Cover Sheet for a Hanford Historical Document Released for Public Availability

Released 1995

Prepared for the U.S. Department of Energy under Contract DE-AC06-76RLO 1830

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

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I

MONTHLY REPORT

JUNE 1971

' DUN-7591

INFORMATION CONCERNING USE OF THIS REPORT

PATENT S T A T U S This document copy, since it is transmiHed in advance of patent clearances, is made available

in confidence solely for use in performance of work under contracts with the U. 5. Atomic Energy Commission. This document is not to be published nor i ts contents otherwise disseminated or used for purposes other %an specified above before patent approval for such release or use has been secured,

. upon reques+, from the Chief, Chicago Patent Group, U. S. Atomic Energy Commission. 9800 So. C a s ,Avenue, Arglonne, Illinois, 60439.

P R E L I M I N A R Y R E P O R T This report contains information of a preliminary nature prepared in the course of work under

Akmic Energy Commission Contract AT(45-1)-1857. This information is subject to correction or mod- ification upon the collection and evaluation of additional data.

NOTICE This report was prepared as an account of work sponsored by the United States Government.

Neither the United States nor the United States Atomic Energy Commission, nor any of their em- ployees, makes any warranty, express or implied, or assumes any legal liability or respoqsibility for the accuracy, completeness or usefulness of any information, apparatus, product, or process disclosed, or represenls %a+ i ts use would no+ infringe privatelyswned rights.

PLEASE SIGN BEFORE READING THIS PUBLICATION

ROUTE DATE

AEGRL niauno. wran

\

JUNE 1971

DOUGLAS UNITED NUCLEAR, INC.

Richland, Washington

Work performed under Supplemental Agreement No. 1 to Contract AT(h5-1)-1857 between the Atomic Energy Commission and

Douglas United Nuclear, Inc.

-1-

UNCLASSIFIED

REPORT DISTRIBUTION

AEC - RICHLAND OPERATIONS OFFICE DOUGLAS UNITED NUCLEAR

1-6. 0. J. Elgert

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

Division of Production

AEC - SAVANNAH RIVER 10. N. Stetson

DU PONT - SAVANNAH RIVER

11-12. tJ. B. Scott

BATTELLE-NORTHWEST

13. R. L. Dillon 14. J. J. Fuquay 1 5 B. M. Johnson, Jr. 16. R. S. Paul 1 7 L. C. Schmid

ATLANTIC RICHFIELD HANFORD

18. L. M. Richards 19 H. P. Shaw 20. R. E. Tomlinson

DOUGLAS UNITED NUCLEAR

21. C. L. Abel 22. D. H. Bangerter 23 J. R. Bolliger 24 * J. J. Bombino

DUN-” 5 91

25 9

26. 27 * 28 29 30. 31 32 33 34. 35 36.

38. 39 40. 41 ., 42. 43 44 * 45 9

46. 47 48. 49 50. 51 * 52 53 9

54

37 0

P. A. Carlson W. G. Catts R. E. Dunn A. E. Engler J. M. Fox, Jr. C. D. Harrington D. L. Hovorka R. T. Jessen H. R. Kosmata H. P. Kraemer, Jr. C. W. Kuhlman A. R. Maguire J. A. Cowan 3. T. Martell W. M. Mathis D. W. Peacock R. S. Peterson C. F. Poor C. A. Priode (Work T. Prudich T. B. Pugh K. L. Robertson R. X. Robinson J. E. Ruffin B. Schauss 0. C. Schroeder W. Seeburger R. H. Shoemaker J. T. Stringer DUN File DUN Record

COPY)

-2- UNCLASSIFIED

UNCLASSIFIED

TABLE OF CONTENTS

SUMMARY

DUN-7 591

Starting Page

A-1

N REACTOR PLANT OPERATIONS

FUEL FABRICATION

TECHNICAL ACTIVITIES

ADMINISTRATION - GENERAL APPENDIX

A. Project Status Summary

B. Ehployment Summary

-3-

B-L

e-1

D-1

E-1

F-1

F-5

UNCLASSIFIED

UNCLASSIFIED DUN-7591

SUMMARY

N REACTOR P M T OPERATION

N Reactor remained shut down, and the maintenance program directed toward startup in July was continued.

FUEL FABRICATION

Production Statistics - Tons Billets extruded

Finished fuel produced

39- 0

7.5

Rework and first-run short end pieces only were processed. Output was as planned.

TECHNICAL ACTIVITIES

Coolant was discontinued in the process channels on the left side of N Reactor for about 14 hours to permit repairs on the CV-3-1 check valve. of performing maintenance in this manner on primary loop components that cannot otherwise be isolated was established by production test in May.

Measurements of N process tubes indicate a tube elongation apparently proportional to cumulative exposure. Future work will be directed toward determining the effect of elongation on operation.

The feasibility

The transfers of DUN operated facilities and related project and fund request activities to ARHCO, BNW and WADCO, were essentially completed on June 30. Company force was reduced 71 employees as a result of the transfers.

There were no disabling injuries and no radiation exposures exceeded operational control.

The

Q r i

Charles D. Harrington ,' President

A-1 UNCLASSIFIED

UNCLASSIFIED DUN- 7 5 91

N REACTOR PLANT OPERATIONS

PRODUCTION

General

Reactor production ( l a rge ly f u e l grade Pu), power level and r e l a t e d s t a t i s t i c s are tabulated below. Input production and time operated e f f ic iency (TOE) fo r

s i x months are shown on t h e following chart :

100

75

50

25

0 Jan Feb Mar June

S t a t i s t i c a l Summary

E l e c t r i c a l Generation (KMWH) - WPPSS - 1 8 4 - N

Fuel Charge (Tons) - 94 Metal - 125 Metal - Natural Uranium

Fuel O i l Usage (bbl. )

00

80 .P c 60 Q,

k" a,

w 0 E-1

40 '?

20

0 .

0 0.3

331.09 64.68

9,670

OPEWITING EXPERIENCE

Reactor Loading

The 27 tubes discharged last month and l e f t empty f o r tube examination were re- charged t o complete t h e scheduled reac tor loading changes. Five tubes of PT 94 Mk-IV-AA mater ia l scheduled f o r discharge were l e f t i n t h e r eac to r f o r t h e next operating period at t h e request of t h e AEC who wish t o consider them f o r a long term i r r a d i a t i o n study. The reac tor loading at month end i s shown on t h e f ront face map on page B-4 .

B-1 UNCLASSIFIED

UNCLASSIFIED, DUN- 7 5 91

Reactor Outages

The reactor 'was shut down the e n t i r e month i n continuation of the January 28 outage. A plant maintenance program was i n progress throughout t he month with s t a r tup scheduled f o r July. a close and spec ia l systems tests were i n progress t o demonstrate readiness of t h e plant f o r s ta r tup . Tota l outage hours f o r the month were 720.

A t month end t h e r epa i r a c t i v i t i e s were drawing t o

Equipment Experience

Primary Loop

J. A. Jones .forces completed V-ll and V-12 valve r epa i r s on June 25. 88 reconditioned V-12 valves and 18 spool pieces were i n s t a l l e d during t h e out- age.. Four addi t iona l leaks not requir ing valve removal were repaired. The t o t a l l eaks repaired involving V-12 valves i s 110. A t o t a l of 229 V-11 valves was repacked with Grafoi l during the outage. remaining leaks i n the V-11 and V-12 areas indicates an accumulated leakage of l e s s than one gal lon per minute at these locat ions.

A t o t a l of

Technical analysis of t he data on

Other primary loop leak r epa i r s were continued during the month by N Plant Mainten- ance forces . A t month end lo7 leaks had been repaired and 17 known leaks remained t o be repaired. The leak maintenance w i l l continue toward t h e goal t o have a l l miscellaneous leaks repaired by p lan t s t a r t u p i n July. A notable achievement during the month w a s t h e r epa i r of t h e CV-3-1 valve bonnet leak. This leak has ex is ted fo r many years and required removing primary coolant flow from t h e l e f t s ide of t he re- ac tor t o e f f e c t repa i rs . Testing showed t h a t l e f t s ide reac tor heat could be satis- f a c t o r i l y removed through t h e graphi te cooling system, thus making it possible t o s top l e f t s ide primary cooling long enough t o r epa i r t he CV-3-1 valve.

Demineralized Water Plant

The degas i f i e r f o r t he demineralized water p lan t w a s removed from service f o r in- t e r n a l inspect ion and repa i r of t h e rubber l in ing . good qua l i ty water at a reduced r a t e during t h e th ree weeks the degas i f ie r w a s un- ava i lab le by removing oxygen using a hydrazine-activated carbon treatment method t h a t w a s developed espec ia l ly f o r t h i s job. The damaged l i n e r area i n the degasi- f ier w a s repaired by an outs ide vendor, but a l l work i n connection with removing and re turn ing the vesse l t o service w a s performed by N Plant Maintenance forces .

The plant continued t o make

B a l l Safety System

The f i v e b a l l channels reported previously t o have l i n e r block separations were resleeved during the month. This completes the resleeving program a t t h i s t i m e , and a l l b a l l channels were i n operating condition a t month end.

Graphite Heat Exchanger

The No. 3 heat exchanger w a s received during the week of June 21, following r epa i r and retubing a t 189-~. It has been s e t i n place i n the auxi l ia ry c e l l i n t he 109 Building and i n s t a l l a t i o n w a s i n progress at month end. It i s expected t h a t r a w water piping connections can be completed p r io r t o s t a r tup and t h a t the remainder of t he i n s t a l l a t i o n can-be completed as time i s ava i lab le during subsequent outages.

B-2 UNCLASSIFIED

UNCLASSIFIED DUN-7591

I -

Boilers

An "all boiler" outage was taken on June 28 for steam and feedwater repairs to equipment common to all boilers. At month end, the boilers are all operating and preparations for boiler load tests early in July are in progress. A leak in a roof tube of CE-1 boiler was repaired.

Process Tube Temperature Monitor

At month end all but four unreadable temperature detectors have been restored to service.

Equipment Modifications

During the outage, 182 repairs have been made to date.

The following equipment modifications were completed:

DC-3092, "Flow Monitor Transducer Calibration Facility Modification. '' Modifications were made to speed up calibration work by changing the required demineralized water source from an accumulator system to the existing 200 psig demineralized water supply.

DC-3119, "Primary Pump Seal Water Versus Pump Suction Transmitter Replace- ment System No. 415.33." More reliable transmitters were provided for ,AP instrumentation in the six 109-N cells.

DC-3157, "IWV-253 Valve Positioners Pressure Change - 109-N." This change provides improved pressurizer level control following a reactor scram.

DC-3159, "Electrical Deactivation of ECS Valves V-19, V-23-1, V-23-2, V-28-1, and v-28-2." These valves are no longer required to be electrically controlled and were electrically deactivated to remove several potential ground sources, thus reducing the possibility of unscheduled reactor outages.

MDC-N-69-67, "Mechanical Control Board and Battery Charger Load Circuit Revision." Modifications corrected wiring deficiencies.

MDC-N-70-74, "Regulated Air Supply for Panels No. 1 and No. 2 - CE Boilers." Precision air regulators were provided in the instrument air system to give more stable boiler control.

MDC-N-70-88, "Mixing Station for Boiler Fuel Oil Additives. " piping were installed to permit supplying additives to fuel oil in efforts to reduce tube fireside scale and acid corrosion in the standby boilers.

Equipment and

MDC-N-71-2, "RTD Communications." Improved communications systems were provided between the various locations where process tube temperature monitor work is performed in order to reduce radiation exposure.

MDC-N-71-15, "Fuel Oil Pressure Annunciator CE-1 and CE-2. 'I circuitry were modified to make these boiler system annunciators operable.

Wiring and

B- 3 UNCLASSIFIED

UNcLAsslCFm DUN-7591

Loading Pa t t e rn - N Reactor

Fuel No. PT-NR NO. No . Tubes -- Code Tubes Description - --

E, ES 166 Mk-IC (94 Metal - Fringe) 94 5 F 2 Mk-IV (94 Metal - High U-236) 11 3 4 G 571 Mk-IV (94 M e t a l - Central & Fringe) N 1 Mk-1B (Matural U) X - 230 Mk-LA & Mk-IV (125 & 94 Metal)

970 T o t a l 34 Tota l PT

1004 Grand Tota l -

18-4

Description

Mk-IVAA Demonstration Evaluation of t h e End Spider Inner Support System

136 8 Mk-IV Non-bonded End Closures

118 16 Mk-IV Performance T e s t - 1 Graphite Sample Channel

34 Total PTs

UNCLASSIFIED

UNCLASSIFIED

FUEL FABRICATION

PRODUCTION

Input Production - Tons Percent of Forecast

Output Production

Finished Assemblies

Tons Output Percent of Forecast

Uranium Utilization - %

BO9 X09

Month-End Inventories - Tons

DUN-7 591

39.0

130.1

7.5 44.4

78.9 -

Bare Uranium Billets Finished Fuel

91 24 5

OPERATING EXPERIENCE

The low level throughput consisted of rework fuels and first-run short end pieces. Defective rates were higher than normal, which is usual when processing short end 'pieces.

PROCESS ASSISTANCE AND CONTROL

Billet assembly problems were recently encountered when the outside diameter on inner copper sleeves for Mark B billets was found to be on the high side of speci- fications and the center hole diameter on end plates was on the low side, resulting in an excessive'interference fit between the sleeves and the end plates. Etching the sleeves for an increased amount of time to reduce the outside diameter tempo- rarily relieves the problem. side diameters on the high side of specifications.

An alternate method is to reduce the sleeve OD by magnetic forces. This arrangement uses a compression coil which s e t s up magnetic forces opposite to those generated in the workpiece (copper sleeve), thus compressing the end of the sleeve. By adjusting the energy delivered to the compression coil the amount of deformation of the work- piece can be controlled. Approximately 120 sleeves have been successfully sized for current production usage. sleeves.

Fifty percent of 3,000 sleeves in inventory have out-

No assembly problems were encountered with the magneformed

TRANSFER OF DUN FACILITIES TO OTmR AEC-RL CONTRACTORS

The transfer of DUN facilities scheduled to go to other contractors was essentially complete by July 1. Included were inventories of essential materials, equipment,

c-1 UNCLASSIFIED

UNCLASSIFIED DUN-" 591

blueprints, operating procedures and other similar information. The following summarizes the major facilities and services transferred:

300 Utilities - t o WADCO

The 300 Area Utility Services consisting of steam generation and distribution, water system river pumps, filter plant and distribution lines, compressed air and distribu- tion, sanitary and process sewer collection and disposition, emergency electrical generation, and all of the associated buildings and equipment.

Central Service Facility - To BNW The analytical laboratory and central maintenance offices were transferred. machine shop and instrument and electrical shops were vacated.

The

Machine Shop .- To ARHCO

Eight major machine tools were transferred to ARHCO and moved to 200-West shops. Machine shop service to DUN will be mainly supplied by ARHCO.

300 Area Trash Collection and Disposal - To ARHCO The responsibility for 300 Area trash collection and disposal was transferred to ARHCO, including all collection equipment, 301 garage and equipment storage. ,

100 Area Export Water System - To ARHCO The 100 Area export water system operation supplying water to 200, 100-D and 100-F Areas was transferred, including all the buildings and facilities.

Attendant with the transfer of the utility services was the transfer of related project and fund request activities as follows :

DAP 536 - No. 2 Boiler Replacement for Increased Steam Generating Capacity

Demolition of the existing boiler is essentially complete. Bumstead-Woolford, Seattle, has been awarded the contract for the new 100,000 #/hr. oil-fired boiler.

Vendor design effort is expected to be completed July 1. Boiler delivery, subject to approval, is anticipated in August and installation may be completed in November and December, 1971.

DAP 541 - Boiler Feedwater System Additions (500,000 #/hr) Deaerator This project, authorized $174,000 for 5OO,OOO #/hr deaerator and related work. Invitations to bid have been issued and responses are due month end. J. A. Jones has been assigned installation responsibility.

DAP 543 - Smoke Density Monitors - 384 Building This project authorized $50,000 for smoke density monitors on three powerhouse stacks. Installation work is in progress. Some equipment on order is not on site. Completion date is scheduled for November 1971.

c-2 UNCLASSIFIED

UNCLASSIFIED DUN-7591

DAP 550 - Steam Distr ibut ion System Additions - 300 Area

A project proposal requesting $130,000 f o r about 1300/ft of 10-inch steam l i n e and about the same length of four-inch condensate re turn l i n e w a s forwarded t o AEC on May 18.

The work proposed would provide looped service t o the Biology f a c i l i t y and other f a c i l i t i e s i n the southeast sect ion of 300 Area.

Design Criteria are i n preliminary form. cannot be funded t h i s f i s c a l year.

AEC-RL has determined t h a t t h i s work

DCP 548 - Sanitary Sewer System Additions - Southeast Section 300 Area

A $49,000 Request f o r Directive has been with AEC-RL s ince December 2, 1970. Consideration of t h e proposal has apparently been deferred pending t h e CH$ report and recommendations f o r sewage treatment. The AEC has determined t h a t t h i s project cannot be funded t h i s f i s c a l year.

Fuel O i l Storage Bunker Addition -

Development of a proposal t o increase ex is t ing capacity by 150,000 gallons, at an estimated cos t of $90,000 i s i n progress. The work would maintain t h e CUT- r e n t seven-day supply when t h e new #2 b o i l e r i s placed i n service. cannot be funded t h i s f i s c a l year.

This. work

Fencing Contaminated B u r i a l S i t e and Burning P i t

Funds have been requested f o r the B u r i a l S i t e only; however, t h e funding situa- t i o n has caused approval t o be deferred.

The Fuels Section force has been reduced by 71 as a result of t h e above trans- fers. Those i n excess of t h e receiving contractors ' needs and of DUN'S require- ments have been o r w i l l be laid-off.

100-KE & K PLANT DEACTIVATION STATUS

A s of June 30, a l l work required t o deact ivate K Plant has been completed, except f o r those f a c i l i t i e s required t o be l e f t i n service f o r t h e s torage and shipment of f u e l elements from t h e s torage basins and f o r t h e Corrosion T e s t F a c i l i t i e s i n 105-KE and 1706-KE which a r e operated f o r BNW.

Project work required t o protect equipment i n K Plant from moisture condensate and f o r f i r e protect ion after t h e remaining f a c i l i t i e s have been deactivated i s approximately 45 percent completed. The s ixef f luent re ten t ion basins have been drained, w a l l s washed down, and approximately two feet of ear th placed i n t h e bottom of t h e bas ins , and a l l i n l e t and e x i t piping has been sealed with blanks. The KE c r i b and trench have a l s o been covered with ear th . A l l material s tored at t h e Burial Ground has been disposed o f , and t h e b u r i a l trenches have been covered.

c-3 UNCLASSIFIED

UNCLASSIFIED DUN-7 5 91.

TECHNICAL ACTIVITIES

N REACTOR OPERABILITY PROGRAM

PT-NR-271 - Primary Loop Coolant Shut-Off Testing and CV-3-1 Repair

PT-NR-271 w a s performed during t h e weekend of May 22-23, for t h e purpose of generating tes t data necessary t o form t h e bas i s t o allow extended (24 hours or grea te r ) r eac to r primary loop coolant shut-off t i m e . This t es t then estab- l i shed t h e f e a s i b i l i t y of performing required maintenance on various primary loop components t h a t cannot otherwise be i so la ted .

The information obtained from PT-NR-271 was immediately u t i l i zed t o r epa i r t h e seal p l a t e of CV-3-1. repaired on June 2.

A supplement t o PT-NR-271 w a s w r i t t e n and t h e check valve

Coolant w a s stagnated i n t h e process channels on t h e l e f t s ide of t h e reac tor core f o r about 21 hours during the i n i t i a l t es t , and f o r about 1 4 hours during t h e check valve r epa i r . Some of t he conclusions obtained from t h e tes t a r e summarized i n t h e following paragraphs.

1. Observed Temperature Transients

Special thermocouples were inser ted i n t h e center f u e l channel of t h ree process tubes t o monitor t h e ac tua l temperature t r a n s i e n t s i n a typ ica l high power f u e l column. I n addi t ion, two thermocouple t r a i n assemblies previously i n s t a l l e d i n t h e core were monitored, mum observed temperatures are summarized below:

Predicted versus maxi-

Test CV-3-1 Repair Observed Approximate Approximate

Predicted Maximum Maximum Steady-State Steady-State Tube No.

1748* 2249* 2750" 1443** 1843**

Steady-State Temp. F Temp. F Temp. F Temp. F

142 136 129 120 134 141 118 123 136 151 147 114 - ,-b 77 74 70 - 8 nJ 77 77 67

* A t center of fuel column ** A t o u t l e t of f u e l column

A s teady-state temperature w a s not a t ta ined during t h e t es t , and t h e temp- e ra ture t r a n s i e n t s reached a maximum and declined s teadi ly . This phenomenon i s a t t r i b u t e d t o t h e onset of na tu ra l convection within each channel and connector, as was evidenced by ATTM system maps taken during t h e test .

A s noted i n the t a b l e of results, t h e approximate s teady-state temperatures observed during the CV-3-1 r epa i r were lower than those observed during the t e s t . forced c i r cu la t ion through t h e reac tor core which r e su l t ed from t he i so la - t i o n procedure used.

The difference is probably due t o the onset of a s m a l l amount of

D-1 UNCLASSIFIED

UNCLASSIFIED DUN-7' 591

2. Hydrogen Detection

Pr ior t o performing t h e t es t , some concern w a s expressed with respect t o t h e rad io lys i s of water t o H;! and 02 and the po ten t i a l accumulation of €I2 i n areas around the primary piping vents , high spots i n the piping, i n areas around CV-3-1, and i n Zone 1. No hydrogen w a s detected during the t e s t or CV-3-1 repa i r .

FUEL PERFORMANCE PROGRAM

Disposit ion of High Exposure Mark I V and Mark IVAA Fuel

The d ispos i t ion of t e n high exposure Mark IV and Mark IVAA f u e l columns charged under PT-94 and PT-107 i s summarized i n the following tab le . t h e AEC, f i v e Mark I V columns were discharged before reaching t h e i r authorized exposure. i r r a d i a t i o n period.

A s d i rec ted by

Five additi 'onal Mark IVAA columns were retained f o r an addi t iona l

TABLE I

Disposit ion of High Exposure Mark I V and Mark I V M Fuel

PT-NR-94 - Mark I V Demonstration

Present Authorized Exposure Tube No. Fuel Type Exposure, MWD/T MWD/T (by PT) Disposit ion .

2359 Mark I V 'L 5445 1443 Mark I V AA 2, 2550 1843 Mark I V AA 2560 2644 ' Mark I V AA 1965 2670 Mark I V AA 2035

6000 Discharged 6/21 6000 ) 2700 ) 2500 ) 2500 )

Retained f o r one i r r a d i a t i o n period

2948 Mark I V AA 2120 2500

PT-NR-107 - Central Zone Loading of Mark IV Fuel

2i54 Mark I V 3530 5000 2160 Mark I V 3520

2460 Mark I V 3580 2452 Mark I V 3760

6000 ) 4000 ) 5000

Discharged 6/21

Column 3262 charged under PT-NR-113, "Evaluation of End Spider Inner Support System," w a s discharged June 21, and i s being held f o r examination t o evaluate the end spider support system.

PLANT LIFE PROGRAM

Steam Generator Tubes

Scanning e lec t ron microscopic (SEN) examination of t he secondary s ide of samples of steam generator 6B ~nconel-600 tubing has shown no indicat ion of corrosive a t tack under t h e tube supports when t h e samples were examined i n t h e filmed con- d i t ion . The examination w i l l be repeated a f t e r t he f i l m has been removed.

D-2 UNCLASSIFIED

B a l l Channel Renovation

On June 7 the removal of the top three b a l l channel l i n e r blocks i n channels 43,52,53,56, and 63 w a s i n i t i a t e d , s t a r t i n g with channel 52. The plan was t o resleeve these channels t o t he top of t he fourth block with new, more f l ex ib l e graphi te sleeves t o correct ex is t ing open block junctions and eliminate inad- ver ten t l o s s of balls in to the graphi te stack. A l l channel renovation and re- sleeving work w a s completed on June 17.

The major portion of the work involved development and t e s t i n g of too l ing f o r removal of t he v e r t i c a l l y "cracked" t h i r d block i n channel 52. t h i s cracked block w a s extremely d i f f i c u l t due t o channel separations i n the reac tor front-to-rear d i rec t ion , which allowed t h e halves of the block t o " f loa t" and wedge i n the channel. June 7 and resleeving was completed on June 14. were a l l completed on June 17 w i t h no problem.

Removal of

Removal e f f o r t s on channel 52 s t a r t e d on The remaining four channels

Design and fabr ica t ion of spec ia l too l ing t o reseat blocks 4, 5, 6 , 7, and 8 i n channels 53 and 56 has been inTtiated.

Control Rod Corrosion

Comparative spectrographic analysis and a search of o r ig ina l purchase orders has ve r i f i ed that the N cont ro l rod aluminum s e a l sect ion tubing is ac tua l ly 6063 aluminum a l loy , r a the r than the 6061 a l loy t h a t t he tubing drawing, H-1-28018, specif ied. Verif icat ion t h a t t he s e a l sec t ion tubing and the t i p sect ion boron carbide containing aluminum extrusions are the sane a l loy rein- forces the concern t h a t the extensive p i t t i n g observed on the s e a l sect ion tubing may a l s o be present on the t i p sec t ion extrusions.

Design c r i t e r i a and a pro jec t proposal were submitted t o AEC-RL ea r ly i n June w i t h the a i m of obtaining FY 1971 funds f o r conversion of t he present s ing le pass coolant system t o a rec i rcu la t ing system u t i l i z i n g high pu r i ty water. The estimated t o t a l p ro jec t cost i s $235,000. t h e pro jec t are 1) a reduction of cont ro l rod replacement cos ts and improved operating cont inui ty by avoiding reactor outages due t o rod leaks , and 2) a reduction i n chemical and radioisotopic pol lu t ion of t h e environs affected by the single-pass coolant e f f luent .

Benefits t o be obtained from

Reactor Layaway Technology

Corrosion coupons placed i n loops KER-2 and 3 i n 1964 t o provide information of the long-term corrosion cha rac t e r i s t i c s of severa l materials i n a stagnant deionized water-hydrazine-lithium hydroxide environment a t room temperature have been examined.

Carbon steel coupons exposed t o the layaway solut ion i n KER-2 showed weight loss equivalent t o a penetrat ion of 0.01 mils/year. had been subjected t o aerated f i l t e r e d water f o r approximately t e n days i n 1970. A s a consequence, the weight l o s s of the carbon s t e e l coupons i n KER-3 w a s about f i v e times the weight loss of the coupons from KER-2 with f i v e t o t e n t i m e s the surface film.

Carbon steel coupons i n KER-3

D- 3

UNCLASSIFIED

The t e s t s c l ea r ly demonstrated the effect iveness of t h e pH 10-11 l i th ium hy- droxide, 500 ppm hydrazine so lu t ion f o r long-term layup of carbon s t e e l systems.

Process Tube Surveil lance

Process tube 3059 w a s removed from N Reactor i n May as pa r t of a continuing survei l lance program t o examine r ad ia t ion After a new tube had been cu t t o t h e prescribed length and inser ted , it w a s observed t h a t t h e nozzle would not fit with t h e connector Grayloc coupling. A comparative measurement w a s then made of t h e length of tube 3054 which had been in-reactor s ince o r ig ina l s ta r tup . It w a s found t o be approximately one inch longer than t h e or ig ina l ly i n s t a l l e d length of 53 feet. This measurement suggested t h a t the tubes i n t h e reac tor s ince s t a r t u p had elongated r a the r than t h e newly i n s t a l l e d tube mistakenly cut an inch too short . A measurement program w a s i n i t i a t e d t o determine t h e extent of tube elongation.

n;e lengths of 122 tubes were i n d i r e c t l y measured by measuring the dis tance from the face of the f ron t gas thimble t o t h e rear of t he f ront nozzle thread boss. Since the gas thimble i s bol ted t o the f ron t sh ie ld while t he nozzle i s f ixed t o t h e tube, t h i s measurement when compared t o t h e o r ig ina l dimensions w a s an indi- ca t ion of tube elongation. elongated 0.8-0.9 inch. t h e t o t a l length increase s ince any outward movement of e i t h e r f ront or r e a r sh i e ld would reduce t h e apparent tube length increase. Subsequent measurements of t he overa l l front-to-rear nozzle dis tances made through four tubes suggests t h a t some outward sh ie ld movement has occurred and t h a t ove ra l l tube elongation is ac tua l ly 0.1-0.2 inches g rea t e r , i .e., approximately one inch. Variation i n tube elongation w a s noted and was found t o be r e l a t ed t o the method by which the tubes were fabricated. The 30 percent cold-work tubes made by Harvey Aluminum Company (approximately 92 percent of reac tor tubes) showed up t o one inch elonga- t i o n while t h e 18 percent cold-work Harvey tubes ( 2 percent of reac tor tubes) showed 0.5-0.7 inch elongation and t h e 18 percent cold-work tubes made by Chase Brass ( 6 percent of reac tor tubes) had elongated only 0.3-0.5 inch. t he elongaticn was approximately proport ional t o the cumulative tube exposure, i . e . , f r inge tubes of a tube group elongated l e s s than c e n t r a l zone tubes i n t h e same grouping.

The s igni f icance of t h e length change from an operat ional viewpoint i s under study. A po ten t i a l problem i s associated with t h e four keys i n the ‘front nozzle which r i d e i n keyways ins ide t h e gas thimble t o prevent t h e nozzle and tube from ro ta t ing . An axial clearance of 1.75 inches w a s o r ig ina l ly provided and w a s ample t o accommodate t h e 0.8-0.9 inch of thermal expansion normally r e su l t i ng from oper- a t ion. a x i a l clearance between the keys and t h e gas s e a l packing between the nozzle and thimble has apparently been used up so t h a t at operating temperature t h e four keys are probably but t ing against t h e gas packing r i n g i n some cases. Any addi t iona l elongation w i l l t r a n s f e r load t o the 0.034 inch w a l l thimble and t o t h e f ront sh ie ld . A key question, therefore , i s t h e mechanism of the observed elongation. If elongation i s due t o irradiation-induced creep, the pressure load w i l l gradually be t r ans fe r r ed from the tube t o the thimble and sh ie ld and t h e creep r a t e would be expected t o drop as the a x i a l s t r e s s i n the tube drops. However, i f t he elongation is due t o a swelling phenomenon (independent of s t r e s s ) , s ign i f i can t ly higher

e f f e c t s on N tubes.

It w a s found t h a t many high flux tubes had apparently It w a s not c e r t a i n a t t h a t time whether t h i s represented

In a l l cases ,

Because of t h e addi t iona l 0.8-1 inch elongation from tube elongation,

D-4 UNCLASSIFIED

stresses could be involved, similar to those associated with attempting to restrain thermal expansion. To date the exact cause of the observed elongation is not known; however, the length increase observed can be predicted utilizing an expression for the creep strain of internally-pressurized Zircaloy tubes developed from in-reactor test data by Watkins and Wood of

Future work involves taking additional reactor measurements to 1) determine the extent of front and rear shield movement, 2 ) evaluate tube elongation mechanisms by measuring thimble strains, and 3) determine whether differential process tube lengths has reduced the spacings between some inlet connectors. The first two of these require the development of equiment and techniques so only the third is scheduled for completion this outage..

N Graphite Surveillance

Two process tube vertical height traverses were completed to determine front-to- rear profiles at the bottom of the stack. Preliminary analyses of the data show channel 0358 has subsided about 0.50 inch and channel 0357 has risen about 0.25 inch. The 0357 measurements confirm the 0.25-0.50 inch rise'measured in a side- to-side traverse channel at the bottom of the stack in May. Lifting of the bottom of the stack while the top subsides w a s anticipated as a result of the keyed structure of the stack.

DECONTAMINATION PROGRAM

A report, DUN-7674, "Radiation Levels - V-12 Pipe Spaces," has been prepared and issued. The results of this analysis have shown the incentive for developing an external decontamination technique which could be applied to the reactor pipe spaces. Radiation levels in excess of 1000 mrem/hr were experienced by pipefitters in V-12 leak repairs with a total exposure of 153 rem f o r 110 leak repairs in V-12 pipe spaces. internally deposited radioactivity.

Of these radiation levels, approximately 150-250 mR/hr were from

Planning has been initiated to provide a method and equipment to decontaminate the external surfaces of the pipe spaces and a program plan has been drawn up. This includes the following:

Evaluation of the effectiveness of several decontaminants to be done off- reactor on valves removed from the reactor.

Analysis of system components which may be exposed to decontamination solutions, to determine the degree to which they will be involved in a mass decontamination effort.

A program to determine short- and long-term corrosion effects of the decontamination chemicals on the materials present in the PCL system.

~

'TRGR-1757, "The Significance of Irradiation-Induced Creep on Reactor Performance of a Zircaloy-2 Pressure Tube," B. Watkins and D. S. Wood, 1968.

D-5 .12

UNCLASSIFIED D P

Study o:P waste disposal problems.

Design of equipment and system for chemical decontamination as dictated by the results of the above.

To provide specific data for evaluation of effectiveness of various reagents, a program for decontamination tests to be performed on V-12 valves which have been removed f r o m the reactor has been established. This would provide the following information :

Reduction in dose rate from valve after typical exterior spray application.

Reduction in dose rate from valve after exterior bath decontamination.

Reduction in dose rate from valve after A'and B porting interior decontam- inat ion

0 Reduction in dose rate from valve after A, B, and C porting interior decon- taminat ion.

Fabrication of fixtures to hold the valves during these tests will permit estab- lishment of standardized locations from which the exposure rate reading will be taken, thereby ensuring consistent results.

OPERATIONAL SUPPORT PROGRAM

The Design Criteria, "Protection of Equipment,lOg-N Building," which describe the operational and technical requirements of the capital funded modification required in 109-N to improve the Zone I11 ventilation, are ready for transmittal to the AEC. months of 1972.

This work is now planned to be accomplished prior to the sumher

In addition to the capital funded work, expense funded work is planned to be performed prior to the startup of N Reactor in 1971 so that some improvements in the 109 ventilation system will be achieved this summer in order to effect some reduction in the basement temperature. 1) increasing the fan speeds on the Zone I11 supply exhaust ventilation fans, and 2) removing the reheat coil from the basement air duct, and increasing the water spray capacity for the Zone I11 air washer.

This work consists primarily of

Project DCP-528, 100-N Area Fire Protection Improvements

At the request of the ABC, the Factory Mutual Research Corporation inspected the N Reactor facilities and prepared an extensive report evaluating the fire protection status of the facilities and recommending a number of improvements.

The scope of the subject project has been revised based on the Factory Mutual Research Corporation recommendations and an indicated fund limitation of about $340,000. The revised scope includes the following:

D-6 UNCLASSIFIED

UNCLASSIFIED DUN- 7 591

lo5 Building

a. Application of f i r e re ta rd ing mastic coating on bundled cables and cable t r a y s i n c r i t i c a l locat ions.

b. I n s t a l l a t i o n of f i r e stops i n cable or cable t r a y penetrat ions through w a l l s .

l o9 Building

a. I n s t a l l a t i o n of water spray system at t h e mezzanine end of t he steam generator c e l l s .

b. Application of f i r e re ta rd ing mastic coating on bundled cables and cable t r ays i n c r i t i c a l areas.

117 Building

Provisions of a w e t pipe spr inkler system f o r t h e c e l l divers ion door hydraulic power pack and main f loo r areas.

The design c r i t e r i a and pro jec t proposal have been transmitted t o AEC-RL f o r approval and a l loca t ion of FY 1971 funds.

NUCLEAR SAFETY ASSURANCE PROGRAM

Process Standards Audit in&

Nuclear Safety Audit - Fuels - A nuclear s a fe ty audi t of t h e f u e l s manufacturing area w a s completed. No standards v io l a t ions or i n f r ac t ions of nuclear s a fe ty procedures w e r e noted.

Nuclear Safety Audit - N Reactor - Auditing of t h e shutdown N Reactor and support f a c i l i t i e s continued. A spec ia l audi t has been completed which involved checking t h e Equipment Maintenance Standards records and Process Operation's s t a r t u p check sheets t o ensure t h a t equipment t e s t i n g i s being done according t o t h e scope and frequency of t h e Operating Safety L i m i t s document. This spec ia l audi t revealed no v io la t ions of t h e t e s t i n g requirements.

The audi t of t he N Reactor and support f a c i l i t i e s involved t h e following systems and spec i f ics .

1. Wergency Cooling Backup Sta tus 2. E l e c t r i c a l System Bus Sta tus 3 . Primary Cooling System Bus S ta tus

a. Primae.- _Temperature b. Process Tube Temperatures

4. Secondary System Pumping Status

UNCLASSIFIED D-7

UNCLASSIFIED DUN-? 591

a. Steam Generators in Use b. Dump Condensors in Use c. Raw Water Temperature to Dump Condensors

5 . 6. Graphite Cooling Status 7. Horizontal Control Rod Status, Rod Out of Service 8. Flux Monitoring System Status 9. 10. 11. Boiler Status

Circulating Raw Water Pump Status

Diesel Oil Storage Tank Inventory Coolant Water Storage Tank Inventories and Temperatures

Project DCE-539 - Nuclear Flux Monitor System The N Reactor nuclear flux monitoring system consists of three subsystems, each of which covers a specific range of flux. These are

Startup Range Subsystem

Intermediate Range Subsystem

* Power Range Subsystem

Less than optimum performance and excessive maintenance costs prompted the deci- sion to upgrade the system. The proposed work is required to improve the opera- tional reliability and maintainability of these systems.

.

This work will reduce spurious malfunctions which have an effect on operating continuity and reduce the related system maintenance and the corresponding personnel radiation exposure. In addition, the replacement system design will comply with the IEEE Criteria 279 fo r Nuclear Power Plant Protection Systems which is currently being used as a guide in the licensing of nuclear power plants.

The scope of work consists of making replacements and modifications to the startup range (subcritical), intermediate-range, and power range instrumentation system components and associated signal-receiving process instrumentation New electronic equipment and detectors will be used in the three subsystems.

After the January 28 announcement of the planned reactor shutdowns, work was stopped on all projects and fund authorizations were cancelled. Funds for this project were reinstated on June 1 and work has been resumed. detail design is estimated at 75 percent complete.

A s of month end

WASTE MANAGEMENT

100-N Ventilation Stack Monitor

To provide quantitative da5a on radioactive gaseous effluent releases would require monitoring of airborne and gaseous effluents from N Reactor. Original monitoring equipment in the 117 exhaust air filter building samples only the Zone I and Zone I1 exhaust for routine condition. The bulk of the exhaust air is from Zone I11 and is not filtered or sampled. The existing sampling equipment does not operate satis- Xactorily because of the excessive moisture in the Zone I exhaust.

D-8 UNCLASSIFIED

UNCLASSIFIED DUN-7591

A project w a s scoped t o provide an extensive monitoring system f o r both rout ine and a11 conceivable accident conditions. The system envisioned with remote readout i s completely automatic. The pro jec t w a s o r ig ina l ly pa r t of t he Efflu- ent Control Pro jec t , but w a s removed because of funding problems. Subsequently, a separate pro jec t proposal was prepared but w a s not implemented because of lack of funds. An a l t e rna te and less extensive sampling system has been developed t o provide f o r rout ine monitoring of gaseous radionuclide re leases t o the environ- ment. release of radionoble gases, radioiodines, rad iopar t icu la tes , and tritium oxide. Although the system i s primarily e f f ec t ive fo r rout ine monitoring during normal reactor operation, it would inherent ly incorporate ce r t a in fea tures with l i m i t e d capabi l i ty for accident analysis and evaluation.

The proposed system would provide qua l i t a t ive and quant i ta t ive da ta on the

The a i r sample would be passed through a pa r t i cu la t e and charcoal f i l t e r car t r idge which i s per iodica l ly taken t o the laboratory f o r analysis . The sample exhaust from the f i l t e r car t r idge would be cooled t o condense out t he moisture together w i t h any tritium oxide. analyzed i n the laboratory. I n addi t ion t o t h i s , an ion chamber would be located i n the base of the exhaust ven t i l a t ion s tack t o provide information on radionoble gases. Exhaust a i r flow measurements would a l s o be taken t o help evaluate the ana ly t i ca l results.

The r e su l t i ng condensate would a l s o be per iodica l ly

Design of the s implif ied monitor system i s complete and construction i s about &it% complete. Existing equipment from shutdown f a c i l i t i e s i s being u t i l i z e d .extensively.

TECHNICAL ACTIVITIES -.FUELS

Zircaloy Cladding Fabrication

Eleven b i l l e t s were extruded t o produce outer Zircaloy clad s h e l l s f o r Mark I V outer b i l l e t extrusions. The b i l l e t s , pierced nearly two years ago, were OD sleeved with .O3O-inch th i ck copper-silicon material, while t h e I D surface was flame sprayed w i t h a . O l O - . O l 5 inch l aye r of copper. Different f ront end con- f igurat ions were used t o determine t h e i r e f f ec t on extruded surfaces as follows:

A .030-inch th i ck end p l a t e welded t o the OD sleeve - on three b i l l e t s .

A l-l/b-inch 'tspun" end t o form a one-piece s leeve and end p l a t e - on three b i l l e t s .

A 1/2-inch ltspun'l end placed between two t h i n end p l a t e s . w a s placed ins ide t h e can and one outside. A weld on the I D joined t h e p l a t e s together - on f i v e bi l le ts .

One end p l a t e

A t the time the b i l l e t s were pierced it w a s an t ic ipa ted t h a t a .065-inch th ick copper s i l i c o n I D sleeve would be used f o r extrusion. copper layer on the I D surface i s considerably more economical but results i n a loose b i l l e t t o mandrel clearance. used a l/b-inch copper s i l i c o n r i n g was also used on occasions i n the be l i e f t h a t the r ing prevented chips from the graphi te cutoff from penetrat ing i n t o the r ea r of t h e extruded tube due t o loose billet-to-mandrel clearance. To confirm the value of the r ing , th ree b i l l e t s were extruded without t he r ing.

Flame spraying a t h i n

Previously, when the copper I D l ayer w a s

D-9 UNCLASSIFIED

UNCLASSIFIED DUN-7 591

It appears t h a t the defect length- i s 8-10 inches long on t h e th ree extrusions without t he r ing and about two inches with t h e r ing. as well as t he f ront lengths , w i l l be compared when the copper i s removed.

The r e a r defect lengths,

The optimum conditions from analysis of these extrusions w i l l be u t i l i z e d on subsequent outer clad fabr ica t ions .

Extrusion Tool R e l i a b i l i t y

Specif ic wr i t t en procedures f o r t h e inspection and reinspect ion of a l l extru- s ion too l ing , excepting containers and l i n e r s , have been i n e f f e c t fo r qu i te some time and have proven highly sa t i s f ac to ry . container and l i n e r growth ind ica te t h a t similar more spec i f i c cont ro l proce- dures would be worthwhile f o r l i n e r s and containers.

Recent problems with excessive

A number of f ac to r s bear on l i n e r l i f e and a l so on po ten t i a l t o o l f a i l u r e s . Included are t h e honing of t h e l i n e r I D , surface preparat ion following honing, and surface preparation of new l i n e r s and lub r i ca t ion during extrusion.

Liner l i f e and t o o l f a i l u r e s are a l so a f fec ted by excessive temperature and overstressing due t o extreme extrusion pressure. Maximum extrusion pressures a r e es tab l i shed and changed from t i m e t o t i m e depending on t o o l type and sizes and s p e c i f i c t o o l conditions. A l l of these f ac to r s a re being reviewed and analyzed t o e s t ab l i sh optimum procedures f o r ove ra l l coordination of t o o l rework, t oo l ing set-up, and extrusion parameters.

D-10 UNCLASSIFIED

UNCLASSIFIED DUN- 7 5 91

ADMINISTRATION - GENERAL

DECLASSIFICATION OF N REACTOR DOCUMENTS

The downgrading of N Reactor documents to Confidential and Unclassified was started with review priority being given to those potentially usable for the BPA study.

EMPLOYMENT SURVEY

DUN personnel totals and employee allocations as of June 30 and May 31 are shown in Appendix B.

UNION RELATIONS

On June 21 the membership of the Hanford Atomic Metal Trades Council accepted a proposed 9-month contract extension. The terms of the extension were:

1. A wage increase of $12.00 per week.

2. Agreement by the.Company to enter into Appendix "A" discussions within 60 days.

3. Commencement of interim discussions on or after November 1, in preparation for 1972 contract negotiations.

The extension agreement will expire on March 30, 1972.

SAFETY

No personnel radiation exposure exceeded operational control.

Month-end safety statistics were:

Disabling injuries - June - CY to date

Days since last disabling injury

Man-hours since last disabling injury

0

0

357

2,300,000

E-l. UNCLASSIFIED

APPENDIX A

PROJECT STATUS SUMMARY - REACTOR FACILITIES

AUTHORIZED PROJECTS

Authorized Number & Title Funds - $

Single-Pass Reactors

DAP-549, Rev. 1, Deacti- 160 , 000 vation of KE Reactor and 100-K Area

Y I P N Reactor

Per cent . Complete Design Construction

80 45

GCP-411, Rev. 3, Effluent 1.,960,000 93 Control Program - 100-N Area

DCE-519, Rev. 4, Replace- ment of Bridge Crane and Hoist System with New Crane System - 105-N Storage Basin Area

DCP-528, Rev. 1, Fire Protection System Improvements - 100-N

465 , 000 100

30,000 33

96

95

0

Remarks

Installation of the new water lines for fire protection has been completed and the lines tested. Title I comment draw- ings have been issued by Vitro for the, electric heat installation. Detail design is nearing completion on the No. 3 KE boiler control modification.

Work resumed June 15. on diesel fire protection. Dump tank instrumentation design complete. ATP on spray diesel jacket cooling completed June 22.

Design proceeding

Sixty-ton crane interlock installation essentially complete.

Revised project proposal based on AEC-RL request completed. Submitted to AEC-RL. I

-4 vl \o !-J

AUTHORIZED PROJECTS

Authorized Percent Complete Number & T i t l e Funds - $ Design Construction Remarks

DCE-539, Upgrade Flux Monitor System - N Reactor

335 , 000 75 0 Purchase Orders have been reopenecl for procurement of t h e con t ro l room panels and hardware. Delivery i s scheduled for l a t e Ju ly .

DAE-540, Smoke Density Monitors for DUN-Operated F a c i l i t i e s - 100-Areas

55,000 100 0 A l l work stopped January 28, pending dec is ion on N Reactor operat ion. posal revised t o include N only. mi t ted t o AEC-RL June 3 , 1971.

Pro- Sub-

DCP-542, High Pressure In j ec t ion System Improvements - N Plant

138,000 0 0 A l l work stopped January 28, pending dec is ion on N Reactor operat ion. Scope being reviewed for three-year campaign.

DAP-547, Permanent Markers for Buried Wastes i n Ter-

r minated 100-D, 100-F and RI, 100-H S i t e s I

28 , 000 8 0 Design i s under way.

PROJECT PROPOSALS AWAITING AUTHORIZATION

Number & T i t l e Funds Requested

DCP-527, Graphite Cooling & Fog Spray Improvements $ 97,000 100-N

Date t o AEC-RL

5/9/69

344,000 2/18/71

7 / 9 / 6 9

DCP-528, Rev. 2, F i r e Pro tec t ion System Improve- ments - 100-N

DCP-529 ;3ravity Drainage System and Disposal Basin for 100-N Area E m

cn H r H M U

200 , 000

PROJECT PROPOSALS AWAITING AUTHORIZATION

Number & Title

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

DCP-538, Heat and Ventilation System Improvement 105-N and 109-N Buildings

DAE-540, Rev. 1, Smoke Density Monitors f o r DUN- Operated Facilities, 100 Areas

DCE-551, Training Mock-up for V-12 Replacements, 185-D Building

DCE-552, Control Rod Coolant Recirculation Facilities, N Reactor

PROJECT PROPOSALS RETURNED BY AEC-RL

None

PROJECT PROPOSAL PREPARATION

Funds Requested

78,000

330,000

38 , 000 (Reduced Total)

49,500

235,000

Number & Title De sign, Criteria

Export Water System Backup - 182-~ ( f o r 200 Areas )

Completed

Stack Monitoring Improvements - 100-N Plant Completed

Date to AEC-RL

8 112 / 69

5/18/70

6/3/71

6/1/71

Project Proposal

Completed - Held by sponsor (ARHCO) pending mode of operation.

Prepared. An AR, however, has been approved to cover routine stack monitoring improvements portion of item.. This item therefore will drop from this report.

PROJECT PROPOSAL PREPARATION

Number & Title Design Criteria Project Proposal

Protection of Equipment, 109-N Building Completed Completed

DCP-528, Rev. 3, Fire Protection System Improvements, 100-N Area Completed Completed

DCE-519, Rev. 5, Replacement of Bridge Crane In preparation In preparation and Hoist System with New Crane System, 105-N Storage Basin Area

DAP-547, Rev. 1, Permanent Markers f o r Buried Wastes in Terminated 100-D, 100-F, and LOO-H Sites

Completed Activity Started

UNCLASSIFIED DUN-7591

APPENDIX B

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

CONTRACT 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

Total - Other Programs

To ta l Contract Personnel

COMMERCIAL ACTIVITIES PERSONNEL

TOTAL FORCE

F-5

794 9 -

803

34 6

40

-

F E

goo 9 -

909

36 7 - 43

18

861

-

UNCLASSIFIED

952

23 - 975

. .... . . , -. -. I