. Seaborg 1965: 25, Golan 1965: 196, Proceedings of the American Power Conference 1971: 48.

119. Van Nort and Copeland 1975: 213.


121. Nuclear Management Company 2002: 4. Website.

122. Nero and Dennis 1984: 397.

123. United Electric Light and Power Company 1926: 40.

124. Engineering 1926: 285; Baumann 1921: 504.

125. Berry and Moreton 1922: 500.



(Page 149) 126. Power 1922: 762.

127. Hopping 1922: 483.

128. E.g., Connecticut Yankee Atomic Power Company 1966-1974: 8.2-1--8.2-5.


130. Huntley 1955: 767-5.


132. General Electric Review 1954: 5

133. Graham 1965: 210.

134. Ibid: 217.

135. Roddis 1964: 22., Edlund 1963: 128. Haley 1971:239

136. Ritchey 1964: 298.

137. Ibid: 297.

138. Agnew 1981: 63.

139. Hilberry 1957: 27, Lischer 1965: 2, Vennard 1965: 11.

140. Simpson 1961: 82.

141. Gaines 1965: 292.

142. Reichle 1958: 140, Hogerton 1968:29.

143. Risser 1964: 584, 586.

144. Dillard and Baldwin 1964: 167; Dunn 1964: 17.

145. Snouffer and Hanson 1964: 589.

146. Ibid: 589.

147. Bergstrom and Hamming 1965: 87.

148. Jensen 1964: 380.

149. Reichle 1958: 140.

150. Roddis 1964: 22.

151. Seaborg 1965: 24, Roddis 1965: 284.



(Page 150) 152. Engineering 1967: 92.

153. Dillard and Baldwin 1964: 169; Hogerton 1968: 27; Spencer & Miller 1973: 24.

154. Johnson, Fax, and Townsend 1957: 642.

155. Rickert 1964: 268.

156. McIntyre 1975: 21.

157. Dillard and Baldwin 1964: 169; Hogerton 1968: 27; Spencer and Miller 1973:24.

158. Hogerton 1968: 30.

159. Connecticut Yankee Atomic Power Company 1988: 1; Southern California Edison 2002: 2.

160. Haueter 1974: 200.

161. Connecticut Yankee Atomic Power Company 1988; Northeast Utilities 2005.

162. Stern 1964: 246.

163. Ibid 252.

164. Ibid: 247

165. Connecticut Yankee Atomic Power Company UFSAR 1998: 1.2-1.



168. Connecticut Yankee Atomic Power Company 1975.

169. Coe 1958: 75.


171. Ibid: 18.

172. Brower 1958: 74, Coe 1958:75.

173. FDSA 10/70: 4.1-2, 4.2-13.

174. vanNoordennen 2005: Personal correspondence.

175. Connecticut Yankee Atomic Power Company 1998: 1.2-11; Gray 1917: 14.

176. Morgan 1950: 9.




(Page 151) 178. Church 1935: 9.

179. Morgan 1950: 8.

180. Church 1953:10.

181. Sinton 1966: 110., Spencer & Miller 1973: 24.

182. Hogerton 1968: 30.

183. Nero 1979: 21; Power 1982: 212.

184. Power 1982: 212.

185. Sinton 1966: 110.

186. Babcock & Wilcox 1972: 23-1.

187. Johnson & McDonald. 93. 188. Power 1971: 94; MacNaughton 1967: 513.

189. Dillard and Baldwin 1964: 169, Hogerton 1968: 27., Spencer & Miller 1973: 24.


191. Morgan 1950: 7.

192. Johnson 1919.

193. Johnson 1919: 1100.

194. Sinton 1966:112.

195. Bauman 1921: 630.

196. Richardson 1911: 231.

197. Morgan 1950: 11.

198. Ibid.

199. Hossli 1969: 106.,

200. Swann 1966: 73.

201. Swann 1966: 73.

202. Whitaker 1981: 9.

203. Morgan 1950: 11.



(Page 152)

204. Hossli 1969: 110.

205. Morgan 1950: 15.

206. Ibid.

207. Brown and Donahue 1964: 15.

208. Sinton 1966: 113.

209. Ibid: 112


211. Lardner 1836: 105.

212. Nagel 1952: 46.

213. Lardner 1836: 105.

214. Morgan 1950: 12.

215. Johnson 1919: 1116.

216. Connecticut Yankee Atomic Power Company 1987-1995: Ch. 23, p. 12.

217. Oxford English Dictionary (1989 corrected to 1991): XV p. 44; Brown & Campbell 1952: 46.

218. Main 1893: 224.



221. Ibid.

222. Power 1982: 225.

223. Ibid: 226; Penner, ed., 1976 : 71.

224. Morgan 1950: 14.

225. Koester 1908: 250.

226. Connecticut Yankee Atomic Power Company 1987-1995: Ch. 41, p. 26

227. Power 1982: 228.

228. Kinsman 2001:1.

229. Ibid: 1.



(Page 153) 230. Merriman 1970.

231. Passer 1953: 136.

232. Hedges 1892: 165.

233. Ibid: 174.

234. Dawes 1928: 104.

235. Hedges 1892:166.

236. Parsons 1911: 190.

237. Ibid: 22.

238. Lafoon 1950: 21.

239. Ibid: 21.

240. Power 1982: 364.

241. Lafoon 1950: 25.

242. Ibid: 25.

243. Passer 1953: 132.

244. Ibid: 135.

245. Dawes 1902, Vol. 2: 209.

246. Hedges 1892: 174.

247. Snyder 1950: 51.

248. Bowers 1982: 183.

249. Snyder 1950: 53.

250. Ibid: 57.

251. Ibid.

252. Williamson 1973: 15.

253. Spencer & Miller 1973: 25

254. Scott, B. L. Jr. - 1973 Materials Performance at Nuclear Plants. Nuclear Safety 14, 5: 507-512. 255. Connecticut Yankee Atomic Power c1975.



(Page 154) 256. Ibid.

257. Connecticut Yankee Atomic Power Company 1987: PDCR 886: page 7.2-1 of 13.


259. Kalderon 1972: 376., Gray 1972: 389.

260. Swann 1966: 81.

261. Jonas 1985: 9:

262. Chetwin 2002

263. Widay 2002

264. Gruber, H. and Reihard. K., 1979: 11.

265. Ibid: 12


267. Jonas 1985: 10.

268. Ibid: 9.

269. Power 1982: 348.

270. Gray 1972: 379.

271. Ibid: 389

272. Merriman 1970; Clark 2003.


274. U.S. Nuclear Regulatory Commission 1988a: 1-1, 2-37.


276. Coughlin, Stark, Brown, and Johnson 1974: 226.

277. Singley, Welinsky, Whirl, and Klein 1959: 752.



280. Randazza 1975: 761; U.S. Nuclear Regulatory Commission 1995: 10.

281. Babcock and Wilcox 1972: 29-16



(Page 155) 282. Sopocy 1984: 1013.


284. Hopkinson 1982: 24; Randazza 1975: 760.

285. Randazza 1975: 762; Whitaker 1981: 8.

286. Whitaker 1981: 9; Berge and Vignes 1981: 763; Schuecktanz, Riess, and Stieding 1981:757.

287. Randazza 1975: 762; U.S. Nuclear Regulatory Commission 1988a: 2-31; Whitaker 1981: 9; Hopkinson 1982: 24.

288. Noll 1964: 753; Whitaker 1981: 10.

289. Lorenzi 1952: 21.16; Baker 1957: 707.

290. Riedel 1964: 791.

291. Ibid: 791.

292. Mundis and Noble 1981: 751; van Noordennen 2005.

293. Connecticut Yankee Atomic Power Company 1966-1974: 8.7-2; Britt, Millard and DiFilippo 1975: 753.


295. U.S. Nuclear Regulatory Commission 1988a: 2-31; Randazza 1975: 762.

296. Necci 1993: 590; Sopocy 1984: 1016.


298. Singley et al 1959: 755, Huffman and Malinowki 1981: 768.

299. Sopocy and Kovach 1984: 1016


301. Britt, Millard, and DiFillippo 1975: 753.

302. Connecticut Yankee Atomic Power Company 1966-1974: 8.7-2; 1987-1994: Chapter 18, page 9.




306. Ibid: 1-2



(Page 156) 307. Ibid: 4-5.


309. Berge and Vignes 1981: 763.

310. Schuecktanz, Riess, and Stieding 1981:757.

311. Dvorin and Schlesinger 1984: 1005.

312. Huffman and Malinowski 1981: 773.

313. Mundis and Noble 1981: 751; van Noordennen 2005.

314. Power 1982: 228.

315. Trent Tube 2004; Northeast Utilities 1985: 2; Northeast Utilities 1986; Connecticut Yankee Atomic Power Company 1987-1994: Chapter 18, pages 11-12, 19.


317. U.S. Nuclear Regulatory Commission 1995: 1.

318. Wald 2002: A1.

319. Danko and Stahlkopf 1984: 654.


321. Connecticut Yankee 1975-1992: Amendment 7, June 8, 1976.

322. Connecticut Yankee 1975-1992: Amendment #42. October 8, 1981.

323. van Noordennen 2005: Personal correspondence.

324. van Noordennen 2005.




328. U.S. Nuclear Regulatory Commission 1984e: 2.


330. U.S. Nuclear Regulatory Commission 1984c: cover letter signed by Walter A. Paulson.




(Page 157) 332. U.S. Nuclear Regulatory Commission 1984a: cover letter signed by Richard C. DeYoung.

333. U.S. Nuclear Regulatory Commission 1984d.

334. Northeast Utilities n.d.


336. Connecticut Yankee Atomic Power Company 1985: Attachment 1, page 3.

337. Connecticut Yankee Atomic Power Company 1984: Attachment 4, p. 1.

338. Ibid: Attachment 2, p. 10.

339. Connecticut Yankee Atomic Power Company 1987-1994: Chapter 9: page 6.

340. Connecticut Yankee Atomic Power Company 1985: cover letter.

341. U.S. Nuclear Regulatory Commission 1984a: letter, Richard W. Starostecki to W.G. Counsil.

342. Connecticut Yankee Atomic Power Company 1985: Program Plan, p. 4.

343. Connecticut Yankee 1975-1992: Amendment No. 91. 12/31/86.

344. Cf. Connecticut Yankee Atomic Power Company 1987-1994, Chapter 3: page 5.

345. U.S. Nuclear Regulatory Commission 1988: Docket No. 50-213. 1-2. Oct 4.

346. Connecticut Yankee Atomic Power Company 1975-1992; 1987-1994, Chapter 3: page 5; 1989.

347. van Noordennen 2005: Personal Correspondence.

348. Coughlin, Stark, Brown, and Johnson. 1974: 226; Connecticut Yankee Atomic Power Company 1996.

349. Union of Concerned Scientists 2005: Website 3 of 7.

350. Suburbanchicagonews 2005: 3 of 4.

351. Connecticut Yankee Atomic Power Company 2005:1.

352. Clauss, Harrison, and Pickens 1993: 234.

353. Davis, Jacobson and Colburn 1993: 605.

354. Necci 1993: 590.

355. Davis 1993: 609.


357. Davis 1993:609.



(Page 158) 358. Martel and Stahlkopf 1981: 741.

359. Slade 1993: 212.

360. Scientific American 1976: 60A.

361. Wald 1999: 16.

362. Hilberry 1957: 27-28.

363. Vennard 1957: 100; Simpson 1961: 81.

364. Rhoddis 1964: 26.

365. Nero 1979: 89, Ford 1982: 101.

366. Nuclear Engineering 1960: 425, Phung 1984: 889.

367. Ibid: 889; Levenson 1981: 40.

368. Agnew 1981.

369. Nuclear Management Company 2002: 5.

370. Wald 1999: 16.

371. Fortner 2001: 6.

372. Gilbert 1975: 157.

373. Wald 2002: A1.

374. Wald 2000: C1.

375. Wald 2005: A16.

376. Ibid 2005: F3

377. Nero 1979: 11.

378. Connecticut Yankee Atomic Power Company 1988: 2

379. Palo, Emmons, and Gardner 1963: 337. 7,000 tons per day for 600 mw x 300 average power days x 28 years = 58,000,000 tons.

380. Babcock and Wilcox 1960: 2-6. 3.9 bbl equivalent of 1 ton of bituminous coal.



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(Page 179)

Figure 1. HADDAM NECK NUCLEAR POWER PLANT ON DEEP RIVER AND HADDAM U.S.

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AN

D D

AT

ES

SU

MM

AR

Y D

ES

CR

IPT

ION

PL

AN

T S

YS

TE

ML

OC

AT

ION

MA

JOR

EQ

UIP

ME

NT

(u

sual

ly in

seq

uen

ce o

f u

se)

TU

RB

INE

BU

ILD

ING

Mai

n B

uild

ing

exte

rior

: gab

le-e

nd s

teel

-C

IRC

UL

AT

ING

Mai

n G

roun

d F

loor

2 M

ain

Con

dens

ers

(E-2

3-1A

, E-2

3-1B

), e

ach

wit

h 2

HA

ER

No.

CT

-185

-Cfr

amed

str

uctu

re 2

68'x

110'

, 12

2' h

igh

WA

TE

Rin

let &

2 o

utle

t wat

erbo

xes,

and

1 p

rim

ing

tank

1964

-196

6w

ith

met

al

roof

; en

amel

ed

flut

ed

ca.

1978

3 co

ncre

te

sum

p ar

eas

crea

ted

in

1979

-198

2W

aste

wat

er T

reat

men

t F

acil

itie

s ad

ded

alum

inum

sid

ing

exce

pt s

outh

wal

l of

conc

rete

blo

ck &

gla

zed

bric

k; 4

.25'

-hi

gh b

rick

bas

e on

nor

th &

wes

t si

des;

on

wes

t si

de,

4 ar

ched

lo

uver

edal

umin

um p

anel

s ab

ove

bric

k ba

se (

2pa

nels

rem

ovab

le),

& p

aire

d 2.

5'-w

ide

plas

tic

win

dow

se

ctio

ns

at

alte

rnat

eco

lum

ns; o

n ea

st s

ide,

107

' of

cont

inuo

usal

umin

um-s

ash

win

dow

s at

el.

62.

5';

plas

tic w

indo

w p

anel

s in

nor

th a

nd s

outh

pedi

men

tsA

uxili

ary

Bay

ext

erio

r: f

lat-

roof

ed s

teel

-fr

amed

str

uctu

re 2

40-2

68'x

27',

38' h

igh

at s

outh

east

cor

ner

of m

ain

buil

ding

,w

ith

enam

eled

flu

ted

alum

inum

sid

ing

Inte

rior

s: c

ircu

lati

ng w

ater

int

akes

and

disc

harg

es

to

el.

-6,

wit

h di

scha

rge

tunn

el i

nclu

ding

de-

icin

g li

ne i

ntak

e; i

nm

ain

and

auxi

liary

sec

tions

, gro

und

floo

rel

. 21

.5',

mez

zani

ne le

vel e

l. 3

7.5;

mai

nse

ctio

n re

heat

er le

vel e

l. 47

.5',

ope

ratin

gfl

oor

el.

59.

5' w

ith

cran

e ra

il t

op e

l.98

.25'

. F

rom

gro

und

floo

r, r

einf

orce

dco

ncre

te p

edes

tals

sup

port

ed tu

rbin

es &

gene

rato

r on

op

erat

ing

floo

r;

stee

l-fr

amed

m

ezza

nine

&

re

heat

er

leve

lsin

depe

nden

t fr

om

turb

ine-

gene

rato

rfo

unda

tion

s.

Aux

ilia

ry B

ay G

roun

d F

loor

to s

outh

end

of

Mai

n G

roun

d F

loor

Mai

n M

ezza

nine

Vac

uum

Pri

min

g T

ank

(TK

-27-

1A)

Mai

n G

roun

d F

loor

2 V

acuu

m P

rim

ing

Pum

ps (

P-3

6-1A

, P-3

6-1B

)

MA

IN S

TE

AM

Aux

. M

ezza

nine

4 24

"-di

a. M

ain

Ste

am L

ines

& 3

6"-d

ia. M

ain

Ste

amM

anif

old

[sus

pend

ed a

t el.

47.

5'];

2 3

0"-d

ia. M

ain

Ste

am P

ipes

Ope

rati

ng F

loor

Hig

h-P

ress

ure

Tur

bine

(T

G-1

) w

ith

2 tu

rbin

e m

ain

stop

trip

val

ves,

& 2

gov

erno

r co

ntro

l val

ves

on e

ach

mai

n st

op tr

ip v

alve

Mai

n R

ehea

ter

Fl.

4 M

oist

ure

Sep

arat

or R

ehea

ters

(E

-28-

1A th

roug

h E

-28-

1D)

Ope

rati

ng F

loor

2 L

ow-P

ress

ure

Tur

bine

s (T

G-1

A, T

G-1

B)

Mai

n M

ezza

nine

Hig

h-P

ress

ure

Tur

bine

Ste

am D

ump

& V

alve

s

Mai

n R

ehea

ter

Fl.

Low

-Pre

ssur

e T

urbi

ne S

team

Dum

p &

Val

ves

MA

IN G

EN

ER

AT

OR

Ope

rati

ng F

loor

Mai

n G

ener

ator

CO

ND

EN

SA

TE

&M

ain

Gro

und

Flo

or2

Mai

n C

onde

nser

s (E

-23-

1A, E

-23-

1B)

FE

ED

WA

TE

R

Mai

n G

roun

d F

loor

2 C

onde

nsat

e F

eedw

ater

Pum

ps (

P-3

5-1A

, P-3

5-1B

)[s

outh

of

cond

ense

rs]

Mai

n M

ezza

nine

2 P

rim

ing

Air

Eje

ctor

s (E

J-2-

1A/1

B)

& 2

Mai

n A

irE

ject

ors

(EJ-

1-A

, EJ-

1-B

)

Mai

n M

ezza

nine

Gla

nd S

team

Con

dens

er (

E-6

4-1A

)

Mai

n M

ezza

nine

#6A

Low

-Pre

ssur

e F

eedw

ater

Hea

ter

(E-2

2-1A

)[t

hrou

gh M

ain

Con

dens

er A

]

Mai

n M

ezza

nine

#6B

Low

-Pre

ssur

e F

eedw

ater

Hea

ter

(E-2

2-1B

)[t

hrou

gh M

ain

Con

dens

er B

]



MICHAEL RABER

Line

AP

PE

ND

IX A

. S

UM

MA

RY

OF

ST

RU

CT

UR

ES

, PR

IMA

RY

FU

NC

TIO

NS

, AN

D M

AJO

R E

QU

IPM

EN

T

NA

ME

AN

D D

AT

ES

SU

MM

AR

Y D

ES

CR

IPT

ION

PL

AN

T S

YS

TE

ML

OC

AT

ION

MA

JOR

EQ

UIP

ME

NT

(u

sual

ly in

seq

uen

ce o

f u

se)

TU

RB

INE

BU

ILD

ING

M

ain

Mez

zani

ne#5

A L

ow-P

ress

ure

Fee

dwat

er H

eate

r (E

-21-

1A)

HA

ER

No.

CT

-185

-C[t

hrou

gh M

ain

Con

dens

er A

](c

ont)

.

CO

ND

EN

SA

TE

&F

EE

DW

AT

ER

(co

nt.)

Mai

n M

ezza

nine

#5B

Low

-Pre

ssur

e F

eedw

ater

Hea

ter

(E-2

1-1B

)[t

hrou

gh M

ain

Con

dens

er B

]

Ope

rati

ng F

loor

#4A

& 4

B L

ow-P

ress

ure

Fee

dwat

er H

eate

rs (

E-2

0-1A

,E

-20-

1B)

[sou

thea

st a

nd s

outh

wes

t cor

ners

]

Ope

rati

ng F

loor

#3a

& 3

B L

ow-P

ress

ure

Fee

dwat

er H

eate

rs (

E-1

9-1A

,E

-19-

1B)

[nor

th o

f L

P f

eedw

ater

hea

ters

4A

& 4

B]

Aux

. M

ezza

nine

#2a

& 2

B L

ow-P

ress

ure

Fee

dwat

er H

eate

rs (

E-1

8-1A

,E

-18-

1B)

Mai

n G

roun

d F

loor

Fee

dwat

er H

eate

r D

rain

Rec

eive

r T

ank

(TK

-23-

1A)

[nor

th o

f C

onde

nser

A]

Mai

n G

roun

d F

loor

2 F

eedw

ater

Hea

ter

Dra

in P

umps

(P

-33-

1A/1

B)

[nor

thof

TK

-23-

1A]

Aux

. G

roun

d F

loor

2 S

team

Gen

erat

or F

eedw

ater

Pum

ps (

P-3

1-1A

/1B

)

Aux

. M

ezza

nine

#1A

& 1

B H

igh-

Prr

essu

re F

eedw

ater

Hea

ters

(E

-17-

1A/1

B)

[fed

com

mon

hea

der

to 4

Ste

am G

ener

ator

Fee

dL

ines

]

Aux

. M

ezza

nine

4 el

ec.

mot

or-o

pera

ted

valv

es, f

eed

regu

lati

ng v

alve

s &

man

ual i

sola

tion

val

ves

per

Ste

am G

ener

ator

Fee

d L

ine

AU

XIL

IAR

YA

ux.

Mez

zani

neA

uxil

iary

Fee

d B

ypas

s V

alve

sF

EE

DW

AT

ER

SE

RV

ICE

WA

TE

RM

ain

Gro

und

Flo

or2

Mai

n T

urbi

ne L

ube

Oil

Coo

lers

(E

-60-

1A/1

B)

Mai

n G

roun

d F

loor

2 C

lose

d C

ooli

ng W

ater

Sys

tem

Hea

t Exc

hang

ers

(E-7

0-1A

/1B

)

low

er M

ain

4 M

ain

Gen

erat

or E

xcit

er H

ydro

gen

Coo

lers

(E

-62-

1AG

ener

ator

cas

ing

thro

ugh

E-6

2-1D

)

Mai

n G

roun

d F

loor

2

Mai

n G

ener

ator

Sea

l Oil

Coo

lers

(E-6

1-1A

/1B

)

Mai

n M

ezza

nine

2 M

ain

Gen

erat

or I

sola

ted

Pha

se B

us C

oole

rs (

E-4

8-1A

/1B

)



AP

PE

ND

IX A

. S

UM

MA

RY

OF

ST

RU

CT

UR

ES

, PR

IMA

RY

FU

NC

TIO

NS

, AN

D M

AJO

R E

QU

IPM

EN

T

NA

ME

AN

D D

AT

ES

SU

MM

AR

Y D

ES

CR

IPT

ION

PL

AN

T S

YS

TE

ML

OC

AT

ION

MA

JOR

EQ

UIP

ME

NT

(u

sual

ly in

seq

uen

ce o

f u

se)

TU

RB

INE

BU

ILD

ING

S

ER

VIC

E W

AT

ER

Mai

n M

ezza

nine

Mai

n E

xcit

er C

oole

r (E

-115

-1A

)H

AE

R N

o. C

T-1

85-C

(con

t).

CO

NT

RO

L A

IRA

ux.

Gro

und

Flo

or2

Con

trol

Air

Com

pres

sors

(C

-3-1

A/1

B)

Aux

. G

roun

d F

loor

2 C

ontr

ol A

ir R

ecei

vers

(T

K-2

9-1A

/1B

)

Aux

. G

roun

d F

loor

4 C

ontr

ol A

ir F

ilte

rs (

FL

-8-1

A/1

B, F

L-9

-1A

/1B

)

Aux

. G

roun

d F

loor

2 C

ontr

ol A

ir D

ehyd

rato

r (F

L-3

5-1A

/1B

)

Mai

n G

roun

d F

loor

Con

trol

Air

Com

pres

sor

(C-3

-1C

)

Mai

n G

roun

d F

loor

Con

trol

Air

Rec

eive

r (T

K-2

9-1C

)

Mai

n G

roun

d F

loor

2 C

ontr

ol A

ir F

ilte

rs (

FL

-10-

1A/1

B)

Mai

n G

roun

d F

loor

Con

trol

Air

Deh

ydra

tor

(FL

-35-

1C)

SE

RV

ICE

AIR

Aux

. G

roun

d F

loor

Ser

vice

Air

Int

ake

Fil

ters

Aux

. G

roun

d F

loor

2 S

ervi

ce A

ir C

ompr

esso

rs (

C-2

-1A

/1B

)

Aux

. G

roun

d F

loor

2 A

fter

cool

ers

& M

oist

ure

Sep

arat

ors

(F-4

9-1A

/1B

)

Aux

. G

roun

d F

loor

Ser

vice

Air

Rec

eive

r (T

K-2

8-1A

)

WA

TE

RA

ux.

Gro

und

Flo

orW

ell W

ater

Fil

ter

(FL

-44-

1A)

TR

EA

TM

EN

T

Aux

. G

roun

d F

loor

Vac

uum

Dea

erat

or (

D-1

-1A

)

Aux

. G

roun

d F

loor

2 V

acuu

m D

eaer

ator

Pum

ps (

P-3

0-1A

/1B

)

Aux

. Gro

und

Flo

or2

For

war

ding

Pum

ps (

P-3

5-!a

/!b)

Aux

. Gro

und

Flo

orD

emin

eral

ized

Wat

er F

ilte

r (F

L-4

5-1a

)



MICHAEL RABER

Line

MICHAEL RABER

Line

MICHAEL RABER

Line

MICHAEL RABER

Line

MICHAEL RABER

Line

MICHAEL RABER

Line

MICHAEL RABER

Line

MICHAEL RABER

Line

MICHAEL RABER

Line

MICHAEL RABER

Line

AP

PE

ND

IX A

. S

UM

MA

RY

OF

ST

RU

CT

UR

ES

, PR

IMA

RY

FU

NC

TIO

NS

, AN

D M

AJO

R E

QU

IPM

EN

T

NA

ME

AN

D D

AT

ES

SU

MM

AR

Y D

ES

CR

IPT

ION

LO

CA

TIO

NM

AJO

R E

QU

IPM

EN

T (

usu

ally

in s

equ

ence

of

use

)

TE

RR

Y T

UR

BIN

ES

teel

-fra

med

st

ruct

ure

wit

h ga

lbes

tos

PL

AN

T S

YS

TE

M

AU

XIL

IAR

YFE

EDW

ATE

RT

erry

Tur

bine

Bld

g2

Aux

ilia

ry S

team

pow

ered

Ste

am G

ener

ator

Pum

ps

BU

ILD

ING

& N

ON

-si

ding

, 40’

wid

e, 4

3.5'

hig

h &

ext

endi

ng(P

-32-

1A/1

B)

RE

TU

RN

VA

LV

E S

TA

TIO

N11

’ w

est

of t

he R

eact

or C

onta

inm

ent:

Gro

und

Flo

or

el.

21.5

’ co

ntai

ned

2H

AE

R N

o. C

T-1

85-E

19

64-1

966

auxi

liar

y fe

edw

ater

pum

ps.

Rem

aini

ngel

evat

ions

(e

l. 31

’,

41’,

49

’,

57’)

cont

aine

d st

ruct

ural

ste

el u

sed

to s

uppo

rtth

e m

ain

stea

m a

nd f

eedw

ater

sys

tem

pipi

ng

and

valv

es;

non-

retu

rn

valv

est

atio

n at

upp

er le

vel.

Non

-Ret

urn

Val

veN

on-R

etur

n V

alve

s (2

eac

h on

4 s

team

line

s)S

tati

on

DIS

CH

AR

GE

CA

NA

L+

/-70

00'

from

T

urbi

ne

Bui

ldin

gH

AE

R N

o. C

T-1

85-D

disc

harg

e tu

nnel

to

Con

nect

icut

Riv

er;

1965

-196

6up

stre

am +

/- 1

62' i

s st

eel-

brac

ed, t

imbe

r-sh

eet-

pile

-sid

ed f

lum

e en

ding

at

rock

wie

r, d

rops

fro

m e

l.-5.

5' t

o -1

0' i

nto

eart

hen

cana

l 65

'-80'

wid

e at

bot

tom

,13

0'-1

60'-w

ide

at o

uter

end

s of

ber

ms

with

gra

vel o

r ri

p-ra

p in

side

slo

pes;

700

'w

ide

at r

iver

SE

RV

ICE

WA

TE

R;

CIR

CU

LA

TIN

GW

AT

ER

12R

SW

ITC

HY

AR

D34

5 K

V31

9 M

ain

Tra

nsfo

rmer

, 309

Rea

ctor

Coo

lant

Bus

Tra

nsfo

rmer

, mai

n tr

ansf

orm

er s

econ

dary

sid

edi

scon

nect

s, m

ain

tran

sfor

mer

out

put m

otor

-ope

rate

ddi

scon

nect

s, 3

20 L

ine

to 1

4B S

wit

chya

rd

14B

SW

ITC

HY

AR

D34

5 K

Vgr

ound

dis

conn

ect,

man

ual d

isco

nnec

ts, p

ower

cir

cuit

brea

kers

, mot

or-o

pera

ted

disc

onne

cts,

blo

ckho

use

AU

XIL

IAR

Y F

EE

DW

AT

ER

[LA

CK

DE

TA

ILS

]M

AIN

ST

EA

ME

lect

rica

l Aux

ilia

ry S

team

Gen

erat

or F

eed

Pum

p (P

-32-

PU

MP

SK

ID E

NC

LO

SU

RE

S1C

)A

& B


HAER No. CT-185(Page 187)

MICHAEL RABER

Line

AP

PE

ND

IX A

. S

UM

MA

RY

OF

ST

RU

CT

UR

ES

, PR

IMA

RY

FU

NC

TIO

NS

, AN

D M

AJO

R E

QU

IPM

EN

T

NA

ME

AN

D D

AT

ES

SU

MM

AR

Y D

ES

CR

IPT

ION

PL

AN

T S

YS

TE

ML

OC

AT

ION

MA

JOR

EQ

UIP

ME

NT

(u

sual

ly in

seq

uen

ce o

f u

se)

SE

RV

ICE

BU

ILD

ING

Ste

el-f

ram

ed,

conc

rete

-wal

led

stru

ctur

eG

roun

d F

loor

Cab

le S

prea

ding

Are

aH

AE

R N

o. C

T-1

85-F

wit

h ga

lbes

tos

sidi

ng,

304'

x43'

-87.

5',

1964

-196

620

'-55'

hig

h: G

roun

d fl

oor

el.

21.

5'

1981

-198

4co

ncre

te w

alls

for

sei

smic

pro

tect

ion,

incl

udes

1-s

tory

, 20'

-hig

h w

areh

ouse

&m

aint

enan

ce

shop

17

9.5'

x43'

&62

', &

lock

er r

oom

s/ o

ffic

es 1

24.5

'x62

'&87

.5';

Mez

zani

ne e

l. 41

.5' 6

2'x1

03' ;

Ope

ratin

gFl

oor

at e

l. 5

9.5

85'x

103'

, ove

rlyi

ng p

art

of T

urbi

ne B

ldg

Aux

. Bay

.

reno

vati

ons

incl

udin

g ne

w

roof

an

d

rais

ed

floo

r in

C

ontr

ol

Roo

m,

&C

hem

istr

y L

ab

Mez

zani

neS

wit

chge

ar R

oom

A

Ope

rati

ng F

loor

Con

trol

Roo

m

Ope

rati

ng F

loor

Pro

cess

Com

pute

r R

oom

SE

RV

ICE

BO

ILE

R R

OO

M1-

stor

y 24

'-hig

h, 4

5'x3

6.5'

ste

el-

HE

AT

ING

sout

hwes

t cor

ner

of2

oil-

fire

d bo

iler

sH

AE

R N

o. C

T-1

85-F

fram

ed, c

oncr

ete-

wal

led

stru

ctur

e w

ith

Ser

vice

Bui

ldin

g19

64-1

966

galb

esto

s si

ding

; 2 s

ervi

ce d

oors

PR

IMA

RY

AU

XIL

IAR

YR

einf

orce

d-co

ncre

te

stru

ctur

e +

/-P

RIM

AR

YG

roun

d F

loor

2

Pri

mar

y W

ater

Tra

nsfe

r P

umps

(P

-29-

1A/1

B)

BU

ILD

ING

70

x150

', 32

' hi

gh w

ith

1'-t

hick

wal

ls,

WA

TE

R(e

ast e

nd)

HA

ER

No.

CT

-185

-Gco

ncre

te f

loor

s. 2

mai

n le

vels

, eac

h w

ith19

64-1

966

4-to

n m

onor

ail s

yste

ms:

gro

und

floo

r at

el.

21.

5',

seco

nd f

loor

at

el.

35.

5.'

Sec

tion

s ab

ove

seco

nd f

loor

gen

eral

lyst

eel

fram

ed w

ith

insu

late

d G

albe

stos

sidi

ng.

At

east

en

d,

Res

idua

l H

eat

Rem

oval

Pit

, +/-

70x

35.5

' ext

ends

to e

l.-1

9.'

Oth

er s

ecti

ons/

leve

ls i

nclu

de:

3-le

vel

Bor

on R

ecov

ery

Roo

m,

+/-

29.

5'sq

uare

at

els.

35.

5-36

.5',

25.5

' & 1

5.5'

;H

igh-

Pre

ssur

e/L

ow-P

ress

ure

Saf

ety

Inje

ctio

n C

ubic

le

at

el.

15.5

;B

low

dow

n/S

ampl

e &

Non

-Rad

ioac

tive

Val

ve

Roo

m,

+/-

15

x25'

at

el

. 22

';R

adio

acti

ve V

alve

Roo

m +

/- 1

0x19

' at

el.

25'

; C

harg

ing

& M

eter

ing

Pum

pR

oom

s at

el.

15.

5';

Sea

l W

ater

Fil

ter

Cub

icle

at e

l. 13

.4.’

Pip

e ga

lleri

es b

elow

cent

ral l

ongi

tudi

nal a

xis

at e

leva

tion

s of

13-1

4'

Gro

und

Flo

or

2 R

ecyc

led

Pri

mar

y W

ater

Tra

nsfe

r P

umps

(e

ast e

nd)

(P-1

18-1

A/1

B)

SE

RV

ICE

WA

TE

RS

econ

d F

l. (

near

2 C

ompo

nent

Coo

ling

Hea

t Exc

hang

ers

(E-4

-1A

/1B

)W

. end

) th

roug

hfl

oor

Bor

on R

ecov

ery

Bor

on R

ecov

ery

Ove

rhea

d C

onde

nser

(E

-14-

1A)

Roo

m (

uppe

r le

vel)

abov

e S

econ

dS

team

Gen

erat

or B

low

dow

n T

ank

Ven

t Con

dens

er

Flo

or, e

l. 4

0'(E

-78-

1A)

Gro

und

Flo

or2

Ste

am G

ener

ator

Sam

ple

Chi

ller

Con

dens

ers

(C-1

6-1A

/1B

)

Res

idua

l Hea

t2

Res

idua

l Hea

t Rem

oval

Hea

t Exc

hang

ers

Rem

oval

Pit

, (E

-5-1

A/1

B)

el. -

19'



MICHAEL RABER

Line

AP

PE

ND

IX A

. S

UM

MA

RY

OF

ST

RU

CT

UR

ES

, PR

IMA

RY

FU

NC

TIO

NS

, AN

D M

AJO

R E

QU

IPM

EN

T

NA

ME

AN

D D

AT

ES

SU

MM

AR

Y D

ES

CR

IPT

ION

PL

AN

T S

YS

TE

ML

OC

AT

ION

MA

JOR

EQ

UIP

ME

NT

(u

sual

ly in

seq

uen

ce o

f u

se)

PR

IMA

RY

AU

XIL

IAR

YS

econ

d F

loor

2

Ada

ms

Fil

ters

(F

L-5

3-1A

/1B

) B

UIL

DIN

G (

cont

.)H

AE

R N

o. C

T-1

85-G

Bor

on R

ecov

ery

Bor

on R

ecov

ery

Dis

till

ate

Coo

ler

(E-1

5-1A

)R

oom

(lo

wer

leve

l)

Sec

ond

Fl.,

nea

r to

p of

Bor

ic A

cid

Mix

Tan

kB

oric

Aci

d M

ix T

ank

Ven

t Con

dens

er (

E-7

8-1A

)

Pri

mar

y D

rain

s T

ank

Cub

icle

in R

esid

ual

Hea

t Rem

oval

Pit

Pri

mar

y D

rain

s T

ank

Ven

t Con

dens

er (

E-1

1-1A

)

CO

MP

ON

EN

TS

econ

d F

loor

2000

-gal

. C

ompo

nent

Coo

ling

Sur

ge T

ank

(TK

-5-1

A)

CO

OL

ING

(nor

thw

est c

orne

r)W

AT

ER

Gro

und

Flo

or3

Com

pone

nt C

ooli

ng W

ater

pum

ps (

P-1

3-1A

/1B

/1C

)(n

orth

wes

t cor

ner)

[bel

ow C

ompo

nent

Coo

ling

Sur

ge T

ank]

CO

MP

ON

EN

TG

roun

d F

loor

2 C

ompo

nent

Coo

ling

Wat

er H

eat E

xcha

nger

s C

OO

LIN

G(n

orth

wes

t cor

ner)

(E-4

-1A

/1B

) [b

elow

Com

pone

nt C

ooli

ng S

urge

Tan

k]W

AT

ER

BO

RO

NB

oron

Rec

over

y2

Bor

on R

ecov

ery

Was

te L

iqui

d T

rans

fer

Pum

ps

RE

CO

VE

RY

Roo

m, 2

leve

l(P

-22-

1AA

/1B

)nd

Bor

on R

ecov

ery

Bor

on R

ecov

ery

Dis

till

ate

Fee

d H

eat E

xcha

nger

(E

-12-

Roo

m, 2

leve

l1A

) nd

Bor

on R

ecov

ery

Bor

on R

ecov

ery

Fir

st S

tage

Eva

pora

tor

Bot

tom

s P

ump

Roo

m, 1

lev

el(P

-23-

1A)

st

Bor

on R

ecov

ery

Bor

on R

ecov

ery

Fir

st S

tage

Eva

pora

tor

Boi

ler

(E-4

3-R

oom

, 1 l

evel

1A)

st

Bor

on R

ecov

ery

Bor

on R

ecov

ery

Fir

st S

tage

Eva

pora

tor

(EV

-1-A

)R

oom

, 1 l

evel

st

Bor

on R

ecov

ery

Bor

on R

ecov

ery

Sec

ond

Sta

ge E

vapo

rato

r B

otto

ms

Roo

m, 1

lev

elP

ump

(P-2

5-1A

)st

Bor

on R

ecov

ery

Bor

on R

ecov

ery

Sec

ond

Sta

ge E

vapo

rato

r B

oile

r (E

-44-

Roo

m, 1

lev

el1A

)st

Bor

on R

ecov

ery

Bor

on R

ecov

ery

Sec

ond

Sta

ge E

vapo

rato

r (E

V-2

-A)

Roo

m, 2

leve

lnd

SER

VIC

E W

ATE

R

(con

t.)



MICHAEL RABER

Line

AP

PE

ND

IX A

. S

UM

MA

RY

OF

ST

RU

CT

UR

ES

, PR

IMA

RY

FU

NC

TIO

NS

, AN

D M

AJO

R E

QU

IPM

EN

T

NA

ME

AN

D D

AT

ES

SU

MM

AR

Y D

ES

CR

IPT

ION

PL

AN

T S

YS

TE

ML

OC

AT

ION

MA

JOR

EQ

UIP

ME

NT

(u

sual

ly in

seq

uen

ce o

f u

se)

PR

IMA

RY

AU

XIL

IAR

YB

oron

Rec

over

yB

oron

Rec

over

y E

vapo

rato

r O

verh

ead

Con

dens

er (

E-

BU

ILD

ING

(co

nt.)

Roo

m, 3

leve

l14

-1A

)H

AE

R N

o. C

T-1

85-G

rd

Bor

on R

ecov

ery

Bor

on R

ecov

ery

Dis

till

ate

Acc

umul

ator

(T

K-1

8-1A

)R

oom

, 3 le

vel

rd

Bor

on R

ecov

ery

2 B

oron

Rec

over

y D

isti

llat

e P

umps

(P

-26-

1A/1

B)

Roo

m, 2

leve

lnd

Bor

on R

ecov

ery

Bor

ic A

cid

Rec

over

y C

oole

r (E

-16-

1A)

Roo

m, 1

lev

elst

Gro

und

Flo

orL

iqui

d W

aste

Con

trol

Boa

rd

thro

ugh

Sec

ond

Fl.

Bor

ic A

cid

Mix

Tan

k (T

K-2

-1A

)

ST

EA

MB

low

dow

n R

oom

Ste

am G

ener

ator

Blo

wof

f T

ank

(TK

-22-

1A)

GE

NE

RA

TO

RB

LO

WD

OW

N

ST

EA

Mab

ove

Sec

ond

Fl.

2 S

team

Gen

erat

or B

low

off

Tan

k C

onde

nser

s (E

-90-

GE

NE

RA

TO

R1A

/1B

)B

LO

WD

OW

N

Res

idua

l Hea

tB

low

off

Tan

k C

oole

r (E

-91)

[at

el.

-8.6

']R

emov

al P

it

RE

SID

UA

L H

EA

TR

esid

ual H

eat

2 R

esid

ual H

eat R

emov

al P

umps

(P

-14-

1A/1

B)

RE

MO

VA

LR

emov

al P

it[a

t el .

-19'

]

Res

idua

l Hea

t2

Res

idua

l Hea

t Exc

hang

ers

(E-5

-1A

/1B

) R

emov

al P

it[b

etw

een

el. -

5' &

-19

']

CH

EM

ICA

L &

Met

erin

g P

ump

Rea

ctor

Coo

lant

Let

dow

n: N

on-R

egen

erat

ive

Hea

tV

OL

UM

ER

oom

Exc

hang

er (

E-7

6-1A

)C

ON

TR

OL

Sec

ond

Flo

orR

eact

or C

oola

nt L

etdo

wn:

Vol

ume

Con

trol

Tan

k (T

K-

6-1A

)

Gro

und

Flo

orP

urif

icat

ion

Pum

p (P

-12-

1A)

thro

ugh

Sec

ond

Fl.

Bor

ic A

cid

Mix

ing

Tan

k (T

K-2

-1A

)

Gro

und

Flo

or2

Bor

ic A

cid

Pum

ps (

P-9

-1A

/1B

)

BO

RO

NR

ECO

VER

Y(c

ont.)



AP

PE

ND

IX A

. S

UM

MA

RY

OF

ST

RU

CT

UR

ES

, PR

IMA

RY

FU

NC

TIO

NS

, AN

D M

AJO

R E

QU

IPM

EN

T

NA

ME

AN

D D

AT

ES

SU

MM

AR

Y D

ES

CR

IPT

ION

PL

AN

T S

YS

TE

ML

OC

AT

ION

MA

JOR

EQ

UIP

ME

NT

(u

sual

ly in

seq

uen

ce o

f u

se)

PR

IMA

RY

AU

XIL

IAR

YG

roun

d F

loor

Bor

ic A

cid

Fil

ter

(FL

-15-

1A)

BU

ILD

ING

(co

nt.)

HA

ER

No.

CT

-185

-G

Gro

und

Flo

orB

oric

Aci

d B

lend

er (

M-9

-1A

) [a

top

Bor

ic A

cid

Fil

ter]

Gro

und

Flo

orB

oric

Aci

d S

trai

ner

(ST

-6-1

A)

Cha

rgin

g P

ump

Roo

m2

Cen

trif

ugal

Cha

rgin

g P

umps

(P

-18-

1A/1

B)

Met

erin

g P

ump

Roo

mC

harg

ing

Met

erin

g P

ump

(P-1

1-1A

)

Sea

l Wat

er F

ilte

rR

eact

or C

oola

nt P

ump

Sea

l Wat

er I

njec

tion

: 2 S

eal

Cub

icle

, el.

13.4

'W

ater

Sup

ply

Fil

ters

(F

L-5

9-1A

/1B

)

Sea

l Wat

er F

ilte

rR

eact

or C

oola

nt P

ump

Sea

l Wat

er I

njec

tion

: 2 S

eal

Cub

icle

, el.

13.4

'W

ater

Ret

urn

Fil

ters

(F

L-3

6-1A

/1B

)

Res

idua

l Hea

tR

eact

or C

oola

nt P

ump

Sea

l Wat

er I

njec

tion

: Sea

l Wat

erR

emov

al P

itH

eat E

xcha

nger

(E

-45-

1A)

Sec

ond

Flo

orC

hem

ical

Add

itio

n T

ank

(TK

-7-1

A)

CO

NT

AIN

ME

NT

Sec

ond

Flo

or2

Pur

e &

Dil

utio

n A

ir F

ans

(F-5

0A/1

B)

PU

RG

E

EM

ER

GE

NC

YH

igh-

2 L

ow-P

ress

ure

Saf

ety

Inje

ctio

n P

umps

(P

-92-

1A/1

B);

C

OR

E C

OO

LIN

GP

ress

ure/

Low

-2

Hig

h-P

ress

ure

Saf

ety

Inje

ctio

n P

umps

(P

-15-

1A/1

B)

Pre

ssur

e S

afet

yIn

ject

ion

Cub

icle

Cha

rgin

g P

ump

Roo

m2

Cen

trif

ugal

Cha

rgin

g P

umps

(P

-18-

1A/1

B)

[shi

fted

suc

tion

from

Vol

. Con

trol

Tan

k (T

K-6

-1A

) to

Ref

ueli

ng C

avit

y W

ater

Sto

rage

Tan

k (T

K-4

-1A

) ou

tsid

e P

rim

. Aux

. Bld

g.]

CO

NT

AIN

ME

NT

Bor

on R

ecov

ery

2 L

ow-P

ress

ure

Saf

ety

Inje

ctio

n P

umps

(P

-92-

1A/1

B)

SP

RA

YR

oom

, 1 l

evel

st

Res

idua

l Hea

t2

Res

idua

l Hea

t Rem

oval

Pum

ps (

P-1

4-1A

/1B

)R

emov

al P

it

LIQ

UID

WA

ST

ER

esid

ual H

eat

2 A

erat

ed D

rain

s T

anks

(T

K-1

2-1A

/1B

)R

emov

al P

it

Res

idua

l Hea

t2

Aer

ated

Dra

in T

ank

Pum

ps (

P-2

0-1A

/1B

)R

emov

al P

it

CH

EMIC

AL

&V

OLU

ME

CO

NTR

OL

(con

t.)



MICHAEL RABER

Line

AP

PE

ND

IX A

. S

UM

MA

RY

OF

ST

RU

CT

UR

ES

, PR

IMA

RY

FU

NC

TIO

NS

, AN

D M

AJO

R E

QU

IPM

EN

T

NA

ME

AN

D D

AT

ES

SU

MM

AR

Y D

ES

CR

IPT

ION

PL

AN

T S

YS

TE

ML

OC

AT

ION

MA

JOR

EQ

UIP

ME

NT

(u

sual

ly in

seq

uen

ce o

f u

se)

PR

IMA

RY

AU

XIL

IAR

YR

esid

ual H

eat

2 A

erat

ed L

iqui

d W

aste

Str

aine

rs (

ST

-1-1

A/1

B)

BU

ILD

ING

(co

nt.)

Rem

oval

Pit

HA

ER

No.

CT

-185

-G

Gro

und

Flo

or2

Was

te T

est T

ank

Pum

ps (

P-2

8-1A

/1B

)

Dru

mm

ing

Roo

m,

Che

mic

al N

ucle

ar P

roce

ssin

g S

kid

Gro

und

Flo

or

Gro

und

Flo

orL

iqui

d W

aste

Con

trol

Boa

rd

WA

ST

E G

AS

pipe

gal

lery

bel

owV

alve

Ste

m L

eako

ff C

oole

r (E

-85)

Blo

wdo

wn

Roo

m

Pri

mar

y D

rain

sP

rim

ary

Dra

ins

Col

lect

ing

Tan

k (T

K-1

1-1A

)T

ank

Cub

icle

inR

esid

ual H

eat

Rem

oval

Pit

Pri

mar

y D

rain

sP

rim

ary

Dra

ins

Tan

k V

ent C

onde

nser

(E

-11-

1A)

Tan

k C

ubic

le in

Res

idua

l Hea

tR

emov

al P

it

Res

idua

l Hea

t2

Pri

mar

y D

rain

s T

ank

Pum

ps (

P-1

9-1A

/1B

)R

emov

al P

it

Gro

und

Flo

orW

aste

Gas

Con

trol

Boa

rd

DIE

SE

L G

EN

ER

AT

OR

1-st

ory

rein

forc

ed-c

oncr

ete

stru

ctur

e,S

ER

VIC

E3

Die

sel G

ener

ator

sB

UIL

DIN

G25

'x32

', 14

' hi

gh

from

el

. 21

.5',

off

WA

TE

R,

HA

ER

No.

CT

-185

-Hso

uthw

est

corn

er o

f P

rim

ary

Aux

ilia

ryE

ME

RG

EN

CY

1964

-196

6B

uild

ing

GE

NE

RA

TIO

N

NE

W D

IES

EL

GE

NE

RA

TO

R1

stor

y re

info

rced

-con

cret

e/co

ncre

teS

ER

VIC

E2

Die

sel G

ener

ator

s (2

A/2

B),

con

trol

cab

inet

s,B

UIL

DIN

G

bloc

k st

ruct

ure,

91'

x45'

&30

'W

AT

ER

,em

erge

ncy

buse

sH

AE

R N

o. C

T-1

85-I

EM

ER

GE

NC

Y19

69-1

970

GE

NE

RA

TIO

N

LIQ

UID

WA

STE

(con

t.)



MICHAEL RABER

Line

AP

PE

ND

IX A

. S

UM

MA

RY

OF

ST

RU

CT

UR

ES

, PR

IMA

RY

FU

NC

TIO

NS

, AN

D M

AJO

R E

QU

IPM

EN

T

NA

ME

AN

D D

AT

ES

SU

MM

AR

Y D

ES

CR

IPT

ION

PL

AN

T S

YS

TE

ML

OC

AT

ION

MA

JOR

EQ

UIP

ME

NT

(u

sual

ly in

seq

uen

ce o

f u

se)

NE

W &

SP

EN

T F

UE

L48

'x11

7.3'

wit

h gr

ound

flo

or e

l. 2

1.5'

. S

PE

NT

FU

EL

Spe

nt F

uel P

it2

Spe

nt F

uel C

ooli

ng P

umps

(P

-21-

1A/1

B)

BU

ILD

ING

Spen

t Fue

l Pit

48'x

49',

with

top

of 3

3.5'

-C

OO

LIN

Gpu

mp

cubi

cle

HA

ER

No.

CT

-185

-Jde

ep,

35'x

37'

stee

l-li

ned

rein

forc

ed-

1964

-196

6co

ncre

te p

it a

t el

. 4

7' &

ste

el-f

ram

ed,

galb

esto

s-si

ded

stru

ctur

e ab

ove

to e

l.75

.5'.

Pit

has

116

8 5'

-dia

., 14

'-hig

h fu

elca

sks

& a

ski

mm

er

syst

em a

t to

p of

pool

; 6-t

on b

ridg

e cr

ane

abov

e po

ol a

t el.

67.3

'. N

ew F

uel B

uild

ing

48'x

38.2

', 54

'hi

gh a

bove

gro

und

floo

r w

ith r

einf

orce

dco

ncre

te t

o fl

oor

leve

l at

el.

47',

stee

lfr

ame

& g

albe

stos

sid

ing

abov

e; g

roun

dfl

oor

serv

es a

s S

pent

Fue

l P

it p

ump

cubi

cle;

flo

or a

t el

. 3

5' s

uppo

rts

PV

Cra

cks

for

114

1'-d

ia.,

12'-h

igh

fuel

asse

mbl

ies;

3-

ton

brid

ge

cran

e at

el

.67

.3'.

R

einf

orce

d-co

ncre

te

Rad

iati

onC

ontr

ol A

rea

48'x

30',

25.5

' hig

h w

ith

3-to

n br

idge

cra

ne a

t el.3

9.5'

Spe

nt F

uel P

it1

Spe

nt F

uel P

ool T

ube-

Typ

e H

eat E

xcha

nger

(E

-10-

pum

p cu

bicl

e1A

) 1

Spe

nt F

uel P

ool P

late

-Typ

e H

eat E

xcha

nger

(E

-10

-1B

)

Spe

nt F

uel P

it2

Spe

nt F

uel P

ool S

kim

mer

Pum

ps (

P-9

0-1A

/1B

)pu

mp

cubi

cle

Spe

nt F

uel P

it2

Spe

nt F

uel P

ool S

kim

mer

Fil

ters

(F

L-6

5-1A

/1B

)pu

mp

cubi

cle

SE

RV

ICE

WA

TE

RS

pent

Fue

l Pit

2 S

pent

Fue

l Poo

l Hea

t Exc

hang

ers

(E-1

0-1A

/1B

)pu

mp

cubi

cle

RE

FU

EL

ING

Spe

nt F

uel P

it e

l. F

uel E

leva

tor

(FU

-5);

Slu

ice

Gat

e (F

U-7

)49

.5'

ION

EX

CH

AN

GE

AR

EA

2 ad

jace

nt

rein

forc

ed-

conc

rete

Was

te L

iqui

d Io

n E

xcha

nger

(I-

6-1A

)H

AE

R N

o. C

T-1

85-K

stru

ctur

es:

Ion

Exc

hang

e S

truc

ture

,19

64-1

967

21.8

'x73

.5',

18' h

igh

from

el.

14.

6', w

ithde

ck

at

el.

22

.5'

in

fron

t of

io

nex

chan

ger

&

dem

iner

aliz

er

faci

liti

esw

hich

are

arr

ayed

wit

hin

Ion

Exc

hang

eS

truc

ture

in

17'-h

igh

cham

bers

; S

pent

Res

in S

tora

ge P

it,

13.3

'x17

.3',

17' h

igh

from

el.

22' w

ith c

ham

ber

for

rem

ovab

leli

ner,

ove

r pi

t ex

tend

ing

to e

l. 7

.5';

with

in p

it, Io

n E

xcha

nge

Sum

p Pu

mp

(P-

63-1

A)

to e

l. 2

.5'.

BO

RO

N R

EC

OV

ER

YIo

n E

xcha

nge

Str

uct.

Ion

Exc

hang

e S

truc

t.W

aste

Liq

uid

Tra

nsfe

r F

ilte

r (F

L-1

3-1A

)

WA

ST

E L

IQU

IDA

erat

ed D

rain

s D

emin

eral

izer

(I-

3-1a

)Io

n E

xcha

nge

Str

uct.

Ion

Exc

hang

e S

truc

t.A

erat

e D

rain

s/S

pent

Fue

l Pit

Fil

ter

(FL

-3-1

A/F

L-6

5)

CH

EM

ICA

L &

Rea

ctor

Coo

lant

Let

dow

n: 2

Pur

ific

atio

n D

emin

eral

izer

VO

LU

ME

Ion

Exc

hang

ers

(I-1

-1A

/1B

)C

ON

TR

OL

Ion

Exc

hang

e S

truc

t.

Ion

Exc

hang

e S

truc

t.R

eact

or C

oola

nt L

etdo

wn:

Rea

ctor

Coo

lant

Pre

-fil

ter

(FL

-5-1

A)

Ion

Exc

hang

e S

truc

t.R

eact

or C

oola

nt L

etdo

wn:

Rea

ctor

Coo

lant

Pos

t-F

ilte

r(F

L-1

1-1A

)

Ion

Exc

hang

e S

truc

t.P

urif

icat

ion:

Deb

oron

atin

g Io

n E

xcha

nger

(I-

2-1A

)

SP

EN

T F

UE

LS

pent

Fue

l Poo

l Ion

Exc

hang

er (

I-1-

1C)

CO

OL

ING

Ion

Exc

hang

e S

truc

t.

Ion

Exc

hang

e S

truc

t.A

erat

e D

rain

s/S

pent

Fue

l Pit

Fil

ter

(FL

-3-1

A/F

L-6

5)



MICHAEL RABER

Line

MICHAEL RABER

Line

MICHAEL RABER

Line

MICHAEL RABER

Line

MICHAEL RABER

Line

MICHAEL RABER

Line

AP

PE

ND

IX A

. S

UM

MA

RY

OF

ST

RU

CT

UR

ES

, PR

IMA

RY

FU

NC

TIO

NS

, AN

D M

AJO

R E

QU

IPM

EN

T

NA

ME

AN

D D

AT

ES

SU

MM

AR

Y D

ES

CR

IPT

ION

PL

AN

T S

YS

TE

ML

OC

AT

ION

MA

JOR

EQ

UIP

ME

NT

(u

sual

ly in

seq

uen

ce o

f u

se)

ION

EX

CH

AN

GE

rein

forc

ed

conc

rete

ad

diti

on

to

Ion

WA

ST

E G

AS

Ion

Exc

hang

e A

dd.

Deg

asif

ier

Pre

-Fil

ter

(FL

-67)

ST

RU

CT

UR

E A

DD

ITIO

NE

xcha

nge

Stru

ctur

e, 2

1'x3

0', 1

0'-2

0' h

igh

1973

-197

4fr

om e

l. 12

.8',

wit

h 4

17'-h

igh

cham

bers

WA

ST

E L

IQU

ID

Ion

Exc

hang

e A

dd.

Was

te L

iqui

d P

olis

hing

Dem

iner

aliz

er (

I-9)

Ion

Exc

hang

e A

dd.

Was

te E

vapo

rato

r D

isti

llat

e F

ilte

r (F

l-69

)

WA

ST

E L

IQU

IDS

pent

Res

inS

pent

Res

in S

tora

ge T

ank

(TK

-102

-1A

)S

OL

ID W

AS

TE

Sto

rage

Pit

Spe

nt R

esin

Tra

nsfe

r P

ump

(P-1

56-1

A)

BO

RO

NIo

n E

xcha

nge

Add

.B

oron

Eva

pora

tor

Dis

till

ate

Fil

ter

(FL

-68)

RE

CO

VE

RY

Ion

Exc

hang

e A

dd.

Bor

on R

ecov

ery

Pol

ishi

ng D

emin

eral

izer

(I-

8-1a

)

WA

ST

E D

ISP

OS

AL

41'x

42'

rein

forc

ed

conc

rete

st

ruct

ure,

WA

ST

E G

AS

Deg

asif

ier

Pre

heat

er (

E-8

6)B

UIL

DIN

G55

.5' h

igh

wit

h fl

oors

at

el.

0',

18.5

' &H

AE

R N

o. C

T-1

85-L

21.5

', &

35.

5'

1973

-197

4

thro

ugh

mid

dle

leve

l

thro

ugh

uppe

r le

vel

Deg

asif

ier

(TK

-58-

1A)

wit

h D

egas

ifie

r V

ent C

oole

r (E

-89

) &

Deg

asif

ier

Ven

t Con

dens

er (

E-8

7)

mid

dle

leve

l2

Deg

asif

ier

Liq

uid

Tra

nsfe

r P

umps

(P

-106

-1A

/1B

)

mid

dle

leve

l - e

l. 30

'D

egas

ifie

r E

fflu

ent C

oole

r (E

-88)

uppe

r le

vel -

el.

50'

Deg

asif

ier

Ven

t Coo

ler

(E-8

9)

low

er le

vel

Was

te G

as S

urge

Tan

k (T

K-5

9-1A

)

uppe

r le

vel -

el.

38.

3'2

Was

te G

as C

ompr

esso

rs (

C-1

3-1A

/1B

)

low

er le

vel

3 W

aste

Gas

Dec

ay T

anks

(T

K-6

0-1A

/1B

/1C

)

Was

te G

as S

ampl

e &

Rel

ease

Hea

der

WA

ST

E L

IQU

IDW

aste

Eva

pora

tor

Fee

d D

isti

llat

e E

xcha

nger

(E

-96)

mid

dle

leve

l - e

l. 3

0'

low

er le

vel

Was

te E

vapo

rato

r R

eboi

ler

Pum

p (P

-114

-1A

)

low

er le

vel

Was

te E

vapo

rato

r R

eboi

ler

(E-9

2)

mid

dle/

uppe

r le

vels

Was

te L

iqui

d E

vapo

rato

r (E

V-4

)

mid

dle

leve

lW

aste

Eva

pora

tor

Dis

till

ate

Tan

k P

ump

(P-1

15-1

A)

uppe

r le

vel -

el.

47.

3'W

aste

Eva

pora

tor

Ove

rhea

d C

onde

nser

(E

-93)



MICHAEL RABER

Line

MICHAEL RABER

Line

MICHAEL RABER

Line

MICHAEL RABER

Line

AP

PE

ND

IX A

. S

UM

MA

RY

OF

ST

RU

CT

UR

ES

, PR

IMA

RY

FU

NC

TIO

NS

, AN

D M

AJO

R E

QU

IPM

EN

T

NA

ME

AN

D D

AT

ES

SU

MM

AR

Y D

ES

CR

IPT

ION

PL

AN

T S

YS

TE

ML

OC

AT

ION

MA

JOR

EQ

UIP

ME

NT

(u

sual

ly in

seq

uen

ce o

f u

se)

WA

ST

E D

ISP

OS

AL

uppe

r le

vel

Was

te E

vapo

rato

r D

isti

llat

e T

ank

(TK

-64)

BU

ILD

ING

(co

nt.)

HA

ER

No.

CT

-185

-L

el. 3

0.’

/upp

er le

vel

Was

te E

vapo

rato

r D

isti

llat

e C

oole

r (E

-94)

low

er le

vel

Was

te E

vapo

rato

r B

otto

ms

Pum

p (P

-116

-1A

)

low

er le

vel

Was

te E

vapo

rato

r B

otto

ms

Coo

ler

Pre

heat

er (

E-9

7)

low

er le

vel

Was

te E

vapo

rato

r B

otto

ms

Coo

ler

(E-9

5)

low

er le

vel

Was

te E

vapo

rato

r B

otto

ms

Coo

ler

Cir

cula

ting

Pum

p (P

-12

0-1A

)

uppe

r le

vel

Flo

or &

Equ

ipm

ent D

rain

Tan

k (T

K-6

5)

low

er le

vel

2 F

loor

Dra

in T

ank

Pum

ps (

P-1

19-1

A/1

B)

low

er le

vel

2 E

quip

men

t Dra

in T

ank

Pum

ps (

P-1

21-1

A/1

B)

CO

MP

ON

EN

Tth

roug

h up

per

leve

lD

egas

ifie

r (T

K-5

8-1A

) w

ith

Deg

asif

ier

Ven

t Coo

ler

(E-

CO

OL

ING

89)

& D

egas

ifie

r V

ent C

onde

nser

(E

-87)

WA

TE

R

CO

MP

ON

EN

Tup

per

leve

l 2

Was

te G

as C

ompr

esso

rs (

C-1

3-1A

/1B

)C

OO

LIN

G-

el.

38.3

'W

AT

ER

mid

dle/

uppe

r le

vels

Was

te L

iqui

d E

vapo

rato

r (E

V-4

)

SE

RV

ICE

WA

TE

Rm

iddl

e/up

per

leve

lsW

aste

Liq

uid

Eva

pora

tor

(EV

-4)

thro

ugh

uppe

r le

vel

Deg

asif

ier

(TK

-58-

1A)

wit

h D

egas

ifie

r V

ent C

oole

r (E

-89

) &

Deg

asif

ier

Ven

t Con

dens

er (

E-8

7)

1-st

ory,

met

al-f

ram

ed, m

etal

-sid

edS

OL

ID W

AS

TE

sout

h of

Spe

nt F

uel

com

pact

ors,

cle

anin

g fa

cili

ty, s

tora

ge a

rea

for

soli

dst

ruct

ure,

±88

'x61

'&37

'B

uild

ing

radi

oact

ive

and

mix

ed w

aste

RA

DW

AS

TE

RE

DU

CT

ION

F

AC

ILIT

YH

AE

R N

o. C

T-1

85-N

WA

STE

LIQ

UID

(con

t.)



AP

PE

ND

IX A

. S

UM

MA

RY

OF

ST

RU

CT

UR

ES

, PR

IMA

RY

FU

NC

TIO

NS

, AN

D M

AJO

R E

QU

IPM

EN

T

NA

ME

AN

D D

AT

ES

SU

MM

AR

Y D

ES

CR

IPT

ION

PL

AN

T S

YS

TE

ML

OC

AT

ION

MA

JOR

EQ

UIP

ME

NT

(u

sual

ly in

seq

uen

ce o

f u

se)

SP

EN

T R

ES

IN S

TO

RA

GE

Par

t of

ori

gina

l S

pent

Res

in S

tora

ge P

itS

OL

ID W

AS

TE

Rem

ovab

le s

tora

ge li

ners

in s

teel

cel

ls;

FA

CIL

ITY

rem

oved

, re

plac

ed

to

nort

heas

t w

ith

WA

ST

E L

IQU

ID3"

PV

C p

ipe

drai

ns f

rom

cel

l bot

tom

s in

to a

djac

ent

HA

ER

No.

CT

-185

-M22

.2'x

29.2

' le

ad-l

ined

re

info

rced

-S

pent

Res

in S

tora

ge P

it19

79-1

980

conc

rete

str

uctu

re w

ith

fl.

el.

19.

2',

10.8

' hig

h w

ith 2

3.5'

-hig

h w

alls

on

nort

h&

eas

t si

des,

con

tain

ing

11 5

.2'-d

ia.,

10.8

'-hig

h st

eel c

ells

PR

IMA

RY

WA

TE

R30

'-OD

, 150

,000

-gal

. st

eel t

ank

PR

IMA

RY

east

of

New

&S

TO

RA

GE

TA

NK

W

AT

ER

Spe

nt F

uel B

ldg.

&(T

K-2

0-1A

) c1

964-

1966

yard

cra

ne

RE

CY

CL

E P

RIM

AR

Y22

'-OD

150

,000

-gal

. st

eel t

ank

PR

IMA

RY

east

of

Rad

was

teW

AT

ER

ST

OR

AG

E T

AN

KW

AT

ER

Red

ucti

onF

acil

ity

(TK

-62-

1A)

1973

DE

MIN

ER

AL

IZE

D W

AT

ER

25'-I

D 1

00,0

00-g

al. s

teel

tank

A

UX

ILIA

RY

in d

iked

enc

losu

reS

TO

RA

GE

TA

NK

F

EE

DW

AT

ER

SW

of

Rea

ctor

(TK

-25-

1A)

c196

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AL

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ER

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NC

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h of

Ser

vice

inst

rum

enta

tion

and

con

trol

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saf

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ant s

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HA

ER

No.

CT

-185

-O

stru

ctur

e 34

'x64

', 37

.5'h

igh

abov

eB

uild

ing

1987

-199

0gr

ound

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el.

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'; ba

sem

ent l

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ea e

l. 1

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cir

cuit

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aker

, 2 tr

ansf

orm

ers

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h li

ghtn

ing

HA

ER

No.

CT

-185

-PR

eact

orar

rest

ors,

man

uall

y- a

nd D

C-m

otor

-ope

rate

d di

scon

nect

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IST

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side

has

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30'

-wid

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nels

w

ith

glas

s an

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bric

k in

fill

& d

oors

in n

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arc

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d in

sim

ilar

sty

le to

sou

th e

nd, w

ith

wes

t fa

ce

exte

nded

no

rth

past

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uth

arch

ed p

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INF

OR

MA

TIO

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o. C

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and

196

4-19

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etal

pan

els

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ovat

ions

HE

AL

TH

PH

YS

ICS

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ory

conc

rete

str

uctu

re, ±

35'x

51.5

',co

unt r

oom

, pro

cess

ing,

off

ice

FA

CIL

ITY

±30.

5' h

igh

HA

ER

No.

CT

-185

-T19

64-1

966

HE

AL

TH

PH

YS

ICS

±20'

x17'

of

fice

PR

OJE

CT

TR

AIL

ER



MICHAEL RABER

Line

AP

PE

ND

IX A

. S

UM

MA

RY

OF

ST

RU

CT

UR

ES

, PR

IMA

RY

FU

NC

TIO

NS

, AN

D M

AJO

R E

QU

IPM

EN

T

NA

ME

AN

D D

AT

ES

SU

MM

AR

Y D

ES

CR

IPT

ION

PL

AN

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sual

ly in

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ce o

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se)

HE

AL

TH

PH

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ICS

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stor

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5'x1

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OD

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1 19

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fram

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etal

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able

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fst

ruct

ure

120'

x122

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20' h

igh

WA

RE

HO

US

E #

2 1

994

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al-s

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str

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re 1

20'x

145,

24'

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ST

EA

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dia

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FF

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1985

stru

ctur

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com

pris

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of

seve

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inte

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ed m

odul

es o

f of

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spa

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r en

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ers;

co

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ted

to

the

Inst

rum

ent

and

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trol

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uild

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c19

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of

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pace

for

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rum

enta

tion

an

d C

ontr

ols

Gro

up a

nd O

pera

tions

; 2 f

loor

incl

uded

nd

a lu

nch

room

/ con

fere

nce

room

use

d fo

rda

ily

plan

t op

erat

ions

m

eeti

ng;

conn

ecte

d to

the

Eng

inee

ring

Mod

ular

.

UN

CO

ND

ITIO

NA

LS

mal

l 1-

stor

y st

eel-

fram

ed

buil

ding

proc

essi

ng a

rea

to r

elea

se e

quip

men

t and

tool

s fo

rR

EL

EA

SE

FA

CIL

ITY

loca

ted

sout

h of

Rea

ctor

Con

tain

men

t ra

dwas

te c

ontr

ol a

rea

TR

AIN

ING

& S

TO

RE

Sm

etal

-fra

med

, met

al-s

ided

, gab

le-

OF

FIC

E c

1978

-197

9ro

ofed

str

uctu

re 9

6'x5

0', ±

20' h

igh

EM

ER

GE

NC

Y1-

stor

y re

info

rced

-con

cret

e st

ruct

ure

OP

ER

AT

ION

S F

AC

ILIT

Y12

6'x9

6', 1

4'-1

7' h

igh.

HA

ER

No.

CT

-185

-W 1

980-

1981

WA

RE

HO

US

ES

A A

ND

B2

1-st

ory

stee

l-fr

amed

met

al-s

ided

1982

stru

ctur

es, e

ach

±100

'x46

'

OF

FIC

E B

UIL

DIN

G #

32-

stor

y, g

able

-roo

fed,

ste

el-f

ram

ed &

c198

9-19

94co

ncre

te-b

lock

str

uctu

re 1

00'x

120'



MICHAEL RABER

Line

MICHAEL RABER

Line

MICHAEL RABER

Line

MICHAEL RABER

Line

MICHAEL RABER

Line

MICHAEL RABER

Line

written historical and descriptive data reduced

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