- Released 1994

Prepared for the US, Department of Energy under Contract DE-AC06-76RLO 1830

Pacific Northwest Laboratory Operated for the U S . Department of Energy by Battelle Memorial Institute

Cover Sheet for a Hanford Historical Document Released for Public Avail a bi I i ty

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DISCLAIMER

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DUN-7060

HANFORD CATEGORIES. ' C - 5 7 A N D C - 6 5

NUCLEAR, INC,

R I C H L A N D , W A S H I N G T O N

. . * .!

PATENT STATUS

INFORMATION CONCERNING USE OF THIS REPORT

This document copy, since it i s transmitted in advance of patent clearances, is mode available in confidence solely for use in performance of work under contracts with the U. S. Atomic Energy Commis- sion. This document is not to be published nor i t s contents otherwise disseminated or used for purposes other than specified above before patent approval for such release or use has been secured, upon request, from the Chief, Chicago Patent Group, U. S. Atomic Energy Commission. 9800 So, C a s Ave., Argonne, Illinois, 60439.

PRELIMINARY REPORT This repor? contains information of a preliminary nature prepared in the course of work under

,Atomic Energy Commissiqn Contract AT(45-1)-1857. This information is subject to correction or modifica- tion upon the collection and evaluation of additional data.

- . . .. .. . -. . . "I .-

L E G A L ..N,OT.LCE. "

nor the Commission, nor ony person acting on behalf of the Commission:

Makes any warranty or representation, expressed or implied, with respect to the accuracy, com- pleteness, or usefulness of the information contained in this report, or that the use of any information, apparatus, method, or process disclosed in this report may not infringe privately owned rights; or

, , . This~epart was prepared QS an account of Government sponsored work. Neither the United States,

A.

B. Assumes any liabilities with respect to the use of, or for damages resulting from the use of any infor motion, apparatus, method, or process disclosed in this report.

As used in the above, 'person acting on behalf of the Commission" includes any employee or contractor of the Commission, or employee of such contractor, to the extent that such employee or con- tractor of the Commission, or employee of such contractor prepares, disseminates, or provides access to, any information pursuant to his employment or contract with the Commission, or his employment with such contractor.

PLEASE SIGN BEFORE READING THIS PUBLICATION

AEC-RL R I C H U I I W . W A S U

DUN-7060

Hanford Categories : c-57 & c-65

This document consists

October 19, 1970

DOUGLAS UNITED lWCLEAR

MONTHLY REPORT

September 1970

NOT UCNf DOUGLAS UNITED NUCLEAR, INC.

Richland, Washington

Work performed under Contract No. AT(k5-11-1857 between the Atomic Energy Commission and Douglas United Nuclear, Inc.

EJ C LAS SIF IED DUN-7060

AEC - RICHLAND OPERATIONS OFFICE

1-6. W. Devine, Jr.

AEC - WASHINGTON 7-9 0 F. P. Baranowski

Division of Production

AEX - SAVANNAH RIVER 10 0 N. Stetson

DU PONT - SAVANNAH RIVER 11-12. W. B. Scott

BATTELLE-NORTHWEST

13. F. W. Albaugh 14. R. L. Dillon 1 5 * J. W. Finnigan 16. J. J. Fuquay 17 e B. M. Johnson, Jr. 18 e L. C. Schmid

ATLANTIC RICHFIELD "FORD

19 W. M. Harty 20. L. M. Richards 21 e R. E. Tomlinson

I DOUGLAS UNITED NUCLEAR

22 e C. L. Abel 23. D. H. Bangerter 24 e 3. R. Bolliger 25 0 J. J. Bombino 26 (. C. E. Bowers

J. A. Cowan J. E. Ruffin

REPORT DISTRIBUTION

- 2-

DOUGLAS UNITED NUCLEAR

27 * 28. 29 * 30 31 32 33 34 35 9

36. 37 38. 39 * 110.

41. 42. 43. 44. 45. 46. 47 48. 49. 50. 5 1 * 52 53. 54 0

5 5 . 56. 57 0

58 59. 60. 61 62. 63. 64 e

P. A. Carlson W. G. Catts G. C. Coleman R. E. Dunn A. E. Engler J. M. Fox, Jr. G. C. Fullmer J. P. Hamric ( O a k Ridge) C. D. Harrington D. L. Hovorka R. T. Jessen V. V. Johnson H. P. Kraemer, Jr. C. W. K u h l m a n 3. W. Heacock A. R. Maguire R. T. Martell W. M. Mathis E. T. Murphy W. S. Nechodom D. W. Peacock R. S. Peterson C. F. Poor C. A. Priode C. A. Priode (Work Copy) T. Prudich T. B. Pugh R. W. Reid K. L. Robertson R. K. Robinson B. Schauss 0. C. Schroeder W. Seeburger R. H. Shoemaker Earl A. Smith J- T. Stringer W. K. Woods DUN File DUN Record

UNCLASSIFIED

U N C M S S I F I E D

TABLE OF CONTENTS

SUMMARY

REACTOR PLANT OPERATIONS

KE Reactor N Reactor

FUEL AND TARGET FABRICATION

KE Reactor N Reactor

TECHNICAL A C T I V I T I E S

ADMINISTRATION - GENERAL

APPENDIX

A. P ro jec t S ta tus Summary B. Employment Summary

DUN-7060

St ar t ing Page

A-l

E-1 BN-1

c-1 CN-1

D-1

E-1

F-1 F-4

- 3- UNCLASSIFIED

I IUCTOR PLANT OPERATIONS

Production S t a t i s t i c s

Input Production - Pu (KMWD)

Time Operated Efficiency - %

83.0

78.6

Steam Ava i l ab i l i t y t o WPPSS - % -

32.5

36.0

32.1

KE Reactor

The reac to r remained down i n cont inuat ion of t h e August extended outage ut9 September 4 when operation w a s resumed following AEC-RL s t a r t u p authorization.

Equilibrium power l e v e l w a s r e s t r i c t e d t o 3,750 MW by b r i t t l e f rac ture tube power l i m i t s . There were two outages and one s t a r t u p in te r rupt ion . One outage w a s t h e continuation of t h e August outage, and t h e other was caused by an unexplained Pane l l i t t r i p . The s t a r t u p in t e r rup t ion w a s caused by a f a u l t y tes t hole Pane l l i t gauge.

N Reactor

N Reactor w a s i n shutdown s t a t u s at t h e beginning of t h e month. Reactor operat ion resumed, following the extended summer outage, on September 14. There were four unscheduled reac tor shutdowns during t h e month, caused by t h e following: a l o w flow monitor t r i p ; a manual shutdown t o invest igate a suspected l eak through t h e primary loop dump valve, V4-4; a manual shut- down t o inves t iga t e an unexplained increase i n primary loop makeup ra te ; and a l o w flow monitor t r i p which occurred when number s i x dr ive turbine t r ippea t o pony motor operation.

I n the l a s t shutdown described above, t he automatic scram c i r c u i t deenergized t h e rod sa fe ty c i r e u i t , but no con t ro l rods entered t h e reac tor . The back-ug con t ro l system, t h e b a l l s a fe ty system, functioned t o shut t h e reactor down. Inves t iga t ion of t h e cause of t h e f a i l u r e of t h e rod sa fe ty system t o func- t i o n properly was s t i l l under inves t iga t ion at month end.

The r eac to r remained i n shutdown s t a t u s at month end. T o t a l outage hours for the month w e r e 472.0.

FUEL FABRICATION

Production S t a t i s t i c s ( tons ) For KE For N

B i l l e t s Extruded - 8.5 Finished Fuel Produced 34.3 37.2

A- 1

KE Fuel

AlSi f u e l s operation continued on a p a r t i a l - l i n e basis f o r t he production of K5E fue l .

N Fuel

Input and output production w a s below forecas t i n order t o assure delivery of spike f u e l t o t h e reac tor .

'TECHNICAL ACTIVITIES

A program has been designed t o establish a basis upon which t o determine t h e in-service l i f e expectancy and maintenance requirements of t he V-11 GRAFOIL valve packing mater ia l .

. .

, .. . : ~ .-. ,..- .I

. .

.

System modifications aimed at reduct ion of b a c t e r i a l contamination have been completed for t he l e f t ho r i zon ta l con t ro l rods on N Reactor, and a surveil lance program has been implemented. con t ro l rod mockup t o check materials compat ib i l i ty , f l u i d s t a b i l i t y , and bac te r i c ide effect iveness .

Tes t ing has been i n i t i a t e d on the horizontal

Fabricat ion of the prototype s u b c r i t i c a l d r i v e system for t he Nuclear Flux Monitor System a t N Reactor i s i n progress and preparat ion fo r environmental t e s t i n g of the dr ive system has been i n i t i a t e d .

Results of ex-reactor impact t e s t i n g have determined that N Reactor fue l has s u f f i c i e n t s t rength t o prevent excessive deformation i n the event of an i n l e t connector f a i l u r e accident.

A KE Reactor ;process tube treated by t h e e l e c t r o l y t i c hydride removal process i n February 1970 w a s found t o be leaking at the rate of 10 t o 12 gallons per day. Laboratory examination revealed t h a t t h e leak w a s caused by a circum- f e r e n t i a l crack a t the Van Stone f lange . The f lange w a s masked from treatment during t h e hydride removal operat ion last February. t h e f a i l u r e and t h e cleaning opera t ion i s apparent. t he tube i s i n progress.

No connection between A complete analysis of

Increased emphasis has been focused on developing a rec i rcu la t ing decontami- nat ion process f o r appl ica t ion a t N Reactor. DUN-7265, "Action Plan - N Reactor Recirculating Decontamination," w a s i ssued and out l ines a program t h a t will permit considerat ion of using t h e technique during next year ' s summer outage.

N Reactor graphite cooling heat exchanger No. 2 has been r ebu i l t , re ins ta l led , and returned t o service.

The N Reactor fog spray system has been func t iona l ly tested f o r pressure surges and ce r t a in c a s t i r o n components have been replaced.

A- 2

DUN- 706 0

Steam c o l l e c t i o n drums which are p a r t of t h e steam generators at N Reactor have been examined and defec ts from o r i g i n a l construction repa i red .

In te rmi t ten t once-through cooling f o r a shut-down N Reactor wa.- es tab l i shed t o e f f e c t r e p a i r s .

The current s t a t u s of t h e N Reactor Regulatory Program was presented t o t h e Hanford Subcommittee of t he Advisory Committee on Reactor Safeguards.

The new emergency coolant supply l i n e and emergency dump tank at N Reactor were t e s t e d fol lowing i n s t a l l a t i o n and accepted fo r reac tor use.

Further t e s t i n g t o i d e n t i f y t h e phenomena causing c i r cu la t ing r a w water in- s t a b i l i t y a t N Reactor and r e s u l t i n g interdependence of t h e two e l e c t r i c a l power sources, w a s no t performed during t h e recent ly concluded summer outage because of o t h e r p r i o r i t y outage work. Plans, procedures, and equipment a r e ava i lab le f o r implementation of t h e test program as outage t i m e i s ava i lab le .

Preliminary results have been evaluated on a confinement leak r a t e tes t at N Reactor. The l e a k r a t e w a s about 2.5 percent as compared t o 1 . 4 percent t he previous year .

Work continues on t h e contaminated coolant p i l o t plant . S ta r tup of t h e f a c i l i t y i s t a r g e t e d f o r October 1, 1970.

DUN waste management p rac t i ces regarding reac tor e f f luent streams were presented t o the ACRS Subcommittee on August 21, 1970.

I r r a d i a t i o n s completed during t h e repor t period included 42 quickie act iva- t i o n samples f o r BNW and WADCO, tantalum fo r a DUN program, and th ree t e n s i l e specimen capsules f o r BNF7. Customer i r r a d i a t i o n s were charged t o produce Dolonium f o r Mound and xenon-128 f o r ANL. ..

GENERAL

Three minority employees were added t o the work force during t h e month. I n t h e c l e r i c a l and r e l a t e d job category, 100 percent of t h e new-hire addi t ions i n t h e quar te r ending September 30 w e r e minori t ies .

There were no d i sab l ing i n j u r i e s i n September, and no r ad ia t ion exposures exceeded opera t iona l cont ro l .

Charles D. Harrington w President

A- 3

REACTOR PLANT OPERATIONS - KE!

PRODUCTION

General

Production ( l a r g e l y weapons grade Pu), power l e v e l s , e f f i c i e n c i e s and r e l a t e d s t a t i s t i c s f o r t h e JSE Reactor a r e tabulated below.

K r eac to r input production and time operated e f f ic iency (TOE) for t h e pas t s i x months are shown on t h e following chart:

300 r

240

180 I

120

60

. -.. . TOE ;

%. , /

- \ / * - - - - - - - - - - -

\ 0 \ / \ /

\ / V

II

--5

Product ion

0 100

80

1 60 - 40

- 20

St a t i s t i ea1 Summary

Input Production - Pu (KMWD) - U-233 (Equiv. MWD)

Power Level (MW) - Maximum - Average

Time Operated Efficiency - %

Number of Outages Number of Star tup In te r rupt ions

Operating Coolant Flow - 1000 gpm

Fuel Charge (Tons) - 94 Metal

Fuel Element Fa i lures

- Natural U

Helium Losses - 1000 cu . f t .

B-1

4J c a, 0 k a, PI

I w 0 w

83.0 0

3,750 3,519

78.6

2 1

201.7

27 215

U

187.1

OPERATIPU’G MPERIEMCE

Reactor Loading

The froEt f ace map showing t h e loading i n KE i s reproduced on page B-5. tonnages l i s t e d a re approximate; ac tua l fue l charge t o t a l s a r e tabulated on t h e preceding page.

The

Power Level

The power l e v e l at KJ3 Reactor was r e s t r i c t e d by b r i t t l e f r a c t u r e tube power l i m i t s dur ing t h e month.

Reactor Outages

Date Down Outage Hours Remarks

September i’? 86.0 A u g u s t outage continued.

September 14 ’* : 67.6 Unexplained P a n e l l i t t r i p . *~

September 17 0.5 Faulty P a n e l l i t gauge 3h t e s t hole.

The au thor iza t ion w a s received a t 0800 hours and operation w a s resumed a t 1400 hours.

August outage was continued u n t i l September 4 when AEC-RL s t a r t u p

EQUIPMENT EXPERIENCE

Ver t i ca l and Horizontal Rods

. . : . . :, :.. All HCRs continue t o be operable with no major binding problems. A l l VSRs a r e serv iceable with an average hot drop t i m e of 1.77 seconds.

Resistance Temperature Detectors

There were no RTD f a i l u r e s during t h e month.

Pane l l i t Scrams

Two P a n e l l i t scrams were experienced during t h e month. The f irst scram oc- curred on September 14 on Pane l l i t gauge 5571. w i t h t h e charge makeup, charge posi t ion, or hardware i n t e g r i t y . The toggle valve was found t o be leaking when depressed and i s bel ieved t o be t h e cause of t he scram. Eight rows of toggle valves were replaced because of aged O-rings. The second Pane l l i t scram w a s experienced on September 17 when the dr ive arm became disengaged from the gauge i n 3A t e s t hole.

No abnormality could be found

B-2

184-DA Boiler

The work of retubing the 184-DA b o i l e r by t h e J. A . Jones Company i s essen- t i a l l y complete and t h e b o i l e r i s scheduled t o be placed i n t o service during t h e week of October 5 .

Eaergency E l e c t r i c a l Power Backup

The program for an emergency power source f o r 100-K i s nearing completion. The i n s t a l l a t i o n of equipment (750 kw d i e s e l generator and No. 2 fue l supply for a KW b o i l e r ) w a s completed September 18. Acceptance t e s t s have been com- p l e t ed on t h e d i e s e l generator , but t h e b o i l e r cold start system has not been demonstrated. It i s planned during t h e next few weeks t o complete t h e demon- s t r a t i o n t e s t s of both t h e d i e s e l genera tor and b o i l e r cold start. Rough-draft emergency procedures have been prepared and will be f ina l i zed following t h e t e s t i n g program.

PROCESS ASSISTANCE AND CONTROL

Operational Physics

Operation was r e l a t i v e l y s t a b l e during September, with one unscheduled outage and a s t a r tup in t e r rup t ion around t h e middle of t h e month. The samaxium di? following t h e s t a r tup on September 4 af te r t h e prolonged August outage w a s made w i t h about 570 cmk i n rods and sp l ines .

The spike inventory i s high and poison requirements f o r t o t a l control are higher than i n the recent p a s t , due t o t h e low average exposure of the fue l . On discharge outages about 140 sp l ines are required f o r t o t a l control . ca lcu la t ions t h e s t r eng th of t h e b a l l system has been reduced t o account f o r t h e four b a l l channels which are out of se rv ice .

I n a l l

Minimal physics ass i s tance w a s required with t h e s t a b l e reac tor condition. The scram recovery on September 17 w a s made without ons i t e physics ass i s tance , using approved procedures.

Some operat ional physics parameters of i n t e r e s t are shown i n t h e following t a b l e :

Effec t ive Central Tubes* 2,126

F la t ten ing Efficiency** - September - 12-Month Average

Maximum Operating Time Permit t ing Scram Recovery - Hours***

.676

.TO2

10

Average C r i t i c a l Predic t ion Error - September (cmk) 1 1 4 - 12-Month Average 131

B- 3

*Reactor power level divided by the average power of the ten most productive tubes which are representative of the reactor loading.

**ECT divided by the number of power generating tubes.

***The maximum operating time subsequent to a cold startup following which a scram recovery could be made using the currently approved startup procedure.

. .

B- 4

Zone

Central

-

Buckled

Fringe

Shield Protect ion

Tons - 226 61

.23

82

Material

Natural Uranium 94 Metal Bismuth

Natural Uranium

Natural Uranium 54

23 .6

94 Metal - Thoria S u g o r t Lithium

Loading Pa t t e rn - KE Reactor

B- 5

. . * . .. .

Helium Losses (1000 cu . f t . )

Fuel O i l Usage ( b b l . )

OPERATING EXPERIENCE

332 1 7

24 , 596

Reactor Loading

The r eac to r loading a t month end i s shown on t h e f ron t face map t h a t follows on page BN-7. month .

No f u e l charging o r discharging w a s undertaken during t h e

Reactor Power Level

The maximum authorized r eac to r power l e v e l of 4000 MWt w a s achieved during equilibrium operation. a t i o n a t power l eve l s g rea t e r than 3600 MW thus permitt ing an e l e c t r i c a l generation by WPPSS of 800 MWe.

The main steam header pressure during reactor oper- w a s maintained above 123 ps ig ,

Reactor 0ut.ages

The r eac to r w a s i n shutdown s t a t u s a t t h e beginning of the month as a result of t h e scheduled summer outage extending from the previous month. Four un- scheduled outages were experienced.

Star tup from the scheduled summer outage w a s made at 0733 hours on September 1L. The reactor w a s automatically shut down by a l o w flow monitor t r i p a t 1303 hours on September 14, and operation w a s resumed a t 2240 hours on September 15. The r eac to r w a s manually shut down a t 1534 hours on September 19 t o invest igate a V-4 valve leak-through problem, with s t a r t u p a t 1918 hours on September 20. A t 1250 hours on September 27, t h e reactor was again manually shut down, t h i s t i m e t o i nves t iga t e an unexplained increase i n pr i - mary loop makeup flow. Star tup w a s e f f ec t ed a t 0222 hours on September 30. An automatic low flow monitor t r i p scrammed t h e reactor a t 0525 hours on September 30. The reactor remained i n shutdown s t a t u s through month end.

Unusual Condition

The r eac to r automatic scram on September 30 t r ipped t h e rod sa fe ty c i r c u i t , bu t no rods entered the reac tor . The backup control system, t h e ba l l s a fe ty system, functioned properly and the r eac to r w a s s a fe ly shut down. Inves- t i g a t i o n of t he cause of t h e fa i lure of t h e rod sa fe ty system and the ensuing correct ive act ion w a s not complete a t month end.

EQUIR4EYT EXPE'RIETJCZ

N Confinement Sysvem

During performance of t he annual confinement system t e s t , four s ign i f icant de f i c i enc ie s w e r e detected;

BN- 2

d

e

e

The backup c losure bag on steam vent CW-206-4 f a i l e d t o inflate due t o over-tightened b a l l de tec t screvs on the bag can i s t e r cover.

Visual inspect ion revealed t h a t seven steam vent valves were not sea t ing properly because t h e cable clamps on t h e valve s h a f t dr ive pul leys had s l ipped . A s a preventive measure, a l l clamps were removed, dressed, and reclamped t o assure a pos i t i ve connection.

The Zone I exhaust backup c losure ga tes leaked due t o a missing rubber s e a l on t h e nor th gate.

The No. 2 exhaust f an valve opened automatically when the "closed" solenoid valves were deenergized. All four solenoid valves were replaced and r e t e s t e d .

The system w a s successful ly t e s t e d and placed in to f u l l service.

N Rupture Monitor

The Equipment Maintenance Standard t e s t of t h e sample delay time was suc- ces s fu l ly performed using a newly developed 60-channel e lectronic data system. This new method el iminates the hand manipulation of each sampling channel 's valves , r e s u l t i n g i n a l a r g e saving of operator and monitor time and about 24 R per year i n r ad ia t ion exposure t o personnel.

Emergency Cooling System

The e n t i r e emergency cooling system valve gas actuating system giping, sole- noids, f i l t e r s , and dip l egs were blown down, cleaned, and repaired t o improve t h e opera t iona l r e l i a b i l i t y of t h e system p r i o r t o the s t a r tup from t he sched- uled summer outage. a l l y t e s t e d and demonstrated to be wi th in a l l Standard requirements.

Dump tank dra in down tes ts were scheduled following the performance of t h e dump tank acceptance test procedure t o provide information f o r analyzing the flow capaci ty of t h e new e f f luen t con t ro l system. Following s tar tup, t h e r e were indica t ions of leakage of hot water i n t o the dump tank, possibly from t h e V-4 o r RV-2 primary loop pressure r e l i e f valve leak-through. Tests were performed t o determine t h e source of t h e hot water and t o provide procedural methods f o r cont ro l of t h e dump tank temperature. Conditions at the dump tank (flow and temperature) w e r e s a t i s f a c t o r y f o r f u l l power leve l operation by month end.

Primary Loop

A l l t h e i n l e t (V-3) and ou t l e t (Y-4) valves were function-

The RV-2 primary loop pressure relief valves were functionally tes ted satis- f a c t o r i l y . V-4 valve leak-through w a s determined t o be nominal, as designed. PCSV-203-5, c e l l 5 i s o l a t i o n valve, f a i l e d t o open when t h e valve stem parted from t h e valve gate. It w a s necessary t o place the l e f t half of the reactor on once-through cooling i n order t o i s o l a t e t h e valve f o r repairs . A proce- dure w a s developed which provided for in te rmi t ten t once-through flow with

SN- 3

demineralized w a t e r , thereby reducing t h e r i s k of using l e s s than primary loop qua l i ty w a t e r i n t h e system. t h e valve bonnet replaced. on t h e PCS header i n c e l l 5.

The valve gate was removed from PCSV-203-5 and The system i s operating with no i so la t ion valve

A t o t a l of seventy (70) V - 1 1 valves was worked on by Construction forces during September. are as follows:

Total. valves worked on during the extended summer outage

1. Manual operators i n s t a l l e d only - valve not rspacked.

2. Valves repacked only - operators previously i n s t a l l e d .

142

47

3. Valves repacked and manual operators ins ta l led - 492

Total 681

Valves y e t t o be repaired:

1. I n s t a l l manual operator only 3

2. I n s t a l l manual operator and repack 22

3. Repack only - 52 . .

T o t a l 77

Two V-12 valves were replaced during t h i s report period; the t o t a l number replaced during t h e summer outage i s now 84. hub l eaks were repa i red f o r a t o t a l of 36 hubs repaired on 30 valves.

However, two add i t iona l Grayloc

Raw Water System

Development T e s t No. 233, Fog Spray Surge Tests, was performed on September 4 t o determine t h e automatic start mode f o r the d iese l driven fog spray pumps t h a t w i l l l i m i t surge pressure t o a maximum of not more than 267 p s i , and t o determine the effect iveness of t h e accumulator tank i n performing t h e function of a surge suppressor. system i n t e g r i t y .

Higher r a t e d flanges were in s t a l l ed t o improve t h e

184-N Turbine Generator

A General E lec t r i c tu rb ine engineer consultant was cal led i n t o help analyze tu rb ine governor problems. w a s loca ted and repaired. binding problem w a s el iminated, t h e linkage adjusted, and the turbine returned t o serv ice .

A leaking hydraulic flange i n t h e governor system Control l inkage w a s found t o be binding. The

Equipment Modifications

The following equipment modifications were completed:

BN-4

~24.1-26, "Relccation of Low Vacuum Trip and A l a r m Switches f o r Drive Turbine - 109-N," provided fsr relocat ing, for convenience and ease of monitoring, the low vacuum t r i p and alarm switches from near t h e tu rb ine exhaust plenum t o t h e d r ive tu rb ine control panel.

E!MP-453 , "Emergency Signals - 1310-N Building," provided for i n s t a l l a t i o n of an evacuation a l a r m system i n t h e 1 3 1 0 - N Building t o provide added as- surance of personnel awareness t o emergency conditions.

PDC-N-70-56, "W Platform Control Power Rewiring," provided f o r c i r c u i t changes i n t h e W platform control console t o permit one person t o t r ans fe r cont ro l power from the W console s t a t i o n t o the charge machine console with- out leav ing t h e cont ro l power i n "off" while changing s t a t i o n s .

MDC-N-70-62, "Rod Low Flow Automatic Reset," provided f o r i n s t a l l a t i o n of a va r i ab le time delay r e s e t i n t h e annunciator i nd ica to r s f o r l o w rod flow t o permit i d e n t i f i c a t i o n of t h e annunciator before it is obscured by t h e automatic r e s e t f ea tu re of t h e system.

MDC-N-70-66, "Emergency Telephone C i rcu i t Revision," provided f o r documen- t a t i o n of changes made i n the emergency telephone communications system t h a t e l iminated excess c i r c u i t s , reconnected deact ivated s t a t i o n s , and cor rec ted an erroneous i n t e r t i e .

MANUFACTURING ENGINEERING ACTIVITIES

EauiDment Maintenance Standards

One new N Reactor Standard w a s issued; of t h e 42 N Reactor Standards scheduled f o r prepara t ion , 17 have been issued, f i v e a r e being routed f o r f i n a l approval, 16 a r e being prepared, and four remain t o be s t a r t ed .

V-12 Valve Linear Actuator

An i n t e g r a l l i n e a r actuator has been designed and a prototype has been fabr i - cated. This ac tua tor combines t h e valve stem packing and d ive r t ing mechanism f o r t h e V-12 valves . Actuation i s e f f ec t ed by primary loop water ac t ing on t h e underside of a pis ton t o "divert", and higher pressure (about 210c) ?s i ) makeup water ac t ing on t h e upper s i d e of t h e pis ton t o "re turn t o nomal". The e n t i r e mechanism has only one s t a t i c s e a l .

The prototype i s curyently undergoing off-reactor t e s t s . Twelve addi t ional u n i t s a r e being fabricated.

P i p e Freeze Technique

Prototype too l ing has been t e s t ed t h a t has t h e capab i l i t y of f reezing and iso- l a t i n g t h e 1 0 5 - N stag-arm connectors i n less than s i x minutes. Liquid nitrogen i s used i n conjunc+,ion with an e a s i l y a t tached self-contained freeze she l l . The t o t a l weight of a f u l l nitrogen container and a t tached f reeze s h e l l i s l e s s than 150 pounds.

BN- 5

Welding Tool Development

An automatic welding head, for welding Grayloc hubs t o connector tubing, has been developed and constructed. Several t e s t welds have been made t o estab- l i s h welding procedures. The tes t welds have passed bend, t ens i l e , and radiograph examinations.

Grayloc Flange - Hub Repair Liners

A new r e p a i r technique i s being developed f o r badly cu t Grayloc flanges. This technique uses a t h i n s tee l l i n e r which i s inser ted i n t o t h e conical surface, thereby renewing t h i s surface and the f lange normal surface. welded t o the outer surface of t h e f lange.

The l i n e r i s

PROCESS ASSISTANCE AND CONTROL

Reactor power l e v e l following the summer outage w a s r e s t r i c t e d u n t i l September 22 t o 3000 M W t or less while measurement and adjustment of rupture monitor hold-up t i m e w a s accomplished.

A ca lcu la t ion of t o t a l cont ro l immediately following the b a l l drop substanti- ated t h a t t h e r e w a s adequate protect ion with 27.5 m k of inser ted rod and 56.8 mk o f i n se r t ed b a l l s f o r an overa l l shutdown margin of 46.8 mk.

Traveling wire flux monitor graphs made subsequent t o t h e September 14 s ta r tup demonstrated t h a t t h e res idua l b a l l s i n t he middle of the reactor are s t i l l present , though addi t iona l burnout i s indicated.

Physics a s s i s t ance w a s given f o r t h ree s t a r tups during t h e month, and t h e four th w a s handled by Operations personnel using establ ished procedures with- out o n s i t e coverage by the assigned phys ic i s t .

Some opera t iona l physics parameters* of i n t e r e s t are shown i n the following t ab le :

Effec t ive Central Tubes

F la t t en ing Eff ic iency - September - 12-Month Average

Maximum Operating T ime Permitting Scram Recovery - Hours

Average C r i t i c a l Prediction Error - September - 12-Month Average

*For def in i t i ons see page B-4.

830

0.83 0.81

rn

125 cmk 125

BN- 6

~ 7 ~ ' i ? 4 1 I I i i I I I I I I I I I I I I I : : 1 1 1 1 1 1 1 I l l i i l l I > 4 ,

Fuel No. Code Tubes Description -- E 220 Mk-IC (94 Metal - Fr inge) F 89 Mk-IV (94 Metal - High U-236) G 444 Mk-N (94 Metal - Centra l ) N 1 Mk-IB (Natural U) X 231 Mk-IA & Mk-IV (125 & 94 Metal) -

985 To ta l

PT-NR NO. No. Tubes Description

949 7 Mk-IV Demonstration

-- 113 5 Evaluation of the

End Spider Inner Sup-por t System

136 6 Mk-IV - Nonbonded End Closures

1 Graphite Sample - Channel

19 Total PTs 19 T o t a l PTs

1,004 Grand T o t a l

*Includes Mk-IV high U-236 content f u e l and 6 tubes with Mk-IV-AA 125 Metal and Mk-IV 94 Metal.

Loading Pa t t e rn - N Reactor

BN- 7

DUN-7060

FUEL FABRICATION - KE REACTOR

PRODUCTION

General

Production of AlSi-bonded f u e l f o r KE Reactor w a s 103.9 percent of forecas t . One hundred percent of t h e elements produced had self-supports attached.

Acceptable Fuel Elements Produced

Tons 34.3

Yields: Current Month - % FY t o Date - %

Month-End Inventor ies - Tons

Bare Uranium Cores

Finished Fuel: AlSi-Bonded Hot-Die-Sized

92.0 92.4

830*

1,411" 14

Thoria Elements 1 4

*These t o t a l s include 73 tons of bare cores and 95 tons of f in i shed f u e l i n s ized used i n t h e smaller reactors .

OPERATING EXPERIENCE

The ove ra l l opera t ing e f f i c i ency f o r t he month w a s 99.7%. t i m e w a s charged t o equipment malfunctions.

A l l of t h e down-

The AlSi f u e l s operat ion continued on a p a r t i a l - l i n e (about 3/4) basis f o r the production of K5E f u e l .

EQUIPMENT EXPERIENCE

The #5 Acme Gridley l a t h e has been too led t o ream t h e base of K5E cans t o re- move a fo ld defec t area, and t h e operation has been successful thus far. Over 10,000 cans have been reamed with one reamer. It w a s necessary t o use a lubr i - cant on t h e t o o l t o obta in a cons is ten t ly smooth surface; however, t he lub r i can t caused t h e chips t o s t i c k i n t h e base of t he cans and they could not be removed w i t h an a i r j e t . It w a s necessary t o design a spec ia l ca.n basket holder so t h e cans could be degreased ( t o remove ch ips) manually before being processed i n t h e cap and can cleaning machine.

PROCESS ASSISTANCE AND CONTROL

Mothing s i g n i f i c a n t t o repor t

c-1

FUEL FABRICATION - N REACTOR

PRODUCTION

Input Production

To ta l B i l l e t s Extruded 62

Tons Extruded Percent of Forecast

8.5 24.2

Output Production

To ta l Finished Fuel Assemblies 1,440

Tons Output Percent of Forecast

uranium U t i l i z a t i o n - %

Month-End Inventor ies - Tons

Bare Uranium Bi l le t s Finished Fuel

37.2 79.0

IF 81.4

130 194

OPERATING EXPERIENCE

The September forecas t attainment w a s s a c r i f i c e d t o assure the del ivery of about four tons of spike f u e l required by t h e r eac to r . I n addi t ion t o t h i s inner spike run, t he shop continued processing pre-extruded material and completed 80% of t h i s mater ia l .

EQUIPlIIENT EXPERIENCE

The Campbell r ecu t s a w w a s provided with a s e t of quick-change s tops t o accom- modate a l l cur ren t recut f u e l lengths . s tops w i t h incremental spacers t o compensate for var i a t ions such as those caused by hea t t reatment , p r i o r machining, or pos i t i on i n t h e f u e l assembly.

The set cons i s t s of s ix bas ic length

PROCESS ASSISTANCE AND CONTROL

Mark I ou te r support inventory includes 80,000 supports tha t were purchased when t h e nominal Mark I height spec i f i ca t ion was 0.125 inch ins tead of t h e pre- s en t 0.128 inch. crown rad ius and crown length. The crown height (measure of crown rad ius ) over a 0.400-inch chord i s spec i f i ed as 0.009 t o 0.011 inch. t h e s e supports ranges from 0.00'75 t o 0.0085 inch.

The supports a r e a lso out of spec i f i ca t ion with respect t o

The crown height on

C N - 1

. . - . .: :

T e s t s have ind ica ted t h a t t h e use of t h i cke r shoes w i l l compensate for t h e s l i g h t deviations from sgec i f i ca t ions i n t h e support dimensions. It i s , therefore , planned t h a t t h e subjec t supports w i l l be used under control led conditions with spec ia l ly shimmed shoes t o provide acceptable support height on the f in i shed f u e l assemblies. cos t savings of about $15,000.

Use of these supports w i l l r e s u l t i n a

. .

I .

CN- 2

TECHNICAL ACTIVITIES

N REACTOR OPERABILITY PROGRAM

Primary Coolant System

V-11 Valve Packing Life Test

A program designed to establish a basis upon which to determine the in- service life expectancy and maintenance requirements of the new GRAFOIL V - 1 1 valve packing material has been initiated. Test fixtures have been designed and fabrication started. Procurement of reactor tested packing material for use as standards during testing is in progress.

Horizontal Control Rod (HCR) System

System modifications to reduce the bacteria1 contamination of the horizontal control rod invert emulsion hydraulic fluid were authorized by Design Change 3098, "Water Oil Invert Emulsion Fire Resistant Hydraulic Fluid for 105-N ECR Drive," and have been completed f o r the left bank horizontal control rods. A surveillance program authorized in PT-NR-153 , DUN-7037, "HCR Invert Emulsion Hydraulic Fluid," has also been implemented. Rod timing is within Equipment Maintenance Standards, and bacterial concentration is diminishing as predicted. Testing has been initiated 3n the HCR mockup to demonstrate an effective bactericide, and will be continued over the next sev- eral months to cheek materials compatibility, fluid stability, and bactericide effectiveness.

Temperature Monitor Systems

No system problems have occurred.

During the extended summer outage, 107 all tube temperature monitor and 61 zone temperature monitor resistancetemperature detectors (RTDs) were repaired.

A development test, DUN-7199, "Development Test Authorization 227, Temperature Monitor Instrumentation Inside N Area Thermo Barrier 105 Building Outlet Site," authorizing the monitoring of rear pipe space, process tube, and structural steel temperatures has been issued. The data will provide a basis for the specifications for strap-on RTDs used in the All Tube Temperature Monitor System.

Nuclear Flux Monitor System

11 In support of Project DCE-539, a schedule of approximately 1 1 5 drawings showing the detailed design of the upgraded nuclear instrumentation system has been made,

Fabrication of the prototype subcritical drive system, using water as the hydraulie fluid, is in progress, and preparation for environmental testing of the drive system has been initiated.

Upgrade Flux Monitor System," dated May 15, 1970,

D- 1

. .

I

FUEL PERFOFMANCE PROGRAM

N Reactor Fuel S t rength Studies

A t e s t i n g program t o determine t h e s t r eng th r e uirement of t h f u e l under accident condi t ions has been completed. The l i m i t i n g accident case iden t i f i ed was t h e failure of an i n l e t connector. I n t h i s event, reverse flow i n the associated process channel would d isp lace t h e f u e l and spacer charge over t h e gap t h a t e x i s t s between t h e most upstream spacer and i n l e t nozzle cap, and would exe r t dynamic loads upon t h e i n l e t nozzle and cap assembly, t he spacer charge, and t h e fuel.

Previous experiments have demonstrated t h a t , under conservatively estimated k ine t i c energy imparted t o t h e f u e l by reverse flow, t h e energy absorbed by the spacer charge upon impact with t h e I n l e t nozzle cap would be su f f i c i en t t o prevent damage t o t h e cap and t h e m i l - j o i n t between t h e i n l e t nozzle and process tube+ I t had not been adequately praved, however, t h a t (1) t he f u e l would withstand t h e impact forces without loss o f cladding in t eg r i ty , or ( 2 ) the spacers would not deform i n such a manner t o obstruct coolant flow thrsugh the i n l e t nozzle por t and c u t through t h e break.

These p o t e n t i a l problems have been resolved by t h e recent tes ts which experi- mentally s imulated t h e accident condi t ions as c lose ly as possible i n ex- r eac to r t e s t i n g f a c i l i t i e s . These experiments consis ted of impact t e s t i n g prototypic fuel-spacer columns i n a v e r t i c a l drop tube and evaluating r e s u l t s by examination of t h e prototypic f u e l and flow t e s t i n g of t h e spacers.

A comprehensive s e r i e s of impact tes ts w a s c a r r i e d out under conditions more severe than those ca lcu la ted f o r an i n l e t connector failure accident with t h e nominal. 28-inch gap between the upstream spacer and i n l e t nozzle end cap. r e s u l t s and conclusions were as follows:

The

0 P l a s t i c deformation occurred i n dummy f u e l elements having cores with y i e l d s t rengths l e s s t han 5,000 ps i . With adjustments made t o account for lower c lad s t r eng ths a t reac tor operat ing temper- atures, a core y i e l d s t rength of 9,200 p s i would be required t o prevent excessive deformation. I n con t r a s t , t he minimum in-reactor expected y i e l d s t rength of uranium under t h e same conditions would be g r e a t e r than 20,000 p s i . This i nd ica t e s a safe ty f ac to r grea te r than two wi th respect t o excessive c l ad deformation under the as- sumed accident conditions. These results and conclusions a r e based upon t h e observation t h a t only t h e outer fuel cyl inders would be deformed by impact. DefOrmatiGn of t h e inner fue l s d id not occur because most of t h e i r k i n e t i c energy w a s transferred t o t h e outer elements by t h e locking c l i p mechanism. Excessive displacement of t h e inner cyl inders w a s prevented by buckling of t h e adjacent downstream spacer.

0 All spacers were bowed and p l a s t i c a l l y deformed by t h e impact. The maJor amount of buckling occurred i n t h e spacer adjacent t o t h e f u e l ; consequently, t he re would be minimal po ten t i a l f o r f low blockage at t he por t l ead ing t o t h e i n l e t connector. Tests

D- 2

of deformed spacer columns i n t h e f u l l s ca l e flow loop indicated a maximum flow reduction of only seven percent, which i s considered t o be neg l ig ib l e -

0 The r e s u l t s of t h e s tud ies have shown t h a t , with t h e current spacer design and f u e l charge makeup, t h e mechanical and hydraulic aspec ts of flow reve r sa l s i t ua t ions would not pose any unresolved sa fe ty quest ions with respect t o f i s s i o n product confinement.

Dichromate Control as a Function of Fuel Surface Temperature

Analysis o f t e s t r e s u l t s (DUN-7166, "K Reactor Dichromate T e s t Results," Ju ly 29, 1970, Secre t ) has i d e n t i f i e d a r e l a t ionsh ip between f u e l c lad corro- s ion and f u e l surface temperature and f u e l residence time as a funct ion of dichromate concentration. Based on t h e data , dichromate concentration should be adjusted upward whenever f u e l surface temperatures exceed 120 C and t h e exposure t i m e i s g rea t e r than 70 operat ing days. For f u e l sur face tempera- tures g rea t e r or l e s s than 120 C , but with exposure t i m e under 70 operat ing days, no requirement for adjustment of t h e dichromate concentration i s indicated.

I n order t o f a c i l i t a t e adminis t ra t ion of such a requirement, cor re la t ions a r e being developed t o enable ready determination of f u e l sur face temperatures from t h e pe r t inen t r eac to r operat ing parameters.

/

PLANT LIFE PROGRAM

K Zircaloy Tube Hydriding

A KE Reactor process tube (3058) treated by t h e combined ( including hydrogen f l u o r i d e ) e l e c t r o l y t i c hydride removal process i n February 1970 w a s found t o be leaking a t t h e r a t e of 10-12 gallons/day. r e a r e ight inches of t h e tube by s e l e c t i v e pressure t e s t i n g e During tube removal t.he nozzle w a s stuck t o the gunbarrel f lange s tuds and subjected t o considerable impact s t r e s s before it broke loose. A c i rcumferent ia l crack was observed a t t h e Van Stone flange. The crack surface had s u f f i c i e n t f i l m buildup t o ind ica t e t h a t it w a s most l i k e l y not caused by t h e removal operation but w a s t h e source of t h e operat ing leak. The f lange w a s not t r e a t e d during the hydride removal operat ions l as t February.

The l eak w a s i so l a t ed t o t h e

N o o ther penetrat ions of the tube were found i n the l as t e igh t inches of t h e tube i n v i s u a l examination i n the laboratory. found a t a dent in t h e tube at about t h e 15-inch poin t , All ind ica t ions are t h a t t h i s damage was caused during tube removal. A complete ana lys i s of t h e downstream 60 inches of the tube is underway. and t h e cleaning operation has been found a t t h i s time.

N B a l l Graphite Channel 3enovation

A f r e s h crack (no f i lm) w a s

N o connection between t h e failure

Physical renovation of t he scheduled b a l l channels w a s completed on August 27, u t i l i z i n g t h e procedures authorized by Design Change 3089. Acceptance t e s t i n g

D- 3

of the renovated channels was completed successfully in accordance with Equip- ment Maintenance Standard N-103-120 A detailed report of' the complete channel rencvation work is in preparation,

DECONTAMINATION PROGRAM

Secirculating Decontamination Process Development

The emphasis on developing a recirculating decontamination process f o r appli- cation to N Reactor was increased during the manth. This increased emphasis will permit consideration of the use of the recirculating technique f o r decon- taminating the reactor at the beginning of next year's long maintenance outage.

In t.he previcus four summers, N Reactor piping has been decontaminated utilizing 8 h:t, inhibited phosphoric acid solution at approximately 10,000 gpm single pass* Because of limited capacity for storing hot water, concentrated chemi- cals, and wastes , ma.ximm chemical contact time with this technique is about thirty minutes

Althcugh these decontaminations have been considered generally satisfactory, xhe post-decontamination radiation levels have been successively higher each year Data show +,he effectiveness of the process on exposed carbon steel sur- faces t3 be the same as always, i.em, essentially ccmplete removal of deposited act iTtity7

Part sf the increase in post-decontamination radiation levels is attributed to incamplete cleaning of crevices and semi-stagnant areas at locations such as Grayloc ccuplings and ends of process tube nozzles? By recirculating the decc,ntamination solGtion, the time of application can be extended f o r at least several hours and, if necessary, for as long as a few days. This longer appli- cati3n period may produce significantly bctter results at hard-to-clean areas. Further, by utilizing a neutral pH e i t r i c acid--EDTA or M'PA reagent--corrosion of reactor materials during the longer perfd should s t i l l be well within safe ranges

An acficn plan for develcping a recirculating process that can be considered for use in a full-plant application in May 1971 was issued as DUN-7265, "Amnion Plan - N Reactor Recirculating Decontamination," September 18, 1970.

- OPERATIONAL SUPPORT PROGRAM

Intermittent Once-Through Cocling Procedure for Isolation and Repair o f PCSV-203-5 - N Reactor In August, a steam generator cell isolation valve (PCSV-203-5) failed upon routine testing Late in t.he extended summer outage, Because the valve (at the cell cutlet 1 is in direct ccntact. with t.he inlet risers, the major problem was i n isolating IC frcm the re8ctoT- core during repair. The solution adopted was t c sectionaiizc- the left and right sides of the reactor, then maintain circu- l a t i c n 9n t he righr side a.nd intermiEt.ently intrcltiuce single-pass cooling to

D- ii

t h e l e f t s ide with the h igh - l i f t d i e s e l pumps through the emergency cooling system piping. nuclear s a fe ty requirements, it w a s c a r r i e d out as a production t e s t (DUN-7240, PT-NR-232, " Iso la t ion of PCSV-230-5 Using In t e rmi t t en t Once-Through Cooling").

The in te rmi t ten t cooling procedure w a s considered t o be required i n order t o enable e f f i c i e n t u t i l i z a t i o n of t h e capac i ty of t h e demineralized water f a c i l i - t i e s . It involved es tab l i sh ing t h e c o r r e c t proport ions between the per iod of t i m e during whieh cooling w a s suppl ied and t h e per iod of flow in te r rupt ion . This required analysis p r i o r t o t h e tes t t o assure t h a t :

To assure t h a t t h e procedure w a s f u l l y consis tent with

0 A "ratcheting" temperature e f f e c t would not occur, i n which event t h e temperahwe decrease during t h e cool ing period would be l e s s than the temperature increase t h a t had occurred during the p r i o r per icd of flow in t e r rup t ion ;

a The pressure d i f f e r e n t i a l a f t e r f lcw w a s res tored would be ade- quate ZCI r ees t ab l i sh cooling i f b o i l i n g and tube voiding occurred during a nonflow period;

e Heat-up r a t e of t he f u e l during t h e flow i n t e r r u p t i m s wouid not be excessive,

Because of t h e low heat generation of t h e f u e l t h a t ex is ted a f t e r t he long shutdown period p r io r t o the t e s t and t h e low temperature of the graphi te , it was poss ib le 'GO develop tes t procedures t h a t would be p rac t i ca l from t h e oper- a t i o n a l standp3int and s t i l l be conserva t ive with respect t o the foregoing considerat ions. The t e s t w a s c a r r i e d out successfu l ly . Based upon temperature measurements with in-core thermocouple probes, t h e following observations have been made:

0 Natural convection w a s not as e f f e c t i v e as o r ig ina l ly expected i n removing heat from t h e a c t i v e zone.

0 The predicted f u e l heat-up rate w a s very conservative, Suf f ic ien t heat t r a n s f e r t o t h e moderator (which w a s a t about 85 F during t h e t e s t ) prevented t h e stagnant water i n t h e active zone from ex- ceeding 104 F i n t h e high f lux reg ion ( cen te r of column) with w a t e r shutoff times up t o one hour.

High Speed Scanner - KE Reactor F The high speed scanner i n s t a l l e d a t KE Reactor measures, displays, and a c t s upon s igna l s received from indiv idua l pmcess tube o u t l e t water temperatures. Data accumulated s ince i n s t a l l a t i o n of t h e high speed scanner ind ica te t h a t c e r t a i n trimming potentiometers o r i g i n a l l y i n s t a l l e d by t h e vendor have f a i l e d a f t e r t h r e e ts s i x months of operat ion. s e t a b i l i t y and f a i l u r e of t he t r impot t o r e t a i n i t s s e t t i n g .

The new state-of-the-art potentiometers ( t r impots 1 i n s t a l l e d i n the KE high speed scanner by production t e s t have net a l l requirements, P. number of new potentiometers a r e being procured cn a requi red design bas is for addi t iona l

The f a i l u r e mode i s usually loss of

D- 5

t e s t i n g i n more c r i t i c a l c i r c u i t s , Monitoring of potentiometer performance i s continuing . Demineralized Water Storage Tank Membrane Removal

High p u r i t y demineralized and degasif ied water f o r N Reactor i s s tored i n a one mi l l i on ga l lon demineralized water storage tank (DWST). The i n t e r i o r of t he carbon steel tank i s painted with a three-coat phenolic r e s i n appl lca t ion t o prevent i r o n pickup by t h e water. The tank has a l s o contained a rubberized f ab r i e bag (usua l ly c a l l e d t h e membrane) fastened t o t h e top of t h e tank w a l l t o prevent oxygen from t h e a i r dissolving i n the water.

,.

In recent yea r s , cemented seams i n t h e membrane have become weakened by expo- sure t o t h e warm water. pe r s i s t ed such t h a t c sn t inua l pumping has been required t o keep t h e membrane f-cm s inking"

Engineering s t u d i e s have determined t h a t t h e best a l t e rna te method of keeping qyger , cu t of t h e s to red water i s t o apply a nitrogen blanket t o t he tank. Wcrk h a s begun on i n s t a l l a t i o n of t h e equipment, but procurement delays have Fcstponed b e n e f i c i a l use.

During t h e r ecen t summer outage, t h e leakage r a t e of t he membrane exceeded t h e pumping rate and t h e membrane sank, It was decided t o remove the membrane, doing without an 3xygen barrier u n t i l t h e nitrogen blanket can be put i n s er:ri c e

Several r epa i r s have been made but t h e leakage has

1

Rebuilding; of Graphite CIsoling System Heat Exchangers

The g raph i t e madera tax cf N Reactor requires addi t ional cooling t o cont ro l graphi te temperatures; A l c w pressure, low temperature r ec i r cu la t ing cooling syszem was provided f o r t h i s function. The four U-tube heat exchangers o r ig ina l ly i n s t a l l e d i n t h i s system have a h i s t c ry of tube failures. A re+ubing Frcgram w a s i n i t s i a+ed and i s current ly underway.

Th? first un i t . ;Graph i t e Cooling Heat Exchanger Noo 2 ) has been removed, disassembled, redesigned, r e b u i l t , r e i n s t a l l e d , t es ted , and returned t o serv 1 ,ae

Disassembly confirmed That t h e tube leaks were due t o a wearing through of the tube wal l s a t t h e tube supports ( b a f f l e p l a t e s ) . t h e r i g i d i t y of t h e tube suppDrt system w a s inadequate. The support system w a s redesigned and t h e unit w a s r e b u i l t with new tubing of Inconel 600.

It w a s determined t h a t

When The unit w a s r e i n s t a l l e d and t e s t e d , no tubing vibrazion could be detec- t e d a+, any flow. The un i t has been returned TO service and a second u n i t i s i n the process of being r e b u i l t ,

105-109-N BlJildings Fog Spray System

The f3g spray system suppl ies t h e gama monitor system and Zone V a i r condi- t l cn ing coc i ing during normal operation and an emergency demand supplies graphize c-o;ing backup, i o9 fcg spray, io5 fog spray, p a r t i a l i o 5 f i r e

D- 6

protec t ion , and backup t o t h e con t ro l rod and thermal shield cooling systens. Normal demand i s suppl ied by t h e e lec t r ic -dr iven "high pressure r a w water" pumps while t h e emergency demands are diesel-driven "fog spray" pumps.

During rout ine t e s t i n g of t h e diesel-dr iven fog spray pumps, i n t h e l a t t e r p a r t of t he extended summer outage, high speed cycling of one of t h e d i e se l d r ives occurred. Concurrently, a c a s t i r o n reducing flange at a s t r a i n e r i n t h e l O 5 f i r e p ro tec t ion system f a i l e d . A review of' t he system revealed t h a t cn one occasion i n t h e pas t a c a s t i r o n reducing flange i n a similar pos i t ion had a l s o f a i l e d and t h a t t h e r e were s i x addi t iona l ca s t i ron reducing f langes s t i l l i n the system. These f langes were replaced with higher ra ted steel f langes p r i o r t o s t a r t u p .

A series of surge tes ts w a s run t o evaluate t h e e f f ec t s of various s t a r t i n g modes of t h e d i e s e l d r ives on t h e system, The t e s t s simulated l o s s of BPA power, and operat ion of t h e system without t h e fog spray accumulator. Loss of pcwer r e s u l t s i n a shock wave from rap id closure of t he high pressure r a w water pump check valves , and a measurable pressure surge results from t h e full speed s t a r t u p of t h e fog spray d i e s e l d r ives . The fog spray accumulator w a s shown t=, a c t as a surge tank on zhe system during t h e d iese l starts.

Although a number of components a r e underrated i n the fog spray system and the f u l l intent. of t h e piping code i s not met, examination of the system design, materials, and opera t ing condi t ions provides assurance t h a t an adequate margin of s a fe ty e x i s t s t o permit continued operation. Code Deviation Record No. 27 w a s prepared and approved cn t h i s basis t c authorize continued operation of the system,

Steam Collect ion Drums

A t N Reactor, t h e r e are two steam co l l ec t ion drums attached t o each s t e m generator. There are two steam generators per c e l l and a t o t a l of s i x steam generat ion c e l l s , Concurrent with t h e steam generator retubing program, the steam co l l ec t ion drums were found t o have a number of i n t e rna l surface defects at t h e 42-inch steam header connection where blanks had been welded fo r t he o r i g i n a l construct ion hydros ta t ic l e a k tests. It w a s found tha t a l l steam co l l ec t ion drums i n t h e o r i g i n a l c e l l s (1-5) had such defects and required r e p a i r , Cells 1 through 4 w e r e previously repaired during the stem gener- a t o r tubing t a s k , The defec ts i n Ce l l 5 were repaired, tcgether with the addi t iona l defec ts found i n t h e C e l l 3 u n i t s during the summer outage. The remainder of t h e c e l l s were found t o be in sa t i s f ac to ry conditicn.

Terminal Marking of Contaminated B u r i a l S i t e s

Design c r i t e r i a have been prepared t o provide terminal markers fo r underground contaminated waste b u r i a l s i t e s no longer considered act ive. The b u r i a l s i t e s under considerat ion are loca ted i n 100-D, 100-F, and 100-H Areas.

N Reac t iv i ty

I n i t i a l c r i t i c a l v e r i f i c a t i o n and excess i n rods as equilibrium was approached both ind ica ted agreement with t h e an t i c ipa t ed excess r eac t iv i ty o f 4-5 nk at

D-7

equilibrium, discharge plans i n t o June 1971 ind ica t e ample r e a c t i v i t y barr ing unforeseen b a l l drop losses i n t h e inter im.

Prs jec t rcns with t h e TRUE-K code based on t e n t a t i v e charge-

REGULATORY PROGRAM

Aagust Regulatory Meeting

. . . ' . . . .

.. I . .

The Hanford Subcommittee of t h e ACRS met a t Hanford on August 20 and 2 1 t o review reac tor s a fe ty experience, t h e N Reactor Regulatory Program, and Hanfcird waste management p r a c t i c e s - Regulatory Program presentat ions t o these regulaxory agency representayives included: 8 review of t h e p lan t s t a tus follcwing i n s t a l l a t i o n of t h e Eff luent Control Pro jec t ; a s t a t u s repor t and t c u r of Effluent Control Pro jec t i n s t a l l a t i o n s ; a l i s t i n g of t he resolved x s u e s , outstanding ques t ions , and pragram cantent t o resolve these outstanding quest ions; and a s t a t u s r epor t of se lec ted t.echnica1 inves t iga t ions . The pur- pose of t h i s presenta t ion w a s t o give the ACRS a bas is f o r evaluating program progress t o show comparabili ty between AEC-Division of' Production and l icensed r eac to r s

Informal supplementary discussions were held between AEC-DRL and DUN personnel cn August 19. These discussions provided valuable guidance on Regulatory Program scope and content. I n addi t ion, it w a s evident t h a t a misunderstanding exist .s on the 1-equired depth and breadth of Regulatory Program studies .

GCP-hll - Effluent Conrrsl P ro je s t

The results of t he acceptance tests f o r t he new emergency coolant supply l i n e and emergency dump Tank were reviewed and these systems were accepted f o r r eac to r use

The gravi ty drains and d i sposa l basin port ion of t h e pro jec t w a s not funded by t h e Atzmic Energy Commission f o r budgetary reasons. Consequently, t he emer- gency dump tank ^Jutflow i s cu r ren t ly routed t o t h e ex is t ing N Reactor c r ib and drainage canal through t h e ex i s t ing 36-inch low pressure f lush [FL(LP)] drain l i n e . This permits b e n e f i c i a l use of t he dump %ank and provides control of ncble gases which would otherwise be released d i r e c t l y t o t h e environment through +,he emergency :solant system l iqu id e f f luent path. Tests have shown t h a t , t h e ex is t ing 36-inch FL(LP) Line and t h e c r i b w i l l su i tab ly accommodate flcwa of approximately 25,000 gpm w i t h the hydraulic head ava i lab le from t h i s i n s t a l l a t i o n . Since t h i s outflow capabi l i ty i s l e s s than f u l l discharge capa- c i t y of t h e emergeney coolant system, procedural ac t ions have been established +,c cb ta in benef ic ia l use i n t h e in te r im pericd u n t i l +,he disposal basin and g rav i ty dra in pro jec ts can be completed.

IEEE-279 Evaluation

The designs of N Reactor pro tec t ion systems a r e being evaiuated f o r compliance wi+h t h e IEEE-279 c r i t e r i a prepared for nuclear power plant pro tec t ion systems. These evaluations show thar, t h e designs of the N Reactor protect ion systems genera l ly comply w i t h t h e provisions of t h e IEEE-279 c r i t e r i a . Apparent

D- 8

I ..

. . ... : . ..

ncnccmpliance with the c r i t e r i a i s genera l ly l imi t ed t o three p r inc ipa l design fea tures . These are:

1. System designs where f a i l u r e of a s ing le component can cause f a i l u r e of system act ion. (This does not apply t o automatic shutdown pro tec t ive systems.)

2. Lack of physical separat ion of system channels such t h a t system funct isn could be l o s t due t o common environmental consequences.

3. System bypass fea tures which do not automatically remove t h e bypass function whenever permissive ccndi t ions are not met.

With these evaluaticns cf ind iv idua l system compliance with c r i t e r i a , it is now pnssible t o evaluate t h e impact c f system noncompliance.

The N Reactor protect ion system design emphasizes redundancy of systems ra rher than redundancy and independence sf ccmponents within each system. Consequently, loss of system funct ion does not necessar i ly mean l o s s of pro tec t ion function es tab l i shed , plant modifications w i l l be necessary t3 provide f u l l comFli- ance with the in t en t of the IEEE-279 c r i t e r i a .

In those instances where t h i s re la t ionship cannot be

NUCLEAR SAFETY ASSURANCE PROGRAM

A-B E l e c t r i c a l Bus Hydraulic Interdependence

Nuclear safety requires t h a t t h e two e l e c t r i c a l power sources a t N Reactor be independent, However, a dependent condi t ion can e x i s t under c e r t a i n conditions because the turbine generator (TG) condenser i s cooled by r i v e r water from both A Bus p3wered and B Bus powered pumps, x d e r these ce r t a in conditions, cause inadequate cooling of t h e TG condenser and subsequent loss of B BGS. d i t i o n s under which the dependency would e x i s t , and appropriate Process Standards l imi t s were establ ished t o permit safe operation. However, t h e phenomena t h a t cause the sudden TG ccndenser flow reductions t h a t amplify the dependency were not prec ise ly defined by t h e test s e r i e s . Further t e s t i n g t o def ine these phenomena were scheduled for t h e long summer outage but w e r e can- c e l l e d becaGse of conf l i c t with o the r p r i o r i t y omage work such as t i e - i n of t he new emergency c o d i n g system, turbo-generator Class A overhaul, and in- spect ion of t he safety con t r c l l e g for debr i s . A l i p lans , procedures, and equipment a re ava i lab le f o r implementation of t he t e s t pmgram as outage time i s ava i lab le .

Thus, a l o s s of A Bus could,

The test s e r i e s of August 1969 defined con-

N Reactor Raw Water Systems Debris

Debris Development Test

Developmsnt Tes? 218,""Invesrigatisn of Large Pa r t i cu la t e Matter in IT Reac- t c r Circulat ing Raw Water System," was czmpieted a d a f i n a l r e p o x was

issued This t e s t w a s designed t o c o l l e c t and charac te r ize deb r i s which passes through t h e t r ave l ing screens i n t o t h e c i r c u l a t i n g r a w water system. The main conclusions were tha t t he concentrations of deb r i s en ter ing t h e system during t h e various t e s t periods were lower than t h e maximum permis- s i b l e concentrations contained i n t h e proposed r a w water c l ean l ines s c r i - teria, but t h a t numerous individual p a r t i c l e s were observed which were i n excess of t h e m a x i m u m volume and surface area spec i f ied i n t h e c r i t e r i a .

Specif ic ac t ions required t o correct t h e def ic ienc ies of t h e system are now being developed.

Graphite Cooling System

Fina l fiesign reviews a re i n progress concerning the e l imina t ion of debr i s accumulation i n t h e N Reactor graphi te cooling system backup supply by i n s t a l l i n g a f lu sh l i n e and screen.

N Reactor Ccnfinement Leak Rate Test

A ccnfinement leak ra te t e s t w a s completed a t N Reactor. The l e a k rate w a s about 2 . 5 percent as compared t o 1 . 4 percent t h e previous year and 5 percent as a l i m i t

A tes t on a se lee ted steam vent i n f l a t a b l e backup c losure showed it would not i n f l a t e . The b u t t e r f l y valve was then closed and w a s found t o l eak s ince it w a s improperly sea ted , Subsequently, a l l steam vent backup c losures and but- t e r f l y valves were inspected and a l l were reset and cycled a t leas t th ree times t o ensure proper c losing. I n addi t ion, t h e r e l ease pressure on t h e i n f l a t a b l e backup closures w a s r e s e t t o ensure proper release.

N Reactcr High Level Protection Studies

Studies are i n progress t o evaluate t h e t r a n s i e n t behavior of N Reactor for various r eac to r events t o more prec ise ly determine t h e margins of s a fe ty pro- vided by t h e flux and termperature monitoring systems. These s tud ie s are ex- pected t o result i n modification of t h e system requirements. The s tud ie s are being done by appl ica t ion of computer coded r eac to r t r a n s i e n t models.

Other N Reactor Control Studies

Current nuclear sa fe ty requirements spec i fy t h a t t h e e n t i r e s a f e t y rod system s h a l l be reca l ibra ted per iodical ly t o assure t h a t loading or o ther changes have not a f fec ted sa fe ty rod system s t r eng th i n an unpredicted manner. The s a f e t y rod ca l ib ra t ion t e s t s require a l a r g e amount of outage t i m e . A study i s i n progress t o evaluate ca lcu la t ion techniques which may be u t i l i z e d t o reduce the t e s t i n g requirements.

Checking cf t h e DUNCAP-3D code against an N Reactor model has been l a rge ly ccmpleted; ca lcu la t iona l r e s u l t s have shown good agreement witn t h e D-pattern ca l ib ra t ion . Runs are current ly being made t o eompare DUNCAP r e s u l t s with var ious rod configurations evaluated during t h e o r i g i n a l s t a r t u p t e s t s .

D-10

:;ASTE MANAGEtENT PROGRAM

Contaninated Coolant Treatment P i l o t P lan t

Work continues on the p i l o t p lan t f a c i l i t y which i s being i n s t a l l e d i n N Area t o develop da ta f o r the design of a Contaminated Coolant Treatment F a c i l i t y (CCTF) f o r N Reactor. i n s t a l l e d . Target date f o r s t a r tup of t he p i l o t p l an t i s October 1, 1970. The process v isua l ized cons is t s of f i l t e r i n g the I? Reactor bleed streams t o remove radio- a c t i v e so l id s with f i n a l cleanup by i o n exchange. Various types of i on ex- change r e s ins w i l l be invest igated. Both car t r idge and Vacco multisegmented f i l t e r s which have reduced w a s t e handling requirements will be t e s t ed . A sampling system ac t iva ted by a timing device w i l l allow measurements f o r t o t a l a c t i v i t y , individual rad io iso topic iden t i f i ca t ion and quant i ty , pH, and conductivity f o r the i n l e t and o u t l e t streams of' each f i l t e r and ion exchange column.

Presentat ion t o ACRS Subcommittee

The bui lding is complete and the process equipment i s The data acquis i t ion equipment i s about 85 percent complete.

The DUN Waste Management prac t ices were described t o t h e Hanford Subcommittee of t he Advisory Committee on Reactor Safeguards during t h e i r recent v i s i t i n August. The presentat ions included descr ip t ion of the source, i d e n t i t y , quan- t i t y , and re la t ionship t o control l i m i t s of t h e pr inc ipa l radionuclides generated by operation of t h e N and K reac tor and fue l f ab r i ca t ion f a c i l i t i e s .

SPECIAL IRRADIATIONS PROGRAM

Isotope Producticn

One hundred f i f t y bismuth pieces were charged in to f i v e KE Reactor process tubes t o produce Po-210 for Mound Research Corporation.

One iodine-containing capsule of the rev ised s t a in l e s s s t e e l design w a s charged i n t o a s ide t e s t channel t o produce xe-128 f o r Argonne PJational Laboratory.

Routine I r r ad ia t ions

Forty-three quickie ac t iva t ion ana lys i s capsules were i r r a d i a t e d i n t h e KE Quickie F a c i l i t i e s f o r BNW and WADCO. A downstream snout capsule containing high spec i f ic a c t i v i t y tantalum w a s discharged from the KE Snout F a c i l i t y f o r DUN, Three BNW t e n s i l e specimen capsules were discharged from t h e cooled magazine f a c i l i t y at KE.

TECHNICAL ACTIVITIES - FUEL FABRICATION

New Design Mark IV Outer B i l l e t s

A study has been completed t o determine t h e incentives and f e a s i b i l i t y of converting Tis fuel. production t o a loager b i l l e t . The study r e su l t ed i n a net;

D-11

design outer b i l l e t having a 21,g-inch length, yielding 1 2 fu l l - length f u e l s per extrusion. Conversion plans include:

& A t e n b i l l e t demonstration of t he new b i l l e t l ength during t h e second quar%er of F i sca l Year 1971.

e Ten percent or" production converted t o the new l eng th by t h e t h i r d quar te r of F i s c a l Year 1971.

The design includes a l s o a secondary Class I11 (20.3" long) b i l l e t which would y i e l d 11 ful l - length pieces .

The cur ren t length of t h e Mark I V outer preshaped b i l l e t i s 19.75 inches, and i t s ex t rus ion r e s u l t s i n t e n 26-inch length and one 17-inch length f u e l s .

New b i l l e t designs considered i n t h e study were based on assuming t h e same amount of defec t ive end material as is now being generated on cur ren t extru- s ions and adding t h e necessary b i l l e t mater ia l t o e f f e c t add i t iona l f u e l pieces , Three b i l l e t lengths were evaluated, a l l of which w e r e within t h e f eeds i t e cas t ing @apabili-i;y and t h e extrusion press capaci ty . Select ion of t h e 21.9-inch length b i l l e t was based both on the r e s u l t i n g savings and t h e production of fu l l - length fue l s i n each extrusion, Fur ther , t h e a v a i l a b i l i t y of a Class I11 b i l l e t y i e ld ing only full-length fue l s i s advantageous.

The cos t savings from converting N f u e l s production t o a longer b i l l e t der ive from increased uranium u t i l i z a t i o n and reduced number of components due t o t h e reduced number of extrusions. These costs have been ca lcu la ted based on scheduled productions f o r F i s c a l Years 1972-75 and are shown below:

Three Bi l le t s @! 21.9 Inches

F i s c a l Year LI 1972 - 1973 1974 - 1975

Savings Due t o Uranium U t i l i z a t i o n $41,432 $43 , 842 $45 Y 022 $46 , 868 Savings Due t o Components 23,239 24,449 24,874 26,314

$64,671 $68,291 $69,896 $73 , 182 Increasing b i l l e t length increases uranium design u t i l i z a t i o n and thus re- duces t h e amount o f sc rap returned t o NLO for reprocessing. A cos t f i gu re of $4.185 per pound of uranium fo r reclaiming extruded uranium was applied t o the d i f fe rence i n scrap generation r e su l t i ng from t h e longer b i l l e t s .

The cos t of b i l l e t components for longer b i l l e t s w a s based upon t h e reduct ion i n t h e number cf extrusions r e s u l t i n g from t h e increased b i l l e t s i z e as well as a compensating s l i g h t increase i n t h e cost of t he longer cladding compo- nents. s icn press t i m e per year made ava i lab le by the reduction i n t h e number of ex t rus ions ,

The cos t s do not r e f l e c t t h e approximate t e n addi-cional days of extru-

I)-12

DUB- 7060

The design of the b i l l e t assembly process allows fo r various b i l l e t lengths, and no problems are expected with t h e 21.9-inch b i l l e t . B i l l e t preheat cans have already been modified t o accomodate b i l l e t s up t o 24 inches i n length. The 2750 ton extrusion press can be loaded with t h i s length b i l l e t , and b i l - l e t s of t h i s s i ze have been successful ly extruded well within t h e press c a p a b i l i t i e s . An assembled 21.9-inch b i l l e t i s calculated t o be approximately 22.7 inches i n length. Approximately 3.9 inches of the press stem w i l l ex- tend i n t o t h e l i n e r pr ior t o b i l l e t upset. shear technique should be about 30 f e e t 6 inches long, which i s well within t h e 42 foo t runout table length.

The extruded tube using t h e bu t t

D-13

UNCLASSIFIED DWJ- 7 06 0

ADNINISTRATION - GENEFLAL

PROP3RTY ACCOUNTING

A phys ica l inventory of catalogued movable equipment assigned t o DUN i s sched- u led t o begin on October 5, 1970 i n t h e 100 Areas and on November 9, 1970 i n t h e 300 Area. value of $5,705,000. provided t o AEC-RL.

DUN has approximately 4,800 u n i t s with an o r ig ina l acqu i s i t i on Copies of t h e inventory procedures and schedule were

APPROVAL LETTERS Eh: A t zhe c lose of t h e repor t ing per iod, f i n a l ac t ion w a s pending on t h e fol lowing reques ts :

Number Sub j ec t

ATD-78 Personnel Chariges Add. #4 ATD-156 DUN Education and Training Add. #1 Program

ATD-196 F e a s i b i l i t y Systems Study

Tank, 1824

sup. #1

Demineralized Water Storage

Date of Transmit ta l t o AEC-SL

September 23, 1970

September 23, 1970

September 1 0 , 1970

September 16, 1970

DECLASSIFICATION REVIEW OF COLUMBIA RIVER DOCUMENTS

The AEC-RL Class i f ica t ion Off icer has been supplied a l i s t of a l l c l a s s i f i e d documents r e l a t i n g t o Columbia River s t u d i e s generated by DUN employees and Washington S ta t e University working under cont rac t with DUN. used t o determine the f e a s i b i l i t y of conducting a dec la s s i f i ca t ion review sponsored by AEC-HQ.

The l i s t w i l l be

EMPLOYMENT SUMMARY

DUN personnel t o t a l s and employee a l l o c a t i o n as of August 31 and September 30 a r e shown i n Appendix B.

AFFIRMATIVE ACTION PLAN

Three minority employees were added t o t h e work force during t h e month. This was o f f s e t by t h e r e tu rn of two Co-op Trainees t o Washington S ta t e Universi ty '

E-1 UTTCLAS S IF IED

UNCLASSIFIED DUN-7 0 60

and one resignation. I n t h e c l e r i c a l and r e l a t ed job category, good r e s u l t s were obtained; 100 percent of t h e new-hire additions i n t h e quarter ending September 30 were minor i t i e s . Conversely, no minority was represented i n t e n addi t ions t o t h e p ro fes s iona l staff during t h e same three-month period.

SAFETY

No personnel r ad ia t ion exposures exceeded operational control.

Month-end safe ty s t a t i s t i c s were:

Disabling i n j u r i e s - September - CY t o date

Days s ince last d i sab l ing in ju ry

Man-haus s ince last d i sab l ing in ju ry

0 3

84

565,519

.

.. , . . .

.

E- 2

I

AUTHORIZED PROJECTS

Number & Title

Single-Pass Reactors

DAP-526, Deactivation of Hanford Production Reactor C

N Reactor r P I GCP-506, Improved Safety

Platforms and Accesses - 100-N Area

GCP-411, Effluent Control Program - 100-N Area

DCE-519, Replacement of Bridge Crane and Hoist System with New Crane System - 105-N Storage Basin Area

DCP-528, Fire Protection System Improvements - 100-N

? rn rn H Y

APPENDIX A

PROJECT STATUS SUMMARY - REACTOR FACILITIES

Authorized Funds - $

105 , 000

300 , 000

1,830 , 000

465 , 000

40 , 000

Percent Complete Desim Construction

100 67

100 99

100 96

100 75

5 0

Remarks

Outside line subcontract work for the fire protection system has been completed.

Miscellaneous cleanup work in progress.

Section V - In service. ATP results being reviewed for compliance with Design Criteria . The two east cranes have been placed on the runway beams and the communication and power rails are being installed. west cranes are being de-bugged at the vendor's plant.

Design activity suspended by AEC-RL. $40,000 authorized of $;lgO,OOO requested.

The

. .

Author i zed Percent Complete --- Number & Title Funds - $ Design Construction Remai ks

DCE- 539 , Upgrade Flux 335,000 lo Monikor System - N Reactor

DAE-540, Smoke Density 55,000 20 Monitors f o r DUN-Operated Faciljties - 100 Areas

PROJECT PROPOSALS AWAITING AUTHORIZATION -

Number & Title

GCP-411, Revc 2, Effluent Control Program - 100-N

DCP-525, Fire Protection Improvements - KE r DCP-527, Graphite Cooling & Fog Spray Improve- Iu ments - ~ O O - N ~ I

DCP-528, Rev. 1 , Fire Protection System Improvements - 100-N DCP-529, Gravity Drainage System and Disposal Basin for 100-N Area'

DAP-530, Upgraded Electrical Services & Lighting 1100-N & 1101-N Buildings

DCP-538, Heat and Ventilation System Improvements 105-N & 109-N Buildings.'

DCP-542, High Pressure Injection System Improvements - N Plant

c:

F U J ClI H r

0 Detail design continues on schedule- in I;

l3 c(

0 Design has ccjntinued by HES-Vitro, The Title 1 drawing depicting equipment layout h a s been issued for comment.

Funds Requested Date to AEC-RL

$ 2,010,000 9 1231 69

225 , 000 5/2/69

394 , 000 3/31/70

200 , 000

78,000

330 , 000

138 , 000

U (5\ 0 I 'To be returned to DUN by AEC-RL because of long holding periods.

%i I w

... . r I

(, 1 1

. .

Number & Title Funds Requested

DCE-544, Critical Instrument Power Supplies - $ 65,000 N Reactor

PROJECT PROPOSAL PREPARATION

Number & Title Design Criteria

Export Water System Backup - 182-D (for 200 Areas) C omple t e d

C omple t e d Stack Monitoring Improvements - 100-N Plant

Permanent Markers for Buried Wastes in Terminated Prepared 100-D, 100-F and 100-H Sites

Date to AEC-RL

6/15/70

Project Proposal

Completed. Held by sponsor (ARHCO). I

Prepared. Weld for later submission.

In preparation.

P 4

UNCL4SSIFIED DUR-7 060

APPERDIX B

EMPLOYMENT SUMMARY (with employee a l l o c a t i o n s )

CONTFiACT PERSONNEL

02 Programs

Douglas United Nuclear Ass i s t ing Other Contractors

To ta l - 02

Other Programs Under AEC Contract

Ass i s t ing Other Contractors and WPPSS Specia l I r r a d i a t i o n s Other Programs - Standards

To ta l - Other Programs

To ta l Contract Personnel

COMMERCIAL ACTIVITIES PERSONNEZ

TOTAL FORCE

8/31/70

1273 17 -

1290

37 9 4 -

50

1340

22 - 1362

9/30/70

1267 17 -

1284

37 6 2 -

45 - 1329

21

1350

-

UNCLASSIFIED F-4