engineering change notice 1 of 2- w. ecn/67531/metadc734491/... · record of revision i i page i 2)...

...................................... ENGINEERING CHANGE NOTICE w. ECN

P..a. 1 O f 2-

ECN Category (mark one) Supplemental 0 J. C. Elgin, SNFP Radiological Control, DirectRevision x3-68, 372-0905

3. Originatots Name, Organization, MSIN, and Telephone No. 4. USQ Required? 5. Date

N Y e s U N O 09/26/2000

0 6. Project TitielNo.fWork Order No. 7. Bidg./Sys./Fac. No. Change ECN

8. Approval Designator

4a. Justification (mark one)

Criteria Chanae IXI

Temporaly 0 Standby

Supersedure 0

- Design Improvement

Environmental 0

N/A N/A N/A 9. Document Numbers Changed by this ECN (includes 10. Related ECN No($ 11. Related PO No.

sheet no. and rev.)

14b. Justification Details This document serves as technical basis for the radiation and contamination trending program for SNFP and change needed to update and maintain current.

Facility Deactivation 0 As-Found 0 Facilitate Const. 0 Const. ErrorIOmission 0

CancelNoid 0

0 Yes (fill out Blk. 12b) No NA Blks. 12b. IXI 42c.12d)

2a. Modification Work

Design ErrorIOmission 0 15. Distribution (include name, MSIN, and no. of Wpies) 3. T. Southerland, 52-44, 1 3. E. Elder, 52-44, 1 P. G. Huntley, X3-68, 1 J. E. Kurtz, X3-68, 1 T . E. Bratvold, X3-68, 1 J. C. Elgin, X3-68, 1 D. J. Watson, X3-79, 1 3. M. Seranno, 52-44, 1

SNF-4358, Revision 1 659750 N/A 12b. Work Package No. 12c. Modification Work Completed 12d. Restored to Ori inai Condition (Temp.

or Standby ECds only)

N/A N/A Design AuthoritylCo Engineer Signature & Design AuthoritylCo Engineer Signature &

d i e B i t e N/A

A-7900-0131 A-7900-013-2 (10197)

- .- -I

of 2 ENGINEERING CHANGE NOTICE

1. ECN(useno frompg 1)

6 6 0 6 0 1

SDDIDD 0 Functional Design Criteria 0 Operating Specification 0 Criticality Specification 0 Conceptual Design Report 0 Equipment Spec. 0 Const. Spec. 0 Procurement Spec. 0 Vendor Information 0 OM Manual 0 FSARISAR 0 Safety Equipment List 0 Radiation Work Permit 0

Environmental Report 0 Environmental Impact Statement

I ~~ ~

6. Design Verification 17. Cost Impact Required ENGINEERING CONSTRUCTION

0 yes Additional $ N/A Additional 0 $ N/A No Savings $ N / A Savings 0 $ N/A

SeismidStress Analysis

StresslDesign Report

Interface Control Drawing

Calibration Procedure Installation Procedure

Maintenance Procedure

Engineering Procedure

Operating Instruction

Operating Procedure

Operational Safety Requirement

IEFD Drawing

Cell Arrangement Drawing

Essential Material specification

Fac. Proc. Samp. Schedule

Inspection Plan

18. Schedule Impact (days)

Improvement 0 N/A Delay 0 N/A

Tank Calibration Manual 0 0 Health Physics Procedure €4 0 Spares Multiple Unit Listing 0

0 Component Index 0 0 ASME Coded Item c7

0 Computer Software 0 Electric Circuit Schedule 0

0 ICRS Procedure 0

0 Process Flow Chart 0 0 Purchase Requisition 0 0 Tickler File 0 0 0

0 Test ProceduredSpecification 0

0 Human Factor Consideration 0

0 Process Control ManualIPlan 0

Environmental Permit 0 Inventory Adjustment Request 0 0 !O. 0ther.Affected Documents:. (NOTE: Documents listed below will not be revised by this ECN.) Signatures below indicate that the signing

organization has been notified of other affected documents listed below. Document NumberlRevision Document NumberlRevision Document NumberlRevision

!I. Approvals Signature Date

Design Authority

QA

Safety

Environ. D. J. Watson p - 2 & c;u / -

Other

Basins RC Manager T. E. Bratval

CVD RC Manager J. E . Kur

CSB RC Manaqer D. T. S O U

Signature Date

Design Agent

PE

QA

Safety

Design

Environ.

Other

QFPARTMFNT OF ENFRGY

Signature or a Control Number that tracks the Approval Signature

ADDITIONAL

SNF-4358 Revision 2

Technical Basis - Spent Nuclear Fuels Project Radiation and Contamination Trending Program

Prepared for the US. Department of Energy Assistant Secretary for Environmental Management Project Hanford Management Contractor for the U S . Department of Energy under Contract DE-AC06-96RL13200

Fluor Hanf ord P.O. Box 1000 Richland, Washington

Approved for public release; further dissemination unlimited

_ _ _ ~ _-

SNF-4358 Revision 2

ECN - 660601

Technical Basis - Spent Nuclear Fuels Project Radiation and Contamination Trending Program

Document Type: TR Division: SNF

J. C. Elgin Fluor Hanford, Inc,

Date Published

September 2000

Prepared for the US. Department of Energy Assistant Secretary for Environmental Management Project Hanford Management Contractor for the U.S. Department of Energy under Contract DE-AC06-96RL13200

Fluor Hanford P.O. Box 1000 Richland, Washington

/" -2-zou Release Approval Date

I STA.

I STA.

Release Stamp

Approved for public release: further dissemination unlimited

..

SNF-4358 Revision 2

1

TRADEMARK DISCLAIMER Reference nerein to any specifc commercia proo-ct. process or sewice oy trade name, trademark. manufacturer or otheruise, does not necessarily constitute or imply its endorsement, recommendation. or favoring by the United States Government or any agency thereof or its contractors or subcontractors.

This report has been reproduced from the best available copy.

Piinled in (ha United SlLes of A m e b

Total Pages: 5c3

RECORD OF REVISION I I Page I

2) Title Cechnical Basis - Spent Nuclear Fuel Project Radiation and Contamination Trending Program

Change Control Record

(1) Document Number

SNF-4358, EDT-624930

(3) Revision

Revision 1, ECN-659750, issued 5/3/2000 N/A

0

N/A I

2

2

2

2

2

2

2

Rs'

Authorized for Release

(5 ) Cog. Engr. I (6) Cog. Mgr. Date (4) Desaiption of Change - Replace, Add, and Delete Pages 17) I I ~I Initial issue, revision 0, EDT-624930

Updated Title pages to reflect revision number change Page 2-5, updated Table of Contents

Page 7, table 3.1, added building 212H to facilities list Pages 9-10. Section 4.1.1, updated area monitoring dosimetry information and data chart Pages 28-35, added section 4.4 for CSB

Pages 51-53, added section 9.4 for CSB

Page 54, updated section 11 to include references for CSB , ECn/ -66060 /

A-7320-005 (10/97)

Revision 2 SNF-4358 Page 1

TECHNICAL BASIS

SPENT NUCLEAR FUEL PROJECT RADIATION AND CONTAMINATION TRENDING PROGRAM


Table of Contents

1 .O Introduction Page 6 2.0 Background Page 6 3 .O Facility Management Responsibilities Page 6

Page 7 4.0 Facility DescriptiondHistory and Radiological Status Page 7

Table 3.1 : SNFP Facility Management Responsibilities

4.1

4.2

4.3

KE Area 4.1.1

4.1.2 4.1.3 4.1.4

4.1.5 4.1.6

4.1.7 4.1.8 4.1.9

4.1.11 4.1.12 4.1.13 4.1.14 4.1.15 4.1.16

4.1.17 KW Area 4.2.1 4.2.2 4.2.3 4.2.4 4.2.5 4.2.6

4.i.10

Page 7 KE Basin Page 7 Figure 4.1 : 4/94 Grid Survey Page 9 Figure 4.2: 9/96 Grid Survey Page 10 Figure 4.3: Area Monitoring Dosimetry Page 10

119-KE: Exhaust Air Sample Building Page 11 1614-KE: Environmental Monitoring Station Page 11 165-KE: Power Control Building Elect switchgear, Page 11 Control Room 18 1 -KE: River Pump House Page 11 183-KE: Clearwells, Filters, Sedimentation Basins, Page 11 Headhouse, and Chlorine Vault 185-K Package Water Treatment Plant Page 12 190-KE: Main Pump House Page 12 1706-KE: Water Studies Semi-works Facility Page 12 1706-KEL: Development Laboratory Page 12 1706-KER: Water Studies Recirculation Building Page 12 1713-KE: Shop Building Page 13 1714-KE: Oil & Paint Shop Page 13 1717-Wl724K: Maintenance Shops Page 13 17 17-AKE: Fan House (Also referred to as 167K) Page 13 1908 KE & 1908 K: Outfall Instrumentation Page 13

Other KE Structures Page 13 Page 13

KW Basin Page 13 165-KW: switchgear, Control Room Page 14 183-KW: Chlorine Vault Page 14 1713-KW: WarehouseiShop Page 14 1714-KW: Oilpaint Storage Page 14 Other KW Structures Page 14

Projections

BuildingIOutfall Structure

Cold Vacuum Drying Facility Page 15 4.3.1 Cold Vacuum Drying Facility Description Page 15 4.3.1.1 Cold Vacuum Drying Operation Overview Page 15 4.3.2 Facility Classification Page 17 4.3.3 Expected Source Term Page 17

Page 18 Page 19

Table 4-1: Radionuclide Concentrations Table 4-2: Source Material Concentration

in Components


Table of Contents: Continued

4.3.3.1

4.3.3.1.a 4.3.3.2 4.3.4 Radiological Posting of CVD 4.3.4.1 CVD Siting 4.3.4.2 CVD Facility Administrative Building 4.3.4.3 Access Corridor 4.3.4.4 Process Bays 4.3.4.5 Process Water Conditioning Room 4.3.4.6 Mechanical Room 4.3.4.7 Health Physics Counting Room 4.3.4.8 Decontamination Room 4.3.4.9 Unintermptable Power Supply Room 4.3.4.10 Security Room 4.3.4.1 1 AidDemineralized Water Room

Methodology and Expected Dose Rates Table 4-3: Calculated Dose Rates in Process Bay

Process Water Conditioning Room Dose Rates Discussion Of expected Dose Rates

4.4 Canister Storage Building 4.4.1 4.4.2 4.4.3 4.4.4 4.4.5 4.4.5.1 4.4.5.2 4.4.5.2.1 4.4.5.2.2 4.4.5.2.3 4.4.5.2.4 4.4.5.2.5 4.4.5.3 4.4.5.3.1 4.4.5.3.1 .a 4.4.5.3.1.b 4.4.5.3.1 .c 4.4.5.3.1.d 4.4.5.3.1.f 4.4.5.3.1.g 4.4.5.3.1 .h

CSB Facility Description CSB Facility Classification ExpectedPotential Source Term Posting Justification Radiological Posting of CSB Radiological Controlled Area CSB Administrative Building Change Room RBA Exit Area HPT Count Room Filter Room HVAC Equipment Area Operating Area Operating Deck Trailer Vestibule MCO Receiving Pit Maintenance Pit Sample/Weld Station MCO Storage/Sampling Tubes VenUPurge Cart MHM Rotating Turret Assembly -

5.0 Historical Basin Radiological Comparison Table 5.1: Radiological Comparison of Basins Routine Water Analysis and Isotopes of Interest 5.1

5.2 Cesium 137 5.3 Strontium 90

Page 19 Page 21 Page 2 1 Page 22 Page 22 Page 22 Page 22 Page 22 Page 23 Page 26 Page 26 Page 27 Page 27 Page 27 Page 27 Page 28 Page 28 Page 28 Page 29 Page 30 Page 30 Page 30 Page 30 Page 3 1 Page 3 1 Page 3 1 Page 32 Page 32 Page 32 Page 33 Page 33 Page 33 Page 33 Page 34 Page 34 Page 35 Page 35 Page 35 Page 35 Page 35 Page 36 Page 36 Page 36

Revision 2 SNF-4358


5.4 Tritium 5.5 Transuranic Radionuclides

6.0 Regulatory and Contractual Basis 6.1. 6.2. 6.3. 6.4 Requirements Discussion

7.1 Contamination Areas (including HCA) 7.2 Radiation Areas (including HRA) 7.3 Radiological Buffer Areas 7.4 Radioactive Material Areas 7.5 Fixed Contamination Areas 7.6 Underground Radioactive Material Areas 7.7 Uncontrolled Areas

10 CFR 835, Occupational Radiation Protection "F-SP-1145, FDH Radiation Protection Program "F-5173-1, Hanford Site Radiological Controls Manual

7.0 Current Required Radiological Surveillance Monitoring Program

8.0 Non-Routinely Surveyed Areas Table 8.1: Non-Routinely Surveyed Areas Figure 8.1 : KE/KW 0' Level Map Figure 8.2: KEKW Basement Map

9.0 Current Radiological Trending Program 9.1

9.2

9.3

9.4

- - KE Bask

9.1.1 KE Basin Radiation and Contamination Trending Figure 9.1: KE Basin Radiation Survey Points Figure 9.2: KE Basin Contamination Survey Points Figure 9.3: KE Main Basin Trending Chart

9.1.2 KE Basin Airborne Radioactivity Trending Figure 9.4: KE Basin Air Quality Chart

KW Basin Radiation and Contamination Trending Figure 9.5: KW Basin Radiation Survey Points KW Basin Airborne Radioactivity Trending Figure 9.6: KW Basin Air Quality Graph

KW Basin 9.2.1

9.2.2

Cold Vacuum Drying Facility 9.3.1 CVD Radiation Trending

Figure 9.7: CVD Radiation Survey Points: First Floor Figure 9.8: CVD Radiation Survey Points: Second Floor

9.3.2 CVD Contamination Trending

Canister Storage Building 9.4.1 CSB Radiation Trending

9.4.2 CSB Contamination Trending 9.4.3 CSB Airborne Radioactivity Trending

9.3.3 CVD Airborne Radioactivity Trending

Figure 9.9: CSB Radiation Survey Points

Page 4

Page 36 Page 36 Page 37 Page 37 Page 37 Page 37 Page 38 Page 38 Page 38 Page 38 Page 38 Page 38 Page 39 Page 39 Page 39 Page 39 Page 40 Page 41 Page 42 Page 43 Page 43 Page 43 Page 43 Page 44 Page 45 Page 46 Page 46 Page 47 Page 47 Page 47 Page 48 Page 48 Page 49 Page 49 Page 49 Page 50 Page 5 1 Page 51 Page 5 1 Page 5 1 Page 52 Page 52 Page 53

Revision 2 SNF-4358


10.0 Other Trending Programs and Corrective Action Programs 10.1 Routine Surveillance Promam 10.2. Radiological Performance Report 10.3. ALARA Committee 10.4. Emplovee Zero Accident Prevention Council

11 .O References

Page 5

Page 53 Page 53 Page 53 Page 53 Page 53 Page 54


1.0 Introduction

This report documents the technical basis for the Spent Nuclear Fuel (SNF) Program radiation and contamination trending program. The program consists of standardized radiation and contamination surveys of the KE Basin, radiation surveys of the KW basin, radiation surveys of the Cold Vacuum Drying Facility (CVD), and radiation surveys of the Canister Storage Building (CSB) with the associated tracking. This report also discusses the remainder of radiological areas within the SNFP that do not have standardized trending programs and the basis for not having this program in those areas.

2.0 Background

The K Basins contain approximately 2000 metric tons of spent nuclear fuel. This fuel is contained in two pools which were designed and built during the 1950’s. This fuel was originally designed for approximately 150 days of storage following use. Some fuel in the basins is approaching 25 years in storage. A significant fraction of the fuel assemblies were breached during discharge from the N Reactor and subsequent handling. A recent survey determined that approximately 40% of the fuel assemblies in KE are damaged to some extent. Corrosion of the exposed metallic uranium results in the continuous release of radionuclides into the basin water.

The current plans consists of removing the SNF from the basins, vacuum drying in the Cold Vacuum Drying facility (CVD), conditioning and sealing the SNF in inert-gas filled canisters for dry vault storage in the Canister Storage Building (CSB), for up to 40 years pending decisions on ultimate disposition. Subsequent to fuel removal, activities include transfer of the basin sludge to Hanford’s T-Plant for management, disposal of non-SNF basin debris in the Environmental Restoration Disposal Facility (ERDF) burial ground at the Hanford Site, disposition of the basin water, and deactivation of the basins pending decommissioning.

In support of moving fuel, major construction involving major modifications to old facilities and construction of new facilities is ongoing.

3.0 Facility Management Responsibilities

Table 3.1 lists the facilities for which the SNF is responsible. These facilities range from uncontrolled facilities to facilities containing High Radiation Areas (HRG).

(11.6. 11.17,11.20)


K EAST K WEST

BLDG

105-KE

119-KE

NAME BLDG NAME

Basin, Office & Fuel Transfer Area 105-KW

Exhaust Air Sample Building 1908 KE Outfall Instrumentation Building

Basin, Office & Fuel Transfer Area

1614-KE

165-KE

181-KE

Environmental Monitoring Station 165-KW Switch Gear, Control Room

Power Control Building Elect 183-KW Chlorine Vault switch gear, Control Room

River Pump House 171 3-KW WarehouseIShop

1 1713-KE I Shop Building I 200 EAST

183-KE

190-KE

I I 1714-KE I Oil & Paint Shop I212H I Canister Storage Building

Clearwells, Filters, Sedimentation 1714-KW OilIPaint Storage Basins, Headhouse, Chlorine Vault

Main Pump House 1908 K Outfall Structure

1 G 6 - m

1706-KEL

1706-KER

Fan House (Also referred to as /1717-AKE 1167K) I

Water Studies Semi-works Facility 185K Package Water Treatment Plant

Development Laboratory 142K Cold Vacuum Drying Facility

Water Studies Recirculation Building

Bechtel Hanford, Inc. (BHI), controls the balance of the 100 K Area facilities for eventual decontamination and decommissioning (D&D).

1717-W

1724-K

(11.6. 11.10, 11.17) 4.0 Facility Descriptions/History and Radiological Status

I

Maintenance Shops Other associated mobile office structures

4.1 KE Area

4.1.1 KE Basin

The 105 KE basin was prepared for the long-term storage of N Reactor fuel in 1973. The water treatment systems initially installed in the 105 KE basin consisted of a primary recirculation loop with a heat exchanger and cartridge filters. The cooling water for the heat exchanger was drawn from the Columbia River. Thus, the basin water temperatures fluctuated seasonally with the

-I_-


river temperature. The cartridge filters provided the only means of removing radioactivity; they were rated to exclude particles with sizes of 5 micrometers or larger.

Modifications to the 105 KE basin proved to be inadequate. The concrete walls and basin floor were not cleaned or coated. The fuel in the basin was stored in open canisters, allowing unrestricted contact between the basin water and the damaged fuel. Much of the fuel shipped to 105 KE had a much higher bumup than the weapons grade fuel and had already been in storage in the N Reactor basin for several years. Filtered water from the Columbia River was used initially to fill the 105 KE pool and replace water lost through evaporation or leakage. The water quality in the basin deteriorated rapidly during the first years of operation, resulting in high radionuclide concentrations in the water and high corrosion rates of some metals (principally aluminum, copper alloys, and carbon steel).

A considerable fraction of fission products that leached from the fuel became imbedded in the bare concrete walls and recirculation piping, resulting in high dose rates, particularly at work locations above the walls.

In 1978, the water treatment loop was modified to include a sand filter and IXCs in an attempt to lower radionuclide concentrations. The IXCs were loaded with a cesium-specific zeolitic resin that reduced the I3'Cs concentrations in the water to acceptable levels. However, 90Sr and tritium concentrations continued to increase, as did concentrations of nonradioactive ions that were not removed by the zeolitic resin. These measures had little effect on dose rates in the basin because the cesium and strontium, already imbedded in the basin walls, were the primary source of the dose. Over the next few years, several steps were taken to improve water quality in 105 KE pool. Strontium-specific ion exchange resins were tested; water lost though evaporation or leakage was replaced with deionized water; hydrogen peroxide was used to control algae growth (replacing the chlorine-based algaecides); and the water-cooled heat exchanger was replaced with a refngerant chiller to maintain consistently low temperatures in the basin. Despite these efforts, the water chemistry in the basin continued to degrade.

In 1986, the IXCs in 105 KE basin were charged with mixed-bed resin, and the basin was slowly deionized over a six-(6) month period. Basin water conductivity was reduced from 280 microS/cm to 5 microSicm, and the 90Sr concentrations were reduced from 75 microCi/L to 1 microCi/L through this effort. The improvement in water quality markedly decreased uniform corrosion and pitting rates on corrosion specimens in the basin. The corrosion rate of the uranium metal in the basin was also markedly reduced, evidenced by a three-fold reduction in '37Cs leach rate, following deionization.

The eleven (1 1) years of operation between 1975 and 1986, with poor basin water quality, have left a legacy that still impacts operation of the 105 KE basin. Deionization of the basin did not significantly lessen the dose rates in the operating area.

Revision 2 SNl:-J358 Page 0

Since this time. dose ralcs have been reduced through the SNFP Dose Reduction Project. This pro-jcct required removing the surface layer of the concrete at the "bathtub ring" and coating ( I'N4), the application of shielding ovei- perimeter areas of the hasin grating. scvcral hydrolasing campaigns, and rcmoval of abandoned pipin%. This has reduced the overall average dosc rates by approximately 60%. This equates in a reduction from about 10-13 niR/lir to 5-7 niR/lir. See Figures 4.1 & 4.2.

Dose rates have began to increase slightly due to continued operation, iis expected, but this increasc can currently only he seen on area monitoring dosimeti-y at radiological boundaries. Figure 4.3 shows current and projec,tcd doses at four monitoring locations showing projected annllal dosc

Figure 4.1: 4/94 Grid Survey

105KE Chest 1,cvel Grid Survey (April 94) Dose Rate in nirem/hr; Water Level at 16'-1.7"

Revision 2

I

I ~ IN

SNF7-4358

k'igure 4.2: 9/06 Grid Survcy

105KE Chest Level Grid Survey (9127196) Dose Rate in mrernillr; Water Level at 16'-10"

Page I O

Figurc 4.3: Area Monitoring Dosimetry Projections

Locations 1. 6 , 22 B 23 Projected Annual Doses

-Location 1 Annual D O S ~

location 6 Annual Dose

*l~ucation 22 Annual Dose

~ Localmn 23 Annual Dosc


The basin is currently posted a RadiatiodContamination Area (RAICA). There is one high radiation area at the Ion Exchange Column (IXC) shielded enclosure. The four roof vents are posted and controlled as CA's. There are various other radiological areas such as Fixed Contamination Areas (FCA), Radioactive Materials Areas (RMA), and Radiological Buffer Areas (RBA).

Dose rates in these "other" types of radiological areas are generally


4.1.7 185-K Package Water Treatment Plant

Located beside 183KE, the 185-K building contains a newly constructed package water treatment plant that provides potable water, service water and fire protection water for the lOOK area. There are no radiological areas in this facility.

4.1.8 190-KE: Main Pump House

Located southeast of 1706KE, this building is currently used as a warehouse for SNFP, paint storage area, and excess equipment storage. Currently contains a small RMA (-400 ft2 ) for the storage of equipment.

4.1.9 1706-KE: Water Studies Semi-works Facility

Located south of 105KE, the 1706-KE building supplied eight single pass process tubes in the KE Reactor core and contained 22 external reactor test loops. All of these facilities have been deactivated. Used as a development laboratory for various Hanford projects over the years, 1706KE currently houses the SNFP Counting facility, and other operations/engineering personnel. Contains various radiological areas throughout its three levels. The 0' level contains the counting facility and all the other manned operations. There is RBAiRMA postings associated with the counting facility and other FCA's and internally contaminated piping. The - 13' contains FCA's and internally contaminated piping but is not routinely occupied. The -27' level (pump room) contains no radiological areas.

4.1.10 1706-KEL: Development Laboratory

Attached to 1706KE, this building was used as a development laboratory for various Hanford projects over the years. Currently houses waste operations group, provides storage, and is being used as a radiological training facility for teaching practical training to radiological workers. Contains potentially internally contaminated piping and ducting.

4.1.11 1706-KER Water Studies Recirculation Building

Located north of 1706KF,, the 1706-KER building supplied four high temperature, high pressure process tubes in the KE Reactor core. This portion of the facility was deactivated in 1964. The facility also served as a hot maintenance shop and decontamination facility and has been used as a development laboratory for various Hanford projects over the years. The 1706KER has a RCRA treatment system (i.e. 1706KE Waste Treatment System) pending closure, procedural or otherwise. The hot cell, hoods and laboratory benches have been removed. The HEPA exhaust system has been deactivated. Currently provides storage and is in process of being completely deactivated. Currently unoccupied and not routinely surveyed, except when entered. The 0' level is a RMA, the -13' level is an RBA and contains a RMA, and the -27' level contains an RBA and a large CA.


4.1.12 1713-KE: Shop Building

Located outside on the East Side of 105 KE, this corrugated sheet metal building is used for storage of non-radioactive materials and currently contains no radiological areas.

4.1.13 1714-KE: Oil & Paint Shop

Located outside on the East Side of 105 KE, this corrugated sheet metal building is used for storage of non-radioactive materials and currently contains no radiological areas.

4.1.14 1717-W1724-K Maintenance Shops

Located by inner fence perimeter access, these buildings currently houses the SNFP Maintenance and Work Control organizations. Contains shop and office areas. Currently contains no radiological areas.

4.1.15 1717-AKE: Fan House (Also referred to as 167K)

Located outside on the south side of 1717 K, this building contains fans for ventilating the tunnels. This is an unoccupied building and has no real access.

4.1.16 1908 KE & 1908 K Outfall Instrumentation BuildindOutfall Structure

Located next to the Columbia River, this structure contains the instrumentation to monitor our compliance with National Pollutant Discharge Elimination System (NPDES) permits. This is a normally unoccupied building and is only accessed to sample outfall or check instruments. Contains no radiological areas. However, the discharge flume that carries water to the outfall is contaminated from the past practice of discharging contaminated water through the process sewers of various facilities.

4.1.17 Other KE Structures

The KE area contains some mobile office trailers, which contain no radiological areas. There are some outdoor areas used for radioactive material storage and most of the areas discussed above, are located in an Underground Radioactive Material Area (UGRMA). There is a outdoor CA located at the railroad tracks between 105KE and 105KW. This was used to store N Reactor fuel shipping railroad cars.

4.2 KWArea

4.2.1 KW Basin

Preparations for use of 105 KW basin for storage of NReactor fuel began in 1978. Basin modifications were similar to those made at the 105 KE basin, but with several important differences. The bare concrete in the KW basin was cleaned and coated with epoxy paint. The water treatment system included IXCs charged with mixed-bed resin, a sand filter, a heat exchanger, and cartridge filters. The KW basin was filled with deionized water in 1979, and all


additions to replace water lost through leakage or evaporation were made with deionized water. In addition, the fuel canisters were sealed.

Therefore, the fuel storage canisters in the 105 KW basin show only minor evidence of pitting corrosion after fourteen years in the 105 KW pool, in contrast to severe corrosion of aluminum alloy canisters in the 105 KE pool.

This basin is posted as a RBNRMA. There is a small (-400 ft’) Contamination area located in the transfer area and a few very small CA around some pump bases and the center of basin sample station. A review of surveys of these CA areas, typically shows, little or no contamination. The CA posting is for the contamination control needed during operations. There are various other radiological areas such as Fixed Contamination Areas (FCA) in the KW basin areas.

4.2.2 165-KW: SwitchGear. Control Room

Located south of the 105 KW building, this building was originally designed as a safety class, bomb resistant structure and was used to house the main control of the K Basins water plants and emergency power service. This building contains no radiological areas.

4.2.3 183-KW: Chlorine Vault

Located south of 105 KW, this building is used to store spent MC after use. IXC have not been used in several years and these are awaiting disposal. This is an unoccupied building and contains a posted High Radiation Area (HRA) as well as a radiation area. Entries are only made to perform minimum survey requirements for monitoring. The building has no active systems or potential for changing radiological conditions.

4.2.4 1713-KW: Warehouse/Shog

Located on East Side of 105 KW, this building is currently used as shop area for Fluor Federal Services crafts. This building contains no radiological areas.

4.2.5 1714-KW: OiVPaint Storage

Located on East Side of 105 KW, this building is currently used for storage of materials and equipment. This building contains no radiological areas.

4.2.6 Other KW Structures

The KW area contains some mobile office trailers, which contain no radiological areas. There are some outdoor areas used for radioactive material storage and most of the areas discussed above, are located in an Underground Radioactive Material Area (UGRMA).


4.3 Cold Vacuum Drying Facility

4.3.1 Cold Vacuum Drving Facility Description

The mission of the Spent Nuclear Fuel (SNF) Project Cold Vacuum Drying Facility (CVDF) is to achieve the earliest possible removal of free water from Multi-Canister Overpacks (MCOs). The MCOs contain metallic uranium SNF that have been removed from the 1OOK Area fuel storage water basins (i.e., the K East and K West Basins) at the U S . Department of Energy Hanford Site in southeastern Washington state. Immediate removal of free water is necessary to halt water- induced corrosion of exposed uranium surfaces and to allow the MCOs and their SNF payloads to be safely transported to the Hanford Site 200 East Area and stored within the SNF Project Canister Storage Building (CSB).

The CVDF is located within a few hundred yards of the basins, southwest of the 165KW Power Control Building and the 105KW Reactor Building. This facility site is in close proximity to all of the required utilities. The site area required for the facility and vehicle circulation is approximately 2 acres. Access and egress is provided by the main entrance to the lOOK inner area using existing roadways.

The CVDF will remove free water from the MCOs to reduce the potential for continued fuel-water corrosion reactions. The cold vacuum drying process involves the draining of bulk water from the MCO and subsequent vacuum drying. The MCO will be evacuated to a pressure of 8 torr or less and backfilled with an inert gas (helium). The MCO will be sealed and leak tested and then transported to the CSB within a sealed shipping cask. (The MCO remains within the same shipping Cask from the time it enters the basin to receive its SNF payload until it is removed from the Cask by the CSB MCO handling machine.)

The CVDF has a minimum of two drying stations (with capability to expand to four stations), and provides excess capacity to simplify transport operations as defined by WHC-SD-SNF-TI-016, Development of Design Basis Capacity for SNF Project Systems. The CVDF subproject acquired the required process systems, supporting equipment, and facilities. The cold vacuum drying operations result in an MCO containing dried fuel that is prepared for shipment to the CSB by the Cask transportation system. The CVDF subproject also provides equipment to dispose of solid wastes generated by the cold vacuum drylng process and to load out process water removed from the MCO for transfer back to the appropriate handling facility.

4.3.1.1 Cold Vacuum Drying Operation overview

The CVDF operations are depicted schematically on the CVDF Process Flow Diagrams, H-l- 81166. The CVDF process and HVAC systems Engineering Flow Diagrams are depicted on drawings H-1-83965 through H-1-83968. The overall drawing list for the CVDF is contained within SNF-5335, Preliminary Spent Nuclear Fuel Cold Vacuum Dy ing Facility Essential & Support Drawing List

The Cask-MCO handling operation starts with the receipt of the Cask-MCO trailer at the CVDF process bay. Operators raise the door to allow the trailer to back into the process bay.


The trailer is backed into a predetermined position, the landing legs are lowered to lift the front of the trailer to allow disengagement of the king-pin, and the tractor is disconnected. The tractor is dnven out of the bay and the bay door is closed. Bay confinement is then established. Radiation surveys are conducted on the caskitrailer, and the quality assurance package is delivered to the CVDF shift operations manager.

Next, the process bay instrument air service is connected to the trailer air supply control station, and the trailer is leveled. Finally, a bridge is installed from the process bay mezzanine to the trailer work platform. Contamination control supplies are installed on the work platform. The top of the cask is prepared for venting, purging, and cask lid removal.

The MCO is vented to the cask headspace (the cask annulus was partially filled with water by operations at the KBasins). The cask headspace was purged and filled with helium at approximately 3 psig before it left the basin. During transport to the CVDF, hydrogen is generated and vented to the cask. Pressure also increases due to temperature increases associated with radioactive decay heat, solar heating, and water-uranium corrosion reactions. This gas is vented to the cold vacuum drying process vent system by means of special venting hardware and flex lines connected to the cask lid port and the CVDF process vent system. After venting, the cask headspace is purged with helium. Following purging, the cask lid is removed by the CVDF process hay overhead crane using a dedicated lifting fixture. The cold vacuum drying process hoodseal ring assembly is installed onto the cask and the MCO is prepped for the process operations described below.

The process is performed per prescribed operating procedures, which include bulk water removal, helium purging, evacuation with and without helium purge, an initial pressure rebound test, an extended operation under vacuum at the base pressure of the system, a final pressure rebound test, an integrated leak test of the MCO mechanical seals, and backfilling the dried MCO with helium. The majority of all process actions are automatically actuated by the Monitoring and Control System (MCS) with input (starthop commands) from the operators. There are minimal manual operator actions in the process sequences. Field operator actions are required, such as connecting MCO valve operators, DVHe rinseiblowdown after draining and tempered water connections. The control room operator actions include acknowledging alarms or instructing the MCS to proceed with the next step. Valve state changes, water temperature control and other process parameter changes are performed by the MCS.

Protection from causing off-site consequences is provided by maintaining water in the Cask- MCO annulus space, pressure venting via a 30 psig rupture disc and automatic actuation of the Safety Class Helium (SCHe) system and the Safety Class Instrument and Control (SCIC) system. The 30 psig vent path is passive and maintaining annulus water is, for the most part, passive. The SCIC will prevent the TW temperatures from exceeding the safety limit of 5 0 T by stopping the TW heater on high temperature detection.

There are two normal operator interfaces; the MCS computers and the SCIC Mode switch, both are located in the CVDF control room.

The SCIC seven position switch defines the MCO process mode and directly feeds this information into the Monitoring and Control System (MCS). While most sequences are held for


operator permissives, some occur based solely on SCIC mode switch position. These are related to keeping hydrogen concentration low in the MCO and PWC receiving tanks.

Following the cold vacuum drying process, the Cask-MCO is prepared for shipment to the CSB. This operation is the reverse of the receipt operation. The Cask-MCO is cooled by the Tempered Water system; the MCO pressurized with helium, and is then closed. The Cask-MCO annulus is drained and dried with an instrument air purge, a He leak test of the MCO is performed, and the cask lid is then reinstalled. The trailer is connected to the tractor and released for shipment to the CSB.

4.3.2 Facility Classification

The CVDF process building is classified as a nonreactor nuclear facility according to DOE Order 6430.1A, Section 1300. The administrative area is a non-nuclear facility rated for office or business use in accordance with the Uniform Building Code (UBC) (ICBO 1994).

4.3.3 Expected Source Term

Potential radiation sources associated with the cold vacuum drymg process are:

The SNF and contaminated water inside the MCO

Fuel will be remain contained inside the MCO/cask assembly and not be directly exposed to the facility environment.

Dissolved fission products and fuel particulate contained in the water extracted from each MCO

The water removed from the MCO will be contained to the PWC system and VPS collection tank.

Radioactive particulate associated with gases that are purged or evacuated from the MCO

These particulates will be vented via the process vent system through High Efficiency Particulate Filters (HEPA).

The radioactive particulate that could be carried by gases is confined, for the most part, inside the shielded MCO by the metal high-efficiency particulate air filter and is assumed to constitute an insignificant radiation source compared to the SNF and contaminated MCO water.

Conversion of fuel mass (grams) to radionuclide activity (curies) was done using the average fuel curies per metric tonne uranium (Ci/MTU) data tabulated in HNF-SD-SNF-TI-009, 1 0 5 4 Basin Material Design Basis Feed Description for Spent Nuclear Fuel Project Facilities.

Table 4-1 lists the significant radionuclides and their activity levels. The activity units in Table 4-1 are microcuries per gram (pCi/g) of fuel plus fission products, which are equivalent to the units of Ci/MTU reported in HNF-SD-SNF-TI-009.


SNF source material compositions have been divided into two classes, as specified in HNF-2850, Shielding Analysis for the Cold Vacuum Drying Project, Appendix A. The first class includes the SNF in an MCO and unfiltered water extracted from an MCO. The composition of the material is listed in the “Unfiltered” column of Table 4-1. In the second class (filtered process water), there is only water and 137Cs/137”Ba, under the assumption that all insoluble constituents included in the first class have been removed by the M M and mechanical filter in the PWC room. The composition of the filtered process water is listed under the “Filtered” Column in Table 4-1.

(1 1.13) Table 4-1. Radionuclide Concentrations in Source Material.

The fuel inventory in an MCO was assumed to be 5.25 MTU from HNF-SD-TI-009. The source material concentrations in “F-2850, Table 4-3 were used to determine radiation source strengths. The table shows the curie concentration for each PWC component. However, actual testing results from HNF-4057, Cold Vacuum Diying Proof of Performance Test Results, indicate that 10% of the estimated mass of particulate actually carried over from the MCO. Table 4-2 below has been corrected to reflect these values.


Table 4-2. Source

Component

Piping

Check valve

Receiver tanks

Ion exchange modules PWC filter

5,000 gallon tank

Material Concentra

Source specification

6.26 g fuel removed with 490 L of process water (unfiltered) 3 g fuel per valve (1 cm3 volume) (unfiltered) 6.26 g fuel suspended in 490 L process water (unfiltered) 2.5 kg fuel in each IXM (unfiltered) Contains I3'cs /'37mBa from 6.26 g fuel (filtered) Contains I3'Cs /'37mBa from 6.26 g fuel (filtered)

Fuel concentration in process water (s/cm3) 1.27 E-OS

Process Water Activity (uCi/cm3) for dose contributing nuclides 7.56 E -02

4.3.3.1 Methodology and Expected Dose Rates

Shielding calculations were carried out using the computer programs ISO-PC version 1.98 (Rittmann 1995) and MCNPH (Carter 1996). ISO-PC is based on the ISOSHLD-I1 code (Simmons 1967) that uses the point-kernel integration technique to compute gamma-ray attenuation. Bremsstrahlung was accounted for in all ISO-PC calculations. MCNPH is the Hanford version of the Monte Carlo N-Particle Code MCNP (Breismeister 1993). The shielding attenuation properties for the bulk materials used in the ISO-PC calculations were obtained from the KO-PC 1.98 data library. ISO-PC converted computed photon fluxes to dose rates (mrem/hr) using the ANSI/ANS-6.1.1-1991 factors for anterior-to-posterior exposure. These fluence-to- dose conversion factors are also included in the ISO-PC data library.

ISO-PC was used to compute dose rates around the piping in the CVD process bay and to generate photon source rate distributions for input to MCNPH. MCNPH was used to compute dose rates throughout the PWC room. Most MCNPH simulations were run in the rigorous photon transport mode. However, dose rates due to the photon source in the two E M S were computed using the point-kernel option in MCNF". This approximation was made because the

3.00 E+OO

1.27 E-05

4.06 E-03

5.52 E-04

3.31 E-07

1.79 E +04

7.56 E -02

Shielded

3.29 E +00

1.97 E -03


IXMs are heavily shielded, and dose rate contribution from photon sources in the E M S is very low compared to the other sources in the PWC room. The heavy shielding would have required modeling detail and computer time that were not warranted by the relatively low dose rates. Nuclear data in the MCNPH calculations were obtained from the MCNP ENDF/B-V library. As in the ISO-PC calculations, photon fluxes were converted to dose rates using ANSI/ANS-6.1.1- 1991 factors. In the transportation, dose rates were tallied throughout the room on a three- dimensional rectangular grid using a special option in MCNPH. In the point-kernel calculation, dose rates were tallied only in the vicinity of the E M S .

Only photon radiation exposures were computed in HNF-2850. Existing shielding, pipes, and vessel walls effectively attenuate alpha and beta radiation. However, bremsstrahlung was accounted for in the ISO-PC calculations. Neutron dose rates from SNF sources are orders of magnitude less than photon dose rates outside unshielded, water-shielded, or concrete-shielded containers. This was verified by estimating the neutron dose rate for one case using the method given in Estimation ofNeutron Dose Rates from Nuclear Waste Packages (Nelson 1996). In that case, the neutron dose rate 30 cm from a check valve with a 3-g deposit of fuel particulate (an assumed worst case) was computed and compared to the photon dose rate reported in HNF-2850. The result of this comparison was that the neutron dose rate was only 0.04% of the photon dose rate (0.022 versus 52.1 mrendh). Steel, on the other hand, is much more effective at attenuating photons than neutrons. Thus, the fractional contribution of neutrons to the total dose rate outside the MCO transportation cask will be higher. However, previous analyses (WHC-SD-SNF-CAVR-001 and HNF-SD-SNF-CN-026) determined that neutrons still account for only about 10% of the dose rate outside an MCO cask. Process Bay Dose Rates

The ISO-PC program was used to compute dose rates as a function of radial distance from, and axial position along, sections of 1 in, schedule 160 pipe containing water contaminated with fuel at 0.000127 gr./cc. ISO-PC was also used to compute dose rates as a function of distance from a check valve hypothesized to contain a deposit of 3 gr. fuel. These results are also shown in Table 4-3.

These calculations were performed using estimated curie contents of the process water, not the actual testing results from HNF-4057. The piping values from Table 4-3 can be reduced to 10% of shown values. The values for the MCO and check valves are expected to be accurate.


Table 4-3. Calculated Dose Rates in Process Bay

Dose Rate (mremh)

From these values it can be seen that dose rates from the piping system in the process bays can be expected to not contribute appreciably to the dose rates in the bay directly from processing operations. However, build up over time could cause this piping to be a significant contributor along the west wall of the bay. By far, the presence of the MCOkask unit will be the overriding contributor to dose rates in the hay. Contributions from the VPS collection tank can be expected to build over time, but due to the very low levels of particulate in the water collected in this tank, this i s not expected to a significant source.

Dose rate contour maps can be found in HNF-2850, Figures 6-1 - 6-6 that show calculated dose rates at different locations in an operating process hay. Include these figures in this document.

4.3.3.1 .a Process Water Conditioning Room Dose Rates

The first source material inventory considered (Case A) was maximum loading of 62.6 grams of fuel that resulted in the fill photon source rate. Case B was like Case A except that there was no


source material in the PWC filter. Case C was like Case B except that the receipt tanks were assumed to contain only 10 percent of the source material present in the first case.

Case C is expected to be the most likely situation for the PWC room. HNF-4057 found that about 10% of the generated fuel particulate is carried out of the MCO during draining. The PWC filter is not expected to have a substantial dose rate initially, plus it is able to be backwashed. The peak dose rate was calculated to be about 24 mrem/hr near the receipt tanks. Also, contour maps can be found in HNF-2850, Figures 6.7 ~ 6.15 that show general dose rate calculations at different locations in the PWC room. Include the contour maps in this document.

4.3.3.2 Discussion of ExDected Dose Rates

The dose rates for each component are calculated for theoretical maximum radionuclide concentrations in the process water. These rates are not expected to be reached until an unknown number of MCO’s have been processed. Radiation and contamination surveys will be performed routinely in order to document radiological conditions that are present and to trend any changes in those conditions. As facility operations progress and radiological conditions change, posting criteria contained in AP-14-009 will be followed to maintain regulatory compliance.

4.3.4 Radiological Posting of CVD

4.3.4.1 CVD Siting

The entire CVD Facility is located inside a posted Radiologically Controlled Area (RCA) established to indicate the presence of radiological areas within its boundary. Requirements for access to this area are a minimum of General Employee Radiological Training (GERT), with no requirement for access control verification of training completion. Access control is maintained by administrative procedures of the PHMC, by requiring all employees to have Hanford General Employee Training (HGET) and any visitor to have GERT to be badged.

4.3.4.2 CVD Facility Administrative Building

There are no radiological areas within the administrative building of the CVD Facility. However, areas adjacent to posted areas will be surveyed to verify effectiveness of controls.

4.3.4.3 Access Corridor

The transfer corridor will have no additional posting at the start of cold vacuum drying operations. This corridor does not contain any contaminated systems and is not expected to have dose rates requiring any further posting. Routine surveillance of this area will be performed to ensure area remains contamination free and does not require posting for dose control. No formal radiological access control is required for this area.


4.3.4.4 Process Bavs

The radiological conditions of a process bay will change when an MCO/Cask shipping container has been brought into the bay and when it is removed because of dose rates from the MCO. Initially, it is not expected that the process systems in the bays will not contain enough contamination to effect the general area dose rates.

As the number of MCO’s that have been processed increases, system contamination levels are expected to increase and therefore the dose rate levels will also increase.

A. Process Bay 5

This bay will be utilized to process MCO containers through the cold vacuum drylng process. As such, many of its systems will become contaminated and require contamination control measures.

When an MCO is received, the bay will be posted as a radiation area (RA)/radiological buffer area (RBA) (for contamination control). During processing, a contamination area (CA) will be established around the work platform of the transport vehicle and the south mezzanine to allow for process connection activities. In addition, depending on the ability to keep the seal ring and process ventilation exhaust hood free of Contamination, it may be necessary to establish a contamination area at the storage location for those times when the unit is not in use.

The associated bay airlock will be posted as Radiological Buffer Area (RBA)/Radioactive Material Area (RMA) as a contamination control buffer and for storage of radioactive material such as laundered protective clothing. An Eberline PCM-1B will be used to survey personnel exiting the RBA as well as for whole body surveys required after CA entries. This personnel monitor is located sufficiently close to the RBA exit to satisfy it use as a exit monitor. It is located approximately five feet from the boundary. This with the shiftly routine surveillance of this area is more than satisfactory in meeting this requirement.

During other activities in the bay such as maintenance, it will be necessary to establish temporary radiological areas for control during these activities.

Radiation and contamination levels in the process bay will be evaluated after the MCO/Cask shipping container has completed the cold vacuum drying process and is removed. A routine surveillance program will be used to characterize the area and determine any needed posting changes that are needed to maintain regulatory compliance.

Formal access control is required for accessing this area to verify qualifications and training of personnel needing access per AP-14-004, Radiological Area Access Control.


B. Process Bay 4

This bay will also be utilized to process MCO containers through the cold vacuum drying process. As such, many of its systems will become contaminated and require contamination control measures.

When an MCO is received, the bay will be posted as a radiation area (RA)/radiological buffer area (RBA) (for contamination control).

During processing, a contamination area (CA) will be established around the work platform of the transport vehicle and the south mezzanine to allow for process connection activities. In addition, depending on the ability to keep the seal ring and process ventilation exhaust hood free of contamination, it may be necessary to establish a contamination area at the storage location for those times when the unit is not in use.

The associated bay airlock will be posted as Radiological Buffer Area (RBA)/Radioactive Material Area ( M A ) as a contamination control buffer and for storage of radioactive material such as laundered protective clothing. An Eberline PCM-1B will be used to survey personnel exiting the RBA as well as for whole body surveys required after CA entries. This personnel monitor is located sufficiently close to the RBA exit to satisfy it use as a exit monitor. It is located approximately five feet from the boundary. This with the shiftly routine surveillance of this area is more than satisfactory in meeting this requirement.

During other activities in the bay such as maintenance, it will be necessary to establish temporary radiological areas for control during these activities.

Radiation and contamination levels in the process bay will be evaluated after the MCO/Cask shipping container has completed the cold vacuum drymg process and is removed. A routine surveillance program will be used to characterize the area and determine any needed posting changes that are needed to maintain regulatory compliance.


C. Process Bay 3

This bay will not be used for process operations but will be fully outfitted for such operations, except for testing and some connections. It is expected that these components will be used for spares for the operating bays and possibly for storage of radioactive material. The PWC line from process bay 4 & 5 operations passes through this bay along the west wall and is accessible. It is not expected that this line will contribute significantly to the dose rates in the bay. A routine surveillance program will be used to characterize the area and determine any needed posting changes that are needed to maintain regulatory compliance.


At the beginning of operations, this area will be posted as “Internally Contaminated Systems., .” due to presence of installed, contaminated components. If radioactive material is stored in this area, a separate RMA will be established in a designated area in accordance with AP-14-001.

Formal access control will be not required for accessing this area normally. If and when a RMA is established, then formal access control will be required for accessing these RMA’s to verify qualifications and training of personnel per AP-14-004, Radiological Area Access Control.

D. Process Bay 2

This bay will not be used for process operations and will not be fully outfitted for such operations. It is expected that this area will be used possibly for storage of radioactive material. The PWC line from process bay 4 & 5 operations passes through this bay along the west wall and is accessible. It is not expected that this line will contribute significantly to the dose rates in the bay. A routine surveillance program will be used to characterize the area and determine any needed posting changes that are needed to maintain regulatory compliance. At the beginning of operations, this area will be posted as “Internally Contaminated Systems.. .” due to presence of installed, contaminated components. If radioactive material is stored in this area, a separate RMA will be established in a designated area in accordance with AP-14-001.

Formal access control will be not required for accessing this area normally. If and when a M A is established, then formal access control will be required for accessing these RMAs to verify qualifications and training of personnel per AP-14-004, Radiological Area Access Control.

E. Process Bay 1

This bay will be used to off-load the processed MCO water collected from operations, into a 5,000-gallon tanker truck for transport to the Liquid Effluent Treatment Facility (LETF) for further processing. The curie content of this water is estimated to be low due to the repeated processing through an IXM and subsequent high efficiency filtration prior to entering the storage tank. However, it is still considered contaminated and is to be treated as such. The PWC line from process bay 4 & 5 operations passes through this bay along the west wall and is accessible. It is not expected that this line will contribute significantly to the dose rates in the bay. This area may also be used for storage for radioactive material and for a work area for work with contaminated materials. With this bays close proximity to the PWC room and the orientation of the building this room will be posted as an RBA for contamination and radiation control.

The associated bay airlock will he posted as Radiological Buffer Area (RBA)/Radioactive Material Area (RMA) as a contamination control buffer and for storage of radioactive material such as laundered protective clothing. An Eberline PCM-IB will be used to survey personnel exiting the RE3A as well as for whole body surveys required after CA


entries. This personnel monitor is located sufficiently close to the RBA exit to satisfy it use as a exit monitor. It is located approximately five feet from the boundary. This with the shiftly routine surveillance of this area is more than satisfactory in meeting this requirement.


4.3.4.5 Process Water Conditioning Room

The PWC room is utilized to collect and process wastewater from MCO processing operations.

This room contains the highest source term next to a loaded MCO. This source term is contained in various shielded and unshielded configurations.

The waste water is collected in two receiving tanks and circulated through an Ion Exchange Module (IXM) which performs ion exchange and some filtering (to -20 micron). Following E M processing, the water is passed through a high efficiency filter and stored in a 5,000 gallon tank awaiting off-load to a tanker truck for shipment to LETF.

This room has the potential to become a High Radiation Area (HRA) due to design configurations. Due to this potential, this room must be able to have HRA controls invoked at any time and must be set up for such. Items such as crash bars on doors and single keyed locks with key control will be required. At the start of operations, this room will be posted as a RA/RE%A and will be closely monitored. The 5,000-gallon tank is not expected to contribute significantly to the overall dose rates.

This room shares a airlock with bay 1 and uses the same egress process and controls.


4.3.4.6 Mechanical Room

The mechanical room contains several radiologically contaminated ventilation systems, the process hood ventilation system and general exhaust system. These systems are made up of contaminated ventilation ducting, HEPA filter housings containing HEPA filter units and pre-filters. These components are located in the south end of the room and are not expected to have a significant external dose rate during processing operations. At the beginning of operations, this area will be posted as “Internally Contaminated Systems.. .” due to presence of installed, contaminated components. If these systems are breached, then radiological conditions will need to be reevaluated and postings changed per criteria HNF-5173, Tables 2-3 and 2-4.


A routine surveillance program will he used to characterize the area and determine any needed posting changes that are needed to maintain regulatory compliance. radioactive material other than the installed systems will not be stored in this area.

Formal access control will not he required for accessing this area normally.

4.3.4.7 Health Physics Counting Room

This room is located off the access comdor and will he used for the counting of radiological swipes, air samples, and other materials. In addition, radioactive sources needed for radiological operations will he stored in this room. Due to the nature of these activities, it is necessary to post this area as a RBA/RMA. Personnel survey will be performed by the HPTs themselves due to their site qualification of self-survey.


4.3.4.8 Decontamination Room

This room is located off the access conidor and will only he used for personnel decontamination if needed. Until such time as the facility is used, it will remain un- posted. Routine surveillance of this area will he performed to ensure area remains contamination free and does not require posting for dose control. No formal radiological access control is required for this area until as such time as the room is used for its intended purpose.

4.3.4.9 Uninterruptahle Power Supplv Room

This room is located off the access corridor and contains backup power for essential systems. This room does not contain any contaminated systems and is not expected to have dose rates requiring any further posting. Routine surveillance of this area will be performed to ensure area remains contamination free and does not require posting for dose control. No formal radiological access control is required for this area.

4.3.4.10 Securitv Room

This room is located off the access comdor and contains security systems for the facility. This room does not contain any contaminated systems and is not expected to have dose rates requiring any further posting. Routine surveillance of this area will he performed to ensure area remains contamination free and does not require posting for dose control. No formal radiological access control is required for this area.


4.3.4.1 1 Air/Demineralized Water Room

This room is located off the access corridor and contains air and DI water systems for the facility. This room does not contain any contaminated systems and is not expected to have dose rates requiring any further posting. Routine surveillance of this area will be performed to ensure area remains contamination free and does not require posting for dose control. No formal radiological access control is required for this area.

4.4 Canister Storage Building

4.4.1 CSB Facility Description

The CSB provides facilities for receiving, sampling, monitoring, and interim storage of high- level radioactive waste. The facility will incorporate a number of methods to accomplish this task. Multi-Canister Overpacks (MCOs) will be used as the transportatiodstorage package for Spent Nuclear Fuel (SNF) currently stored within the 100 K Area storage basins. In the future, Shipping Port fuel, fuel from the Fast Flux Test Facility (FFTF), and waste that has been through the glass vitrification process will be stored here as well. The scope of this document will only cover the process for posting radiological areas during operations for storage of MCOs from lOOK area.

The building is steel-framed and encloses the operating area and the support area. The operating area consists of the load-idload-out area and MCO Handling Machine (MHM) maintenance pit on the northern end, three equally sized, below grade concrete vaults, and the Sampling/Welding Stations on the southern end. The concrete vaults are covered by a concrete operating deck. Only the northernmost vault (vault 1) in the operating area is equipped with steel tubes for stagindstoring mechanically sealed MCOs. This vault has been designated for storage of MCOs from the lOOK Basins and future storage of Shipping Port Fuel. Future plans will utilize vaults 2 (center) and 3 (southern) for storage of Glass Vitrification Logs.

The Support area contains systems that will be used to support and monitor work evolutions in the operating area. It is located adjacent to the north side of the operating area. These systems include separate HVAC systems for the operations and support areas, Instrument air equipment, an Unintermptible Power Supply (UPS), the Distributed Control System (DCS), Continuous Air Effluent Monitoring (CAEM) system, and administrative offices.

The Interim Storage Area (ISA) is comprised of three concrete pads within a fenced area, located outside and to the west of the CSB. The ISA will be used for interim storage of FFTF fuel at a future date and will not be covered by the scope of this plan.

An MCO received at the CSB contains five or six baskets of dried SNF that are stacked one on top of the other. The stainless steel outer shell and bottom of the MCO and the mechanically- sealed top carbon steel shield plug provide effective long-term confinement of the enclosed SNF. The shield plug also provides access to the interior of the MCO through four process ports on the top of the shield plug assembly. These ports will be sealed at the Cold Vacuum Drying Facility (CVDF) using gasketed, bolted cover plates before the MCO is transported to the CSB. Select


process ports can be removed for sampling the internal atmosphere of the MCOs at the CSB if necessary.

At the load-idload-out area, a receiving crane is used to lift the shipping cask that contains the MCO off the transport trailer and then transfer the MCO/Cask to the MCO receiving pit. The MHM is used to lift the MCO out of the shipping cask and transport the MCO to a pre-selected storage tube in vault 1. A permanent cap will be welded onto the top of each MCO at a future date.

The CSB operating deck is a 5-ft-thick, at-grade, reinforced concrete structure. There are numerous full-thickness, embedded steel sleeves spaced throughout the deck. Each sleeve attaches to a standard storage tube or overpack tube. The tubes have only been installed for vault 1 , with the tubes for vaults 2 & 3 to be installed at a future date. These embedded sleeves also provide a location for securing the tube plugs and cover plates.

Vault 1 contains 220 standard storage tubes and 6 overpack storage tubes. Vault 1 is cooled by natural convection through above-grade concrete intake and exhaust plenums. The 3-ft-thick interior walls of the subsurface structure provide shielding from the source term associated with SNF storage. The exterior walls are designed to meet the shielding criteria given in Title 10, Code of Federal Regulations, Part 835, “Occupational Radiation Protection” (10 CFR 835), for uncontrolled access areas.

The tubes are supported from the foundation base slab of the vault and are accessed through the operating deck. The storage tubes are normally closed with removable tube plugs and then cover plates are installed to complete the deck surface. Each standard storage tube is capable of staging or storing two MCOs. The standard tube plugs provide radiation shielding, a filtered vent, and connections for sampling the storage tube atmosphere. Each standard storage tube contains both a bottom impact absorber and an intermediate impact absorber to mitigate the impact of a dropped MCO.

Each overpack storage tube is available to accommodate one abnormal or suspect MCO. The overpack storage tube plugs have connections for sampling, purging, and pressure-relief of the overpack tube atmosphere; pressure gauges for surveillance of the tube pressure; and lock-down devices. Each overpack storage tube is fitted with a bottom impact absorber. The SampleiPurge cart will be used to sample or purge a tube with a suspectedknown off-normal MCO.

4.4.2 CSB Facilitv Classification

The CSB is classified as a hazard category 2, nonreactor nuclear facility according to DOE Order 6430.1A, Section 1300. The administrative area is a non-nuclear facility rated for office or business use in accordance with the Uniform Building Code (UBC) (ICBO 1994).


4.4.3 ExpectedPotential Source Term

Potential radiation sources associated with the MCO handlingktorage at the CSB include:

Spent Nuclear Fuel inside an MCO

Fuel will remain contained inside the MCO/cask assembly and not be directly exposed to the facility environment.

Release of potentially contaminated atmosphere from the inside of the MCO

The cold vacuum drying process should eliminate the potential for over-pressurization that could potentially result in a release of contamination from the MCO.

(ll.18,11.19) 4.4.4 Posting Justification

Two documents were used to gather calculated dose rates for the CSB. The first was the CSB ALARA Analysis 09 and the second was HNF-SD-SNF-SARR-005, Multi Canister Overpack (MCO) Topical Report, Revision 2. These two documents summarize calculated dose rates for the different work locations and structures in the CSB operating area. The calculated dose rates were compared to requirements listed in “F-5173, Chapter 2, Part 3 “Posting” and AP-14-009 “Radiological Posting Process”. From the comparison, dose-related posting determinations were made for different areas and work evolutions.

CSB Radiological Control personnel determined contamination postings by evaluating the potential for systems and controls used to handle and store the MCOs to become contaminated. This evaluation was made based primarily on experience with similar systems. The choices for radiological postings for the facility were made on the conservative side in order to insure adequate warning to personnel.

4.4.5 Radiological Posting of the CSB

Potential radiological areas at the CSB have been analyzed for the potential to exceed radiation and contamination levels set by 10 CFR 835 and HNF-5173. Use of the areas was also evaluated. This section will document and justify posting selections for those areas.

4.4.5.1 Radiological Controlled Area

Prior to receipt of SNF at the CSB, an RCA will need to be established around the perimeter of the facility. Posting the perimeter of the facility will alert personnel of the presence of radiological areas at the facility. RCA signs will need to be added to each access point of the RCA that describe the entry requirements for the area. These access points include gates 793 (Turnstiles), 813,823, & 836 (Figure 3).

Requirements for access to this area are a minimum of General Employee Radiological Training (GERT), with no requirement for access control verification of training completion. Access

~ . ~ .... ~. .... .___-

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control is maintained by administrative procedures of the PHMC, by requiring all employees to have Hanford General Employee Training (HGET), any visitor to have GERT.

4.4.5.2 CSB Administrative Building

The administrative area includes the following rooms:

Compressed Air Equipment Area Filter Room HVAC Equipment Area Security Room UPS Room Utility Rooms 1 & 2 Control Room Telephone Room

Switchgear Room Corridor and Vestibule

0 Shift Manager’s Office HPTOffice ChangeRoom HP Counting Room

0 RBAExitArea

The Administrative Area of the CSB has both radiological and non-radiological areas. The non- radiological areas of the Administrative Area will be surveyed to verify effectiveness of contamination and dose controls. The radiological areas within the administrative building include the Change Room, the HP Counting Room, the RBA Exit Annex, the Filter Room, and the HVAC Equipment Area. Access controls and posting justifications for these areas will be discussed in the following sections.

4.4.5.2.1 Change Room

The change room is located off the main entrance corridor between the HPT Office and Shift Managers Office. The change room is the main personnel entry point to the CSB Operating Deck. After entering the Change Room, a Radiological Buffer Area (RBA) will be established for contamination control purposes. The associated airlock will also be posted as an RBA. A Radioactive Material Area (RMA) will also be established inside the Change Room for storage of radioactive material such as Anti-Contamination clothing. This room will be used to change into Anti-C’s when required by the applicable Radiological Work Permit (RWP). Surveillance of this area will be performed to verify effectiveness of controls.


4.4.5.2.2 RBA Exit Area

The RBA Exit Area and the associated airlock will be the typical personnel exit point from the Operating Deck. The airlock and RBA Exit Area will be posted as an RBA. The airlock will have an Eberline PCM-1B that will be used for personnel whole body exit surveys which are required after Contamination Area (CA) entries. The RBA Exit Area will have two HandiFoot Monitors, an Eberline HFM-6 and HFM-7A, that will be used to perform exit surveys from the RBA. Personnel decontamination will also be performed in the RBA Exit Area if the need arises. Routine surveillance of this area will be performed to insure adequate contamination and dose controls.


4.4.5.2.3 HPT Count Room

This room is located off the airlock associated with the RBA Exit Area and will be used for the counting of radiological swipes, air samples, and other materials. In addition, radioactive sources needed for radiological operations will be stored in this room. Due to the nature of these activities, it is necessary to post this area as a RBA/RMA.


4.4.5.2.4 Filter Room

This room contains two filter housing that are part of the Operating Area HVAC system exhaust. These filter housings each consist of a HEPA filter and a pre-filter. At the beginning of operations, the room will be posted as “Potentially Internally Contaminated.” If these systems are breached, then radiological conditions will need to be reevaluated and postings changed per criteria in HNF-5173, Tables 2-3 and 2-4. A routine surveillance program will be used to characterize the area and determine any needed posting changes that are needed to maintain regulatory compliance. Radioactive material other than the installed systems will not be stored in this room.

Formal access control will be not required for accessing this area normally.

4.4.5.2.5 HVAC Equipment Area

The HVAC equipment in this area contains two potentially contaminated ventilation systems; the Support Area HVAC (SHVAC), and the Operating Area HVAC (OHVAC). The OHVAC system consists of potentially contaminated ventilation ducting and an Air Handling Unit with a filter. The SHVAC has a HEPA filter housing containing a HEPA filter unit and pre-filter. External dose rates from the HEPA filter are not expected to be significant. At the beginning of operations, this area will be posted as “Potentially Internally Contaminated Systems.. .” due to presence of installed components that have the potential to become contaminated. If these systems are breached, then radiological conditions will need to be reevaluated and postings changed per criteria in HNF-5173, Tables 2-3 and 2-4. A routine surveillance program will be used to characterize the area and determine any needed posting changes that are needed to maintain regulatory compliance. Radioactive material other than the installed systems will not be stored in this area.

Formal access control will be not required for accessing this area normally.

.


4.4.5.3 Ouerating Area

4.4.5.3.1 Operating Deck

All access points of the operating deck will initially be posted as an RBAf'Potentially Internally Contaminated systems within.. .". There is a potential for contamination to be present when the cask lid is removed, on the exterior of the MCO once it is removed from the shipping cask, and during sample/weld station operations. Formal access control is required for accessing these areas to verify qualifications and training of personnel needing access per AP-14-004, Radiological Area Access Control (Figure 5).

The southeast comer of the operating deck may also be used for storage of radioactive material and for a work area for work with contaminated materials. It will be appropriately designated at that time. Formal access control is required for accessing this area to verify qualifications and training of personnel needing access per AP-14-004, Radiological Area Access Control.

The following sections will describe potential conditions and the corresponding postings for different areas and work evolutions for the Operating Area.

4.4.5.3.1.a Trailer Vestibule

This bay is located on the northwest comer of the operating deck and will be utilized to receive the MCO/Cask transport trailer. When an MCO is received, the bay will be controlled as a radiation area (RA). A radiological receipt survey will be performed to verify radiological conditions. Then the MCO/Cask is staged and prepared to be moved into the MCO Receiving pit. The area will be maintained as an RBA after the trailer is removed for purposes of contamination control.

Radiation and contamination levels in the Trailer Vestibule will be evaluated after the MCO/Cask shipping container has been removed from the trailer. A routine surveillance program will be used to characterize the area and determine any needed posting changes to maintain regulatory compliance.

4.4.5.3.1.b MCO Receiving Pit

Located on the north end of the operating deck, the MCO receiving pit is used to stage the MCO for removal from the shipping cask by the MCO Handling Machine (MHM). The MCO/Cask is removed from the transport trailer and lowered into the MCO Receiving Pit. The area surrounding the MCO Receiving pit will be controlled as an RA due to the dose rates from the exterior of the MCO/Cask.

After the MCO/Cask has been lowered into the pit, the head space between the shipping cask lid and the MCO shield plug will be sampled. Then the cask lid will be removed to allow access to the MCO. An IWCA will be established around the pit as a precautionary measure until contamination levels have been verified. Decontamination efforts by the CVDF should minimize the amount of contamination present on the top of the MCO. Routine surveillance of the annular space between the MCO and shipping cask will also be performed. This will verify the


effectiveness of the Immersion Pail Lid at maintaining a barrier between basin water and the IWTS Return water to prevent contaminating the outside of the MCO. The need to establish a CA during this work evolution will also be evaluated after data has been collected from a number of processed MCO’s.

Cover blocks will be put in place around the top of the pit, which will most likely eliminate the need for the RA control. The MHM will be positioned over the pit and will extract the MCO from the Cask. Shielding has been incorporated into the MHM design to reduce radiation levels from the MCO.

A routine surveillance program will be used to characterize the area and determine any needed posting changes that are needed to maintain regulatory compliance. During other activities in this area such as maintenance, it will be necessary to establish temporary radiological areas for control during these activities.

4.4.5.3.1.c Maintenance Pit

The maintenance pit located at the north end of the operating deck will be used to make repairs to the MHM grapple, which is not expected to have significant external dose rates. This area will be posted as a CA during maintenance operations because of the potential for contamination on the interior of the MCO chamber of the MHM rotating turret. Radiological surveys of the grapple will be used to characterize the area before, during and after maintenance to determine any necessary posting changes changed per criteria in HNF-5173, Tables 2-3 and 2-4. A routine surveillance program will be used to characterize the area and determine any needed posting changes that are needed to maintain regulatory compliance.

4.4.5.3.1.d SamDle/Weld Station

Stations #2 & #7 located at the south end of the operating deck are planned to be used for sampling/welding operations that occur at the CSB. Sampling will usually take place in pit #7 and welding in pit #2. During any sampling/welding work evolution, the area will be posted as a CAIRA.

In addition, depending on the ability to keep the MCO sampling hood free of contamination, it may be necessary to establish a contamination area at the hoods’ storage location for those times when the unit is not in use.

Posting for the sampling hood ventilation system, which will be “potentially internally contaminated”, is covered by the access posting on the entry way. If the ventilation system is breached, then radiological conditions will need to be reevaluated and postings changed per criteria in HNF-5173, Tables 2-3 and 2-4. A routine surveillance program will be used to characterize the area and determine any needed posting changes that are needed to maintain regulatory compliance.

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4.4.5.3.1.f MCO Storage/Sampling Tubes

The MCO Storage/Sampling tubes will not be posted as a High Radiation Area (HRA). Each MCO Storage Tube will eventually contain two MCOs and a shield plug will be placed into the tube. Each shield plug weighs approximately 5,000 pounds, which exceeds the requirements for posting each tube.

ITEM I3’cs Concentration 9 0 ~ r Concentration Total alpha Concentration Tritium Concentration Average Dose Rate Floor Sludge Volume

4.4.5.3.1.g Ventpurge Cart

The Ventpurge Cart will be used to monitor the pressure of the MCO overpack tubes and inert the tube atmosphere with helium if necessary. The cart contains a HEPA filtered ventilation unit that is used to ventilate the overpack tubes. Once the cart is used, it will be posted as “Potentially Internally Contaminated.” If these systems are breached, then radiological conditions will need to be reevaluated and postings changed per criteria in “F-5173, Tables 2-3 and 2-4

105 KW BASIN 0.16 microCi/L 3.7 microCiL 0.01 microCi/L 1.7 microCi/L 2.5E-3 rnicroCi/L 1.9E-2 microCi/L 0.07 microCi/L 3.0 microCi/L 0.1 mRem/h 5 mRem/h 129 feet3 1,480 feet3

105 KE BASIN

4.4.5.3.1.h MHM Rotating Turret Assembly

The MCO chamber of the MHM rotating turret will be posted as “Potentially Internally Contaminated” after handling the first MCO. The MCO chamber is maintained at a negative air pressure by the Cask Extract System that exhausts through high efficiency particulate air (HEPA) filtration whenever the MHM is powered up.

5.0 Historical Basin Radiological Comparison

The Table 5.1 compares the radionuclide concentrations, radiation fields, and sludge volumes in the 105 KW and 105 KE basin water and operating areas.

This demonstrates the beneficial effects of facility modifications, the encapsulation of fuel in sealed canisters, and better water chemistry control in the 105 KW basin versus the 105 KE basin.

(11.10)

Table 5.1: Radiological Comparison of Basins

I COMPARISON OF 105 KW AND 105 KE BASIN RADIOLOGICAL CONDITIONS

The table’s radionuclide concentrations are the typical in calendar year 1999, and the dose rates are from radiological surveys conducted in April 1999.


5.1

Samples are collected routinely and analyzed for radiological constituents in order to monitor performance of the water treatment system and to characterize spent water treatment components for disposal. The radionuclides of particular interest are the fission products 137Cs, 90Sr, tritium,

Pu. Due to the age of the fuel, short lived activation and the transuranic isotopes 241Am and products such as 54Mn, 59Fe, and to a lesser extent, Co, are no longer present in significant amounts.

Routine Water Analysis and Isotopes of Interest

2391240

00

5.2 Cesium 137

The isotope 137Cs is a medium-energy gamma emitter that is released directly from the uranium metal fuel as it corrodes. The I3’Cs embedded in the concrete is the primary source of radiation exposure in the KE basin. This “bath tub ring” was established in the first years of operation, and even ten (IO) years of maintaining good water quality had not lessened its impact on general area dose rates. Localized “hot spots” caused by 137Cs that is trapped in piping and valves also contributes to dose rates. However, the in the water is a secondary source contributing 3 to 5 mRem/h to the general dose rates.

The walls in KW basin were coated with epoxy paint so there is no “bath tub ring”, and 137Cs concentrations in the water are much lower. Cesium is present in the water as a highly soluble cation (positively charged ion). It is removed from the water by the cation resin bed of the IXMs. Only a small fraction is removed by filtration.

5.3 Strontium 90

90Sr is a high-energy beta emitter also released directly to basin water from the corroding fuel. Because the beta radiation is easily shielded, the ”Sr does not contribute much to dose rates in the basin. Its primary hazard is in the form of contamination and internal deposition.

5.4 Tritium

Tritium is a radioactive isotope of hydrogen and is present in the basin water as part of the water molecule. The water treatment system can neither filter or ion exchange the tritium out of the basin water. The concentration of tritium is a function of its release rate from the corroding fuel and evaporation or leakage from the basin. Fortunately, the concentrations of tritium found in K- Basins do not present a significant hazard in terms of dose rates, contamination, or internal

(11 1 1 ) deposition .

5.5 Transuranic Radionuclides

By definition, a TRU isotope has an atomic number greater than 92 (uranium), has a half-life greater than 20 years, and is an alpha emitter. The K Basins contain many TRU isotopes;

239’z40Pu account for greater than 95 percent of all alpha activity. In the KW however, Am and basin, the TRUs are mostly contained inside the fuel storage canisters and don’t present significant contamination problems. In the KE basin, where the fuel is exposed directly to the

241


basin water and a large volume of sludge is present, the TRU concentrations are several times higher and present many problems.

The chemistry of TRU isotopes is very complex. They are present as both cations (positive ions) and anions (negative ions) that can be removed by the IXMs. They can also be removed as large particles by filtration. Fin