. .. . - . ,

Cover Sheet for a Hanford Historical Document Released for Public Availability

~~

Released 1994

Prepared for the U.SI 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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R I m m

02 R&D PROGRAM

RC 6-01

This document consists of

- 4'2 pages,- - /2 9pL-T HANFORD CATEGORY C-65 and C-57

SEMIANNuAt REPORT

SUMMARY

D e c e m b e r 31, 1969

RICHLAND OPERATIOMS OFFICE I

ATLAEFIC R I C H F I E L D WFORD COMPANY . - : ----..

DOUGLAS UNITED NUCLEAR, IMC.

PREPARED BY RICHLAND STAFF

coordinated by

CHEMICAL PROCESSIBG DIVTSIOEJ

and -.

PRODUC2ION REACTOR DIVISIOM

CGNTEHTS

INTRODUCTION

PROGRAM COST

BASIC PROOUCTIOn PROGRAM

P3ODUCT FI;EXZBILITY PBOGRAM

Enriched Fuel

Plut onium438

Other Isotopes

HVIRONM2NTAL & REGULATORY TECMJOLOGY PROGFW4

Nuclear Safety

Radiocuclides in Soils

Columbia River

WASTE VW?AGEE1ENT

IISLEREiUCES

CI STRI BljTI0N

A% 6 4 1 ?age 2

PAGE

3

5

a 21

2 1

-

22, 23

22, 24

20

26, 28

29

28

32

%I

). 1 -A

?

RL 6-01 Page 3

RICRLAMD FIVE-YEAR 02 R&D PROGRAM

The compiete semiannual R&E report consis ts of t h i s summary and two companion reports, one from DUN which covers reactors and another from ARECO which covers separationso These companion reports a re l i s t e d in t he set of references i n the back o f t h i s summary.

This semiannual report covers progress i n R&D during the first half of FY 1970 and i s based on plans outlined i n t h e June 30, 1969 annual report . The financial s t a tus i s shown i n t h e tab les following this introduction. seen t o be 45 percent of che $3.5 million R&D budget for t he year, which is sat isfactory. were i n the sarae order as the or ig ina l budget figures, t h a t is, i n the order of descending magnitude es follows:

Overall expenditures f o r the first half of FY 1970 ere

Expenditures under major program headings

Waste Management

Environmental arid Regulatory Technology

Basic Production

Product Flexibi l i ty

Within major program headings, there were some adjustments i n t h e d i s t r ibu t ion of f b d s . For example, reactors reduced emphasis on Pu-238 and Colmiia River Studies, and increased it on Other Isotopes (C-233) and Nuclew Safety;

Development work on the Basic Production and Product F l ex ib i l i t y ?ro- grams has been consistent with funding leve ls . bas been placed on I7 Reactor operational poblems as w e l l as on assessing the K reactor i n l e t cooling piping in t eg r i ty and z i r con im hydriding problems. plant extraction performance by in-plant t e s t i n g of t he "Zirf lex Tonic" e f f ec t e

Increased enphesis

Work has been i n progress t o improve t h e Pwex

RL 6-01 Page 4

Gn t h e Product F l ex ib i l i t y Program, developnental work was under XL~- t o 2s tab l i sh e i the r enriched uranium or nondefense plutonium (or both) as a reac tor Puel and t o provide fu r the r technology f o r t h e production of high-pwity p.Iutoni~~t1-238.

Ac t iv i t i e s on the Environmental and Regulatory Technology Program are engaged i n developing t e c h n o l o o t o assure safe operation of t h e Haaford project faci l i t ies . Major emphasis w a s placed on the study of metal-water ( i r r ad ia t ed N fue l s ) react ions t o simulate loss of coolant conditions, and with a t ten t ion t o t h e r e l a t ed sa fe ty assur- ance features. capab i l i t i e s for detemfning t h e consequences of accidental re lease of radioactive materials.

Work is underway t o develop accurate pred ic t ive

Increased emphasis has been placed on t h e Waste Management Program with in ten t t o develop technology f o r a su3s tan t ia l reduction i n releases of low- and intermediate-level radioactive waste t o t h e environment. The salt waste processing and the waste so l id i f i ca t ion in-tank subprograms have both been reduced i n l e v e l of R&D e f f o r t f o r t h e reason t h a t B Plant and the in-tank so l id i f i ca t ion u n i t s are now operating successfully and do not need fur ther R&D support. Emphasis has been maintained on t h e development program f o r purif'ying and encapsulating fission-product strontium and cesium. Ef fo r t s t o reduce the salt content of Purex wastes have been increased b e c a u e of the projected waste management cost savings and concomitant improved strontium recovery-

--

Program

02 R&D PROGRAM COST (Dollars in Thousands 1

fl 1970 Budget

Basic Production 735

Product Flexibility 300 Transplutonium - Plutonium-238 19 5 Other Isotopes 5 Enriched Fuel 100

Environmental & Regulatory Technology 1040 Nuclear Safety ’740 Radionuclides in Soils 2 50 Columbia River 50

1425 Waste Management - Total 3500

Expenditures Estimated to 12-31-69

$ of ~udget

337 46

110 37 - - 63 32 16 320 30 30

549 53 *35 * 5-9 108 43

6 12

- 579

1575

41 - 45

L ht

RL 6-01 Page 6

02 R&D PROGRAM COST (Dollars i n Thousands )

REACTORS ONLY

Program

Basic Production N Reactor OFerability K Reactor Risers K Reac, Zirc. Hydride Fuel Performance

Characterization Reactor Plant L i f e P l a n t Cost Reduction Decontamination Studies Oper. Support Services

Product F l e x i b i l i t y Enriched Fuel ~lutonium-2 38 Other Isotopes

Environmental & Regulatory Technology

Nuclear Safety Regulatory Tech Nucl, Safety A s s u r .

Columbia River

FY 1970 Budget

480 154

28 -

95

163 - -

40

200 102

65 33

460 234 26

Waste Management

T o t a l

12 5

1525

- -

Expenditures Estimated t o 12-31-69

% of Budget

40 70

8 - 45

29 -

29

73

10

- -

31

4; - 25

74 37 30 30 14

-397 55

329 62 6

29 46 42

71 26 23

23 - 684

18

45

-

Note: Overhead costs a re included i n t h i s tabulat ion.

RL 6-01 Page .7

Program

02 R&D PROGRAM COST (Dollars in Thousands)

SEPARATIONS ONLY

FY 1970 Budget

Basic Production 255 Proc. Devlp. Xxtrac. Process Devlp. Pu Matl. Equip. & Inst. Nondefense Pu

Product Flexibility Transplutonium Plutonium-238 Other Isotopes Enriched Fuel

100

Environmental & Regulatory Technology Nuclear Safety Atm. Dispersion Seismic Invest,

Radionuclides in Soils Radio. Trans. Model Soil Waste Reactions Field Meas. Methods

320

1300 -

Waste Management Lou & Inter. Level

Salt Waste Processing Waste Solidif, in Tanks Waste Packaging S a l t Waste Vol. Reduc. Alter. Solidif, Methods

Waste

95 5

Total 1975

Expenditures Estimated to 12-31-69

% of Budget

147 58 60 .

34 40 13

36 1 33 2

36

152 47

20 24

556

43 22 43

43

74 168

- 45

RL 6-01 Page 8

BASIC TRODUCTIOZ PROGRAM

This program covers that research and development work which is conducted in support of plutonium production, this pertains primarily to the production of weapons-grade plutonium; while for E Reactor, it pertains primarily to the production of plu- tonium of 12 percent Pu-240 content. the work provides for the isolation and purification of the reactor- produced plutonium. The work involves activity in a wide variety of fields such as physics, thermal hydraulics, materials technology, fuels technology, and separations chemistry. The program spans the range from basic research to in-plant development and new process demonstration.

For the K reactors,

For the chemical facilities,

Objectives and Incentives

Advancements in basic production technology provide the following benefits :

1. Improved product quality.

2, Reduced costs.

3. Incressod plant utilization.

4. Extended plant life.

5* Improved plant performance.

6 , Reduced product losses,

In this way, the Richland complex can continue to perform consistently with bcth the near-tern w d the long-term requtrements.

Progress W i n g Report Period - Reactors N Reactor Operability

The object of this activity is to reduce unscheduled shutdowns and! extended outages so that a high level of operationd relizbility will be achieved.

RL 6-01 Page 9

V-11 Valve Development and Evaluation

Alternate seal designs w e r e conceived, fabr icated and tested f o r V-11 valve development and evaluation, An engineering study and analysis w a s completed f o r t he specif icat ion of primary loop COM- ponents. Engineering and economic analyses were completed f o r evaluation of t h e var ious design approaches. Design and fabri- cat ion of t he "sleeve ins ide the nozzle" concept w a s completed, and off-reactor t e s t i n g w a s started, An engineering study and evaluation of sleeve i n s t a l l a t i o n equipment w a s started, The Mark I prototype cap remover was designed, fabr icated, and suc- cess fu l ly t e s t ed on-reactor.

V-12 Valve Development and Evaluation

An engineering study and analysis has been made t o def ine the leakage problem of t h e V-12 valves and t'o develop correct ive concepts, and the components which w e r e fabr icated from corrosion r e s i s t an t material are being tested, A unique concept of a multi-ply m e t d l i c be lhws seal to eliminate t h e valve stem packing leak- age w a s designed, fabr ica ted , and successfully tested off-reactor i n a l i f e cycle test equivalent t o 20 years of operation a t loop temperature and pressure.

Redesign of t h e present actuator has been completed,

System I so la t ion Techniques

Work continued on development of t h e "balloon" concept of system i so la t ion whereby a boss would be added t o t h e object pipe through which a f l ex ib l e , high s t rength balloon OF bladder would be in- serted and i n f l a t e d t o seal off coolant flow. T h i s would supple- ment t h e freeze technique where coolant flow through t h e l ines is too great, t o permit freezing.

Exposure Reduction

An engineering study f o r exposure reduction and evaluation was con- t inued to achieve a better def in i t ion of rad ia t ion sources and exposure levels. fabricated, and used i n both the f ront and rear pipe spaces, Location md in tens i ty of radiat ion levels have been estab- l ished, and several concepts fo r custom shielding have been designed.

Special columnated detectors were designed,

RL 6-01 Page 10

Zone I Stainless Steel Lines - Development and Evaluation A nuclear safety study was conducted to determine whether or not there were any problems attendant to deteriortion of the Zone I stainless steel sensing lines. It was concluded that in all cases, except in flow monitoring, sensing line failure would be fail-safe. For individual tube flow monitoring, close surveillance and cor- rection of Leaking sensing lines should prevent a safety problem from developing.

Evaluation of the Fresent primary raw water screening system effec- tiveness was completed and recommendations ere being reviewed. Raw water quality criteria are n e a r l y complete, system modification schedule guidelines on a priority basis was completed to provide the basis for decisions relative to which systems should be considered when planring backup screening. proved freeze protection has been implemented in the primary screening facility.

A document outlining

Im-

Flux Monitor Systems

All nuclear flux monitor instrument systems were modified to assure operational reliability. Prototype chambers with new seal and cable configurations were received, and operational tests were started. completed, and a laboratory prototype was fabricated. A specifi- cation for intermediate range readout instrumentation was pre- pared, and a market survey was Conducted. Testing of a vendor supplied intermediate range instrument was completed, A new sub- critical chamber drive system was designed, and fabrication is 90 percent complete.

Design of revisions to the readout instrumentation was

Reliability Appraisal Efforts

Reliability appraisal efforts were started on the confinement, the AC electrical power, the primary coolant4 and the emergency raw water systems. fault-free and critical failure modesp mechanisms, and effects analyses. The reliability analysis in support of high pressure injection and seal water subsystem control was also completed.

Individual system models are being developed for

Appraisal of the fog spray subsystem has been completed.

IU 6-01 Page 11

B a l l Safety System

Work accomplished included addition of the dryer t o the t r i p mechanism air purge system, accumulation of r e l i a b i l i t y data , conversion of t h e cocking and locking devices t o manual operation, continued development of a l t e r n a t e t r i p mechanisms, and completion of an engineering evaluation of the ball recovery system.

A and B Bus Independence

Operational and shutdown t e s t i n g of the c i rcu la t ing r a w water system was completed t o determine safe operating conditions f o r N Re8CtOr while t h e A-B Bus i n t e r t i e study is progressing. were performed which provide t h e necessary da t a t o develop a s e t of operating requirements which meet t h e independent e l e c t r i c a l system c r i t e r i a under c e r t a i n conditions of r i v e r temperature and elevat ion, of assuring A-B Bus independence during operation and shutdown w a s completed.

Tests

A preliminary engineering study on a f eas ib l e method

Horizontal Contrcl Rod System

Performance of t he l e f t - s ide ha l f of the control rod system has been monitoPod following conversion, afu the mid-year, of t h e hy- d rau l i c f l u i d t o EL water-in-oil emulsion, Several slow rods were noted, analyses w e r e performed, and correct ive act ion w a s taken, Buildup i n the bticteria content of t he emulsion w a s detected, aEd a study i s underway t o provide an e f fec t ive bacter iacide.

K Reactor Risers

The object of t h i s a c t i v i t y i s t o assess t h e i n t e g r i t y of t he K reac- t o r i n l e t coolant supply piping, define reac tor operating modes con- s i s t e n t with piping system l imi t a t ions , identi-fy and car ry cu t improvements required tQ. enhance the in t eg r f ty of t h e K reac tor piping, and r e tu rn t h e r eac to r t o a normal operating mode consis tent with r e s t r a i n t s which a r e found t o 'De =.equired,

Accamplisbents

Planning w a s completed f o r replacement of t h e e i g h t 2b-hch bj 36-inch wye J o i n t s i n t h e i n l e t piping systems. t o seismic analysis of ;piping, res idual crack detection, and proof

Progreqs related

RL 6-01 Page 12

tes t planning e i t h e r continued o r were started. The f i n a l e v d u a t i o n of f r ac tu re mechanics w a s e s sen t i a l ly completed and t h e t r a n s i e n t pressure analysis i s being prepared. Basic methods f o r development of technical bases f o r tQbe power l i m i t s f o r K reac tors t o prevent f u e l clad Icelting in t h e event of coolant l o s s due t o i n l e t riser f a i l u r e were developed and applied.

K Reactor Zircaloy Hydride

The obgect of t h i s e c t i v i t y i s t o determine t h e causes and necha- nisms contr ibut ing t o t h e hydriding of K reac tor Zircaloy Frocess tubes, promote measures t o minimize addi t iona l hydriding, and de- velop methods and techniques fo r hydride case layer removal f o r r e s to ra t ion of t h e base m e t a l .

Mechanisms

Further ins ight i n t o t h e mechanics of the hydriding phenomenon w a s gained from t h e current Zircaloy coupon s tudies . Preliminary data ind ica t e t h e following trends:

1. Bare aluminum dummy elements i n Zircaloy tubes have the greatest hydriding poten t ia l . will cause some hydriding i f good e l e c t r i c a l contact is maintained. Regular f u e l supports do not give a good contact. Although s t a i n l e s s s teel itself w i l l . not cause hydriding, it w i l l nat prevent f w t h e r hydriding when a case layer i s already present ,

Anodized aluminum dummies

2, A case layer of hydride p l a t e l e t s on t h e inner tube sur- face causes t h e continued buildur, of hydrogen i n the base m e t a l

3* Removal of a l l t r aces of a case layer from previously hydrided surf_aces r e s to re s them t o a condition where anodized aluminum spacers w i l l prevent fu r the r hydriding.

Case Layer Removal

Development of an e l e c t r o l y t i c method of case layer removal t o sup- p lan t t h e g r i t b l a s t ing process w a s continxed, It was demonstrated t h a t t h e case l aye r can be removed at a rate of 0.1 m i l per minute using a current densi ty of 0.1 ampere per square c e n t b e t e r .

RL 6-01 Page 13

However, t he process leaves an objectionable t race of hydride needles covered by an anodic oxide film. secondary electrochemical e tch t o remove the t race.

Work i s under way on a

Tube Examination

The downstream 66-inch lengths of 25 Zircaloy tubes were removed from C reactor f o r use i n an electrochemical cleaning development. Samples from the tubes revealed very e r r a t i c hydriding, more so than a t the K reactors .

Fuel Performance Characterization

The object i s t o conduct a f u e l development program aimed a t fue l monitoring and u t i l i z a t i o n , f ie1 cladding performance improvement, elimination of f r e t t i n g failures, supporting t rans i t ion t o full- sca le Mark I V usage, ver i fying technica l f e a s i b i l i t y of long hot- die-size fue l , supporting t r a n s i t i o n t o ful l -scale hot-die-size usage and reducing N f u e l cos t .

Fuel Monitoring and Ut i l i za t ion

Routine f u e l monitoring w a s continued, I n the l i gh t of lower exposure operation m d improved measuremcot cap&i l i t i e s , a review of current monitoring requirements has been s tar ted. The charging of four types of Class II category fuel (minor defects) i n the f r inge and upstream cen t r a l tube posit ions of N Reactor was authorized. The preparation of a document w a s i n i t i a t e d which w i l l describe the object ives , bases, and administration of the HDS Class I1 fue l s program f o r t he K reactors , The premise i s t h a t both cost savings and problem-free e-qerience can be achieved by use of hot-die-sized f u e l i n the Class I1 category,

Fuel Cladding Performance Improvement --

A modified fie1 support design has been selscted t o alter coolant flow cha rac t e r i s t i c s i n t h e vicinity of supports on A l S i elements i n an attempt t o rrinimize fuel f a i lu re s resul t ing frcm erosion- corrosion adjacent t o the supports, Procurement has been s t a r t ed so t h a t fuel equipped with the improved support may be in-reactor by t h e summer of 1970. A program was developed f o r establishing the proper cladding a l loy condition p r io r t o fue l fabr icat ion by t h e hot-die-size p- A ocess.

e

RL 6-01 Page 1 4

Elimination of Fret t ing Failures

S igni f icant and detai led study of factors and mechanisms contrib- u t ing t o N fue l f r e t t i n g failures was accomplished, and an inter im repor t i s Seing issued. Apparently work hardened Zircaloy buildup on t h e support shoes was providing a cu t t ing t o o l t h a t made the f r e t t i n g debris. A set of c r i t e r i a f o r sfmulation of support environment and i t s f’unction w a s derived and used t o design a support w e a r t e s t machine which i s being constructed.

Transi t ion t o Mark I V

Conversion of t h e cent ra l zone of N Reactor t o Mark I V fue l i s approximately 70 percent complete. The swelling charac te r i s t ics of t h e Mark I V fue l s discharged t o date have been analyzed, and the results show t h a t alloy-503 cores s w e l l less than do the design cores vhich w e r e made of B r i t i s h alloy-601. f u e l s s w e l l with increasing enrichment and a lso with increasing clad thickness. An evaluation i s i n progress t o assess, reactivity-wise, t he r e l a t ive merits of t he Mark TA and the proposed Mark IVAA designs. A t es t i s being prepared t o irradiate 25 Mark I V fue l columns i n the cent ra l zone with exposures ranging from 3000 t o 7000 MWD/T t o determine the inherent design strengths and weaknesses of t h i s fue l model.

Both of these alloy

Technical Feasfb i l i ty of Long Hot-Die-Size Fuel

An i n i t i a l 42 columns of 11.23-inch7 94 metal, hot-die-size fue l was charged i n t o a K reactor t o es tab l i sh the technica l feasi- b i l i t y of longer f u e l which would have s igni f fcant economic advantages. Two d i f fe ren t sizes of aluminum drunmies, one simulating the 8-inch model and the other simulating t h e pro- posed If-inch model, w e r e charged separately i n t o the same process tubes f o r a comparison test . n i f i can t difference between the two insofar as moderator dis tor- t i o n w i t h resu l t ing tube d is tor t ion i s concerned,

mis t e s t showed no sig-

Transit ion t o Hot-Die-Size Usage

Tentative schedules have been made up for t he t r a n s i t i o n from AlSi canning t o hot-die-size usage and are based on t h e aot ic ipated adequacy and acceptabi l i ty of long f u e l perforclance.

RL 6-01 Psge 15

N Fuel Cost Reduction

Fabrication development and ex-reactor t e s t i n g of protot;ypic unbonded end closure designs were accomplished i n an effort t o reduce N fue l cost . End support development w a s concentrated on fabricat ion ar;d ex-reactor flow loop t e s t i n g of an inner support system which would replace only the conventional inner support hardwake. a l te rna te Si l le t fabr icat ion techniques continued a t a r e l a t i v e l y l o w l e v e l of e f f o r t .

I r r ad ia t ion t e s t i n g and fabricat ion processing of

Reactor Plant Life

The object of t h i s a c t i v i t y i s t o assure t h a t t he a b i l i t y of t he fianford reactors t o produce i r r ad ia t ion products economically does not become constrzined by irrespfaceable f a c i l i t i e s . faci l i t ies a re defined as those which cannot be replaced without essent ia l ly rebuilding t h e e n t i r e complex ( the graphite stack o r reactor shield fo r example), o r those which are so expensive t o replace t h z t economic considerations suggest reactor deconmissioning as the more accepta5le a l t e rna t ive (external piping complex o r water treatment system f o r example).

Irreplaceable

Water Chemistry and Corrosion - K Reactors

The half-plant t es t of 1 .0 versus 0.5 ppm dichromate additior: t o t h e process water i n t h e K reactors was caropletcd. The 35 evaiu- a t ion f u e l c o l m s are being examined. As a r e s u l t , t he reactors are being operated on interim specif icat ions permitting 0.5 ppm dichromate and 6.7 pH at below 90°C coolant o u t l e t , a d 1 ppm and 6.6 pH above 90OC. Laboratory tes ts t o evaluate the e f f ec t s of coolant chemistry variables on t h e erosion-corrosion character- istics of aluminum f u e l cladding al loys were terminated f o r budget reasons, During t h e tests, a laborato,ry technique w a s developed fo r use in screening alloys and/or process water compositions with respect t o corrosion-tendencies. a l te rna te methods of preventing p i t t i n g m d tuberculation i n l o w veloci ty carbon steel piping were continued,

Laboratoqy s tudies t o compare

Water Chemistry and Corrosion - N Reactor

An f ~ . ~ o n e i - 6 0 0 tube wiis removed from stemi generztw 6A of N Reactor after 40 months' residence and 24 months' operation i n

RL 6-01 Page 16

the continuing program of steam generator tube corrosion surveil lance. removed.

The tube i s being compared t o others previously

Seal sect ion tubes from N Reactor control rods 46 and 34 were removed f o r destruct ive examination. Two problems, loss of cooling flow and progressive p i t t i ng , with respect t o t h e control rod cooling system were defined.

A t es t program at N Reactor t o measure the in-reactor corrosion of acid s ta ined Zircaloy-2 w a s started. P l a s t i c r ep l i cas of the steel nozzle Zircaloy process tube junction taken before and after the 1969 through-reactor decontamination have confirmed the bore- scopic observation t h a t scat tered local ized corrosion occurred on the carbon steel adjacent t o the tube.

Graphite - K Reactors

Samples of graphite from KW reactor continued t o be irradiated i n the Engineering T e s t Reactor.

An in-reactor program of tube diameter measurements and probing w i t h f ue l s i z e probes has confirmed tha t process tubes are being f l a t t ened a t some trunion block locations by trunion block shrink- age onto t h e tube. continued.

Oxidation monitoring of t he graphi te stack

Graphite - N Reactor

Dimensional measurements of irradiated fu l l - s ize graphi te bars and laboratory-size samples i n N Reactor showed t h a t dimensional changes i n the a l l - s i z e bars were subs tan t ia l ly grea te r than those occurring i n t he laboratory samples. were inspected i n addition t o rout ine measurement of stack move- ment. Two were found t o be i n good condition, one had a 1-inch misalignment , and t h e -fourth had an apparent bowing. Oxidation monitoring of t h e graphite stack was continued.

Four ball channels

N Reactor Process Tube Surveillance

Nine N Reactor tubes were borescoped. No de t r imen td e f f e c t s due t o in-reactor service were noted.

RL 6-01 Page 17

K Reactor Aluminum Tubes

Examination of leaking aluminum tubes removed from t h e KE reactor revealed those tubes with cha rac t e r i s t i ca l ly t h i n w a l l s and low r i b s which have reached t h e end of t h e i r normal corrosion l i f e .

N Reactor Primary Pipe Survei l lance

High s t r e s s areas were u l t r a son ica l ly examined t o determine whether o r not there were weld flaws i n t h e N Reactor primary piping, and i f so whether o r not t h e f l a w s were growing. detected

No flaws have been

Plant Cost Reduction

The object of t h i s a c t i v i t y i s t o adapt ex is t ing technology o r de- velop new technology which, when implemented, w i l l r e s u l t i n a d i rec t reduction i n plant operating cos t o r avoidance of costs which can be projected on t h e b a s i s of current experience and conditions.

Accomplishments

The reactor fiystems study w a s completed. I n se lec t ing systems o r components f o r appl ica t ion of automatic survei l lance, the principal emphasis w a s placed on Process Standards, systems c r i t i c a l t o nu- c l e a r safety, i d e n t i f i e d problem areas within t h e process audit check l i s t , and consol idat ion of control i n the 105-N Control Room. Software programs have been iden t i f i ed and progran require- ments specified. upon performance and r e l i a b i l i t y requirenents. elimination of t he cont ro l computer, t h i s work i s being discontinued.

Hardware requirements have been established based Due t o budgetary

Decontamination Studies

The object of t h i s a c t i v i t y is t o evaluate effect iveness , corrosivity, and cost of ava i lab le decontaminating reagents; t o develop pract ical , safe and economic procedures f o r t h e i r use; and t o develop and eval- uate possible new approaches t o t h e a t t a in ing of reduced radiat ion l eve l s and attendant personnel exposwe.

G

I

RL 6-01 Page 18

K Reactors

A program was s t a r t e d t o determine the p rac t i ca l i t y of adapting through-reactor decontamination t o the K reactors. Material com- p a t i b i l i t y tests were performed t o assure acceptable corrosion rates for a l l reac tor materials i n contact with the decontami- nation solutions. Tests were made t o evaluate decontamination effect iveness as a function of solution temperatures, reagents, so lu t ion concentrations and application times, as w e l l as valving and cont ro l procedures and t h e use of externally heated solu- t i ons . ex te rna l ly heated t o 85O-9OoC.

The bes t results were obtained with Turco 4528 solution

N Reactor

The t h i r d successful through-reactor decontamination o f N Reactor w a s car r ied out. Steam generators i n c e l l No . 1 were decontami- nated as w a s t he N o . 4 graphite coolant system heat exchanger. Screening tests of commercially available decontaminating reagents were continued.

Operational Support Services

The object of t h i s a c t i v i t y i s t o maintain continuity of reactor plant production and t o provide the technical support required t o maintain the plant i n a high state of efficiency. This i s t o be accomplished primarily through the maintenance and improvement of physics computer codes and through development work associated with product assaying problems.

Physics Code Development

The input sect ion f o r using the WIGL2 two-group k ine t i c s physics code on t h e UNIVAC 1108 computer has been largely completed, and subroutines have been-tested by using sample cases. m a n u a l for the FLARE: flux dis t r ibu t ion and control computer code i s being prepared. An analysis of performance of t he HAMMER l a t t i c e computer code was completed, and minor problems with t w o subroutines were found.

An input

Product A s saying

Arrangements hzve been made f o r individual batch dissolut ion and sampling of PT-NR-95 material fo r product assay which conteined higher than noma1 i n i t i a l uranium-236.

RL 6-01 Page 19

Progress During Report Period - Chemical Separations Process Development - Extraction Laboratory Studies demonstrated that fluoride, when added to the feed just prior to extraction or when the feed is digested in the presence of fluoride, is effective in improving the decontamination factor .(DF) for zirconium by the "Zirflex tonic" effect. The major improvement in the zirconium DF occurs in the first cycle codedecon- tamination column ( H A ) .

The processing of long-cooled (2-year) Fast Flux Test Facility fuels in the Purex Plant was studied. sible. million would be required to modify a dissolver cell for the task, Hcwever, the fuel could be received after 60 days' cooling (with a lead heat-transfer media) and stored for processing after 180 days' aging for an additional $200,000. A transfer cask was estimated to cost $150,000.

This processing appears to be fea- It is estimated that a capital investment of about $1.6

Neptunium Chemistry

Rate accelerating material (M) for neptunium(V) oxidation has been prepared in engineering-scale prototype equipment. to achieve a product with a yield amounting to 85 percent of that achieved on a laboratory scale. Further work on RAM has been de- ferred pending evaluation of recent new possible improvements in neptunium recovery.

It was possible

Equipment, Instrumentation and Materials

Analytical methods studies indicate that x-ray fluorescence has a high potential for successful application in monitoring neptunium and plutonium in the Purex extraction column waste. This method generates the fewest installation problems, is highly selective, and may require the least amount of separations. The method and instrumentation is approaching the stage of final evaluation testing by the use of Purex waste samples.

A specific gravity and conductivity analysis can be used as a method to accurately determine the concentration of nitric acid in the rec- lamation facility slag and crucible dissolver. Development 0% in- strumentation for the measurement of these parameters under process conditions is nearing coqletion.

RL 6-01 Page 20

3arium-133 shows advantage over cesium-137 and has been adopted as the isotopic source t o nieasure self-absorption i n bulk plutonium package samples being analyzed by gama spectroscopy. The optimum lithium-drifted germanium detector system for nondestructive pluto- nium assay can measure as l i t t l e as 1 gram of plutonium i n an 18" x 18" x 24" waste carton with a precision of -+ 5 percent.

Evaluation of Effort

Expenditures under the Basic Production Program f o r t he f i s c a l year t o date zre 46 percent of the FY 1970 budget fo r t h i s program.

Progress has been i n general conformance with or iginal planning and a l loca ted funding.

Defini te progress has been made i n defining B Reactor deter iorat ion problems and i n providing solutions. Licensabili ty i s s t i l l required. Assessment of K reactor i n l e t cooling piping i n t e g r i t r and the zirco- nium hydriding problems have been defined and solved, and corrective measures remain t o be put i n to e f f ec t . Conversion t o Mark I V N Reactor f u e l i s w e l l developed and proved f o r 94-metal elements, but evaluation of the spike elenent i s s t i l l i n progress, s i ze fuel has been successfully developed, but t e s t ing remains t o be completed. This work appears t o be on schedule for the changeover from AlSi t o hot-die-size fuel. and graphi te problems have continued. reac tor graphite trunnion block shrinkage is being studied t o deter- mine a solution. provides a sound bas is f o r reducing personnel exposure.

Long (11.23-inch) hot-die-

Evaluation of routine plant corrosion A new problem involving K

Studies have shown tha t decontamination development

Lzboratory s tudies have confirmed t h a t zirconium decontamination can be increased by adding f luoride t o the %rex coextraction column feed and also t h i s may enhance neptunium recovery. Developnent work on t h e rate accelerating material f o r neptunium(V) oxidation has been deferred, pending evaluation of improvements i n neptunium recovery. --

IFIE

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PRODUCT FLEXIBILITY PROGRAM-

The purpose of t h i s program is t o develop production capabi l i t ies other than tha t of weapons-grade plutonium.

Objectives and Incentives

The i n i t i a l objective was t o develop the f u l l production potential of t he Richland s i t e . The individual R&D items included were tritium, transplutonium, plutonium-238, other isotopes, and enriched fuel. Ef for t s within the reporting period have been directed toward the development of highly-enriched (oral loy) fuels t o be used as the sole f u e l o r i n conjunction with s l i g h t l y enriched uranium fue l elements (spike concept) t o produce a desired product mix i n the reactors. Development of the technology f o r t he production of medical-grade plutonium-238 includes t h e advantage of the reactors favorable ther- m a l neutron spectrum. The incentives are , for example, improved production capabili ty, reduction of product un i t costs, and bene- f i c i a l safety analyses. Since t h i s program i s composite, each sub- item has a unique objective.

Progress During Report Period - Reactors

Enriched h e 1

The purpose of t h i s a c t i v i t y i s t o develop and demonstrate technol- ogy f o r the use of enriched uranium o r nondefense plutonium 85 reactor fie1 and t o perform t h e nuclear safety elvations required for the use of enriched fue ls .

Oralloy

Planning f o r the oral loy tes t load, which involves charging 20-30 columns of fuel i n the upper enriched r ing of a K reactor, was completed, ment work on fie1 element fabr icat ion has been s tar ted. Tube power limits have been determined and loading patterns estab- l ished.

The required oral loy has been obtained and develop-

Planning f o r loading of the t e s t is under way.

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Plutonium

A prel ininary evaluation has been made on the use of nonweapon plutonium fuel as spike enrichment. Preliminary planning iden t i - fying methods of allowing a nominal. annual throughput of 200 kg without introducing nuclear sa fe ty problems has been done. De- t a i l e d s tudies a re being s ta r ted .

Plutonium-238

The purpose of t h i s ac t iv i ty i s t o demonstrate continued capab i l i t y t o produce medical-grade Pu-238, t o develop improved technology f o r medical-grade Pu-238 production, and to evaluate a l t e rna t ive methods f o r producing t h i s isotope.

Accomplishment

The 1200-gram neptunium i r r ad ia t ion t e s t consis t ing of f i v e columns w a s completed. One column of aluminum-clad neptunium-graphite matrix t a rge t elements experienced f a i l u r e s i n all s i x elements; one cen t r a l zone and one f r inge column of zirconium-clad neptunium- graphi te matrix elements were discharged even though no f a i l u r e s had been experienced. The two cent ra l ly charged aluminum matrix columns were irradiated t o e s sen t i a l ly goal and discharged. The graphi te matrix elements have been analyzed, but t he aluminum matrix elements have not. matrix elements w a s ax ia l growth of t he wafer, f o r some undeter- mined reason. charged i n KE reactor i n M a y 1969, i s continuing.

The cause of t h e f a i l u r e of the graphi te

The i r r ad ia t ion of the 150 grams of americium oxide,

Other Isotopes

The obgect of t h i s ac t iv i ty is t o enhance capabi l i ty of t h e Xanford reac tors t o produce various isotopes and t o determine t h e cost and capabi l i ty of production f o r those isotopes not included i n . o t h e r a c t i v i t i e s ,

cobalt-60

Calorimetric measurements were completed on a l l C reac tor overbore cobal t t a rge t s , nominally 51 percent cobalt) which w a s i n s t a l l e d i n €3 reac tor has now oDerated over s i x months and has been recharged twice.

The cobalt base a l loy tube (Haynes Alloy-25,

There

RL 6-01 Page 23

have been no significant problems. strips irradiated in N Reactor were

Battelle-Northwest cobalt discharged.

Lithium-Bearing Spline Development

Spline samples containing lithium in three different chemical forms were received and are now being irradiated. The forms include lithium-manganese alloy particles dispersed in a powder metallurgy matrix, lithium-carbonate crystals dispersed in a powder metallurgy aluminum matrix, and lithium-fluoride crystals dispersed in a powder metallurgy aluminum matrix. All smples are clad with 3-6 mils of aluminum.

Plutonium Burning

The 13-column slab of plutonium-aluminum alloy fuel elements in KE reactor continued to operate without incident. The reactivity of the block has decreased steadily allowing replacement of essen- tially all of the 94 metal and thoria with natural uranium.

Progress During Report Period - Chemical Separations Plutonium-238

One kilogram of neptunium was converted from nitrate to oxide for use in further target fabrication and irradiation studies. The neptunium was denitrated continuously by calcination in a heated tube provided with a conveying scraper blade agitator. Although product quality was adequate and operability was considerably im- proved over that of the vibrating tube calciner, the performance required of a production unit was not achieved.

Due to scheduling of the integrated program, no plutonium-238 was processed during the report period.

Chemical flowsheets for-either coextraction or consecutive extrac- tion of neptunium and plutonium-238 from nitric acid solutions of irradiated aluminum-neptunium oxide targets have been devised and demonstrated, Unexpected difficulties were encountered in parti- tioning neptunium from plutonium-238, and work is in progress to improve this process. ated most efficiently with dilute plutonium-238 feed.

-

The second and third plutonium cycles oper-

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Separation requirements t o provide capabi l i ty f o r t o l l i r r ad ia t ion were reviewed. $2,100,000 f o r t he 234-5 building. kilograms of neptunium i s required f o r processing i n the Purex Plant. capabi l i ty can be provided i n Pac i f i c Northwest Laboratory f a c i l i - t i e s f o r an estimated s ta r tup cos t of about $600,000.

Capital costs estimated were $950,000 f o r Purex and A campaign s i z e of at least 100

When production rates are less than 100 kilograms per year ,

Other Isotopes

A minor amount of work was performed t o scope t h e recovery of precious metals from Hanford wastes. It was found t h a t act ivated carbon se lec t ive ly sorbed palladium from a lka l ine waste. I n i t i a l palladium breakthrough on act ivated carbon occurred at 72-bed volumes (BV) , 50 percent breakthrough occurred a t 94 BV, and 100 percent breakthrough occurred at 110 BV. Rhodium w a s not renoved from the feed by the carbon. Technetium w a s held by the carbon for 1 0 BV with 100 percent breakthrough a t 40 BV. netium i n t h e palladium-free e f f luent were recovered on IRA-401 ( t r ade name--Rob & Haas Company) anion exchange r e s in . Palladium w a s removed from the carbon column by treatment with hot water and d i l u t e ammonium hydroxide.

Attempts t o recover rhodium from Hanford a lka l ine wastes using mossy zinc o r mossy t i n were unsuccessful. a lka l ine wastes were contacted with zinc metal powder on a batch bas is f o r 24 hours.

The rhodium and tech-

N o rhodium w a s reduced when

Evaluation of Ef for t

Process Development expenditures for t h e f i s c a l year t o date are 37 percent of the FY 1970 budget f o r Product F l ex ib i l i t y . The m i n k a l e f for t on t h i s program w a s consis tent with t h e l e v e l of funds al located.

The oralloy program i s pro-ceeding as planned- being redirected t o include use of plutonium as a f u e l as w e l l as

However, t he program i s

oralloy 0

The Pu-238 program has been delayed by t h e unexpected rupture of the neptunium oxide-graphite matrix elements. p a r t i a l l y i r rad ia ted elements ind ica tes higher than expected (0.28 ppz) of plutonium-236, but new analyses are being made with more pre- c i s e instruments. l u s of t a rge t material.

Original analysis of the

Work has j u s t s t a r t e d on elements using a t h i n anm-

RL 6-01 Page 25

To date no problems have arisen from irradiation of a Haynes Alloy-25 (51 percent cobalt) tube in a tube channel.

Irradiation of 13 columns of plutonium-aluminum alloy is proceeding according to plan with first discharge planned for the fourth quar- ter FY 19709

One kilogram of neptunium was converted to the oxide in a partially successful continuous calciner test unit. No irradiated neptunium was scheduled for separation duriug the report period.

Difficulties in the partition of plutonium-238 from neptunium will require fbther flowsheet development.

F

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ETIVIROMMENTAL AND REGULATORY TECHNOLOGY PROGRtZM

The Emironmental and Regulatory Technology Program contains the following research and development items: Nuclear Safety, Columbia River Studies, and Radionuclides in Soils. The activities under these items are conducted to assure that the Hanford production reactors and separation plants are maintained and operated safely, to minimize the heat and isotope loading of the Columbia River, and to determine those parameters by means of which the consequences of accidental releases ( including those caused by earthquakes) of radioactive and chemically toxic materials to the atmosphere or groundwater can be predicted.

Objectives and Incentives

Effort is under way to model the consequences of reactor loss of cooling accidents including temperature transients, reactions in the overheated core, and fission-product release from overheated fuel. Development of this technology will permit a quantitative assessment of the basic risk of facility operation. Methods for controlling the fission gas released following reactor accidents are being developed. This will reduce the most acute consequences of reactor accidents. The characteristics of aersol formation from nuclear materials are being determined in support of shipping hazards analyses. A transport model is being developed to predict the rate and direction of travel of fission-product materials from accidentid release to the soil.

Progress W i n g Report Period - Reactors Regulatory Technology

General.

In-place iodine removal tests of charcoal filters at the K and N reactors have been completed. New Process Standards requiring semiannual charcoal sample tests from each reactor cell and annual in-place tests of all reactor cells have been approved. oratory facility for testing charcoal samples has been reassembled in the 190-H building water laboratory and used extensively.

The lab-

RL 6-01 Page 27

B Reactor

Experimental apparatus t o simulate loss-of-coolant incidents has been constructed by t h e thermal hydraulics laborator ies of Battelle-Northwest, and tests are i n progress. Additional work w a s car r ied out t o develop ana ly t i ca l t o o l s f o r coordinated cal- culations of primary loop blowdown t r ans i en t s , confiner tran- s i en t s , and f i s s i o n product t ranspor t . Building 324-D and 292-T f a c i l i t i e s have been used f o r m e t a l - w a t e r react ion tests of irra- diated N Reactor f u e l pieces . A major e f f o r t was made t o ver i fy t h e travel t i m e of radioiodine from the 1301-N c r i b t o the r iver- bank szrings. A minimum t i m e o f f i v e days seems most l i ke ly . A qual i ta t ive study w a s performed t o co r re l a t e I? Reactor fue l element failures with increased iodine-131 a c t i v i t y i n spring water. Mete- orological data acquis i t ion has been continued using the automated N Met System, and software f o r handling data from the system is about 90 percent completed. E s t i m a t e s have been obtained of t he maximum rainout of noble gases by fog sprays i n the confinement zones, and of the amount of f i s s i o n gases t o be expected i n a loss- of-coolant accident.

K Reactors

Evaluation of t h e seismic adequacy of c r i t i c a l K reactor safety systems w a s started by John Blume and Associates.

The heats of react ion f o r t he various uranium-aluminum compounds i n the standard state were obtained.

Muclear Safety Assurance

An in-depth analysis of t h e Operating Safety L i m i t s f o r both the K reactors and N Reactor w a s i n i t i a t e d , A program was s t a r t ed t o pro- vide an overa l l review, cor re la t ion and updating of t he Safety Analysis Reports f o r the K and N reac tors , several areas toward upgrading t h e technical bases f o r both t h e Safety Analysis Reports and the Process Standards. Analytical pro- grams w e r e car r ied out t o enable development of technical bases f o r process limits fo r production loading of Mark I V fuel . Parametric re la t ionships a f fec t ing trade-offs i n K reac tor u t i l i z a t i o n of t he high speed scanner f o r nuclear s a fe ty protect ion were developed and published. I n i t i a l t echnica l bases f o r K reactor tube power l i m i t s f o r a na tura l uranium load were developed.

Progress w a s made i n

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Columbia River

Routine monitoring of t h e r ive r continued, as did r ive r model s tud ies at Washington S t a t e University where the primary emphasis was placed on cor re la t ing r i v e r model results with infrared imaging data. Lab- oratory s tudies of e f f luent f locculat ion indicated t h a t t h i s i s a po ten t i a l method of cleaning up radioact ivi ty . Planning f o r p i l o t p lan t tests i s under way, and development of a preliminary concept of a Azll plant operation has been s ta r ted .

Progress During Report Period - Chemical Separations

Nuclear Safety

Atmospheric Dispersion of Radioactive Pa r t i c l e s

Laboratory s tudies , t o characterize plutonium aerosols produced during combustion, were scaled up t o inves t iga te release of aerosols from large-scale burning incidents w a s completed, and renovation and modification of an unused building (242-B) su i t ab le fo r hous- ing t h i s f a c i l i t y w a s s ta r ted . To reduce t h e hazard of alpha con- tamination from these large-scale tes ts , uranium w a s se lected as a su i t ab le stand-in f o r p1utoni.m campounds

Seismic Invest igat ions

The frequency, i n t ens i ty , and s p a t i a l d i s t r ibu t ion of microearth- quakes occuring i n t h e region about Hanford are being determined by the U. S o Geological Survey National Earthquake Research Center with an array of highly sens i t ive geophones. U, S. Geological Survey, together with a de ta i led f i e l d study of possible young f au l t i ng i n the region, is being sponsored j o i n t l y by the Division of Reactor Development and Technology and the Division of Production. Forty of t h e 107 microearthquakes, de- t ec t ed during a six-month period, were s i tua ted near t h e center of t h e Pasco Basin where %he geologic s t ruc ture of t he bedrock is con- cealed by more than 1100 feet of sedimentary rock. microearthquakes detected along the a n t i c l i n a l r idges on the Banford Reservation where faults have been inferred. In support of t he U. S. Geological Survey invest igat ion of young f au l t i ng on Gable Mountain and Gable Butte, approximately 2000 feet of trenching w a s excavated. As many as four d i f fe ren t hypotheses w e r e advanced t o explain t h e geological s t ruc tures exposed i n the trenches.

This work by t h e

There were no

Radionuclides i n Soi l s

Movement of Groundwater

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A general solution method w a s developed f o r predicting the r a t e and d i rec t ion of water flow through porous media. method, a number of computer rout ines have been writ ten t o solve groundwater flow problems. The computer routine, developed f o r solving flow problems i n t he vadose zone beneath waste disposal si tes, was extended during t h i s report period t o the en t i r e satu- ra ted zone of the groundwater flow system beneath the Hanford Reservation. Error and s e n s i t i v i t y analyses a re being made on a sorption model which relates t h e r e l a t i v e rate of movement of con- taminants t o t h a t of groundwater. based on t h i s sorption model, which provides a l imited capabi l i ty f o r predicting the r a t e and d i rec t ion of strontium-90 and cesium-137 through the vadose zone underlying a w a s t e disposal s i te and i n t h e saturated portion of the groundwater flow system.

Using t h i s solution

A t ransport model was developed,

Long-Term Storage of Nuclear Waste

Rock and water samples of t h e geological formations underlying the Hanford Reservation have been co l lec ted t o a depth of 5661 f ee t . Data obtained from these s m p l e s and from hydrological t e s t s and geophysical logs made i n the borehole indicate t h a t t h e basal t flows are e f fec t ive ba r r i e r s t o v e r t i c a l groundwater movement; i n some zones the permeability of t h e basa l t s was so low t h a t measure- ments could not be made by standard hydrological t e s t ing techniques.

S o i l mositure poten t ia l s of 20 t o 100 negative atmospheres have been observed i n t h e vadose zone of t he sand-sil t region south of the 200-East area. These desiccated zones range from 20 to 30 feet i n thickness and occur 15 f e e t t o 45 f e e t below grade. bases of these data , it has been determined t h a t incident precip- i t a t i o n i n t h i s region - cannot percolate t o the underlying water tab le .

On the

Evaluation of Effor t

Process Development expenditures f o r the f i s c a l year t o date are 53 percent of the FY 1970 budget f o r Environmental and Regulatory Technology.

RL 6-01 Page 30

Evaluation testing of the reactor confinement system charcoal filter for removal of iodine from exhaust air was completed. formed very satisfactorily throughout the testing.

The filter per-

Work under the N fie1 heatup transient experimental program is prog- ressing on schedule.

Resolution of the groundwater travel time between the 100-N crib and the river has removed a major barrier to the completion of studies of consequences of accidents involving fission-product transport by water leaving the reactor and reactor building.

Experimental measurement of metal-water (irradiated N fuels) reactions have progressed very satisfactorily within the past six months. Mea- sured hydrogen releases are in close agreement with analytical models. Tests are continuing at temperatures 50' to l0O'C below the uranium melting temperature.

Work under the meteorological studies program is progressing satisfac- torily, except that diffusion test planning is undergoing review be- cause of the high costs.

Rapid progress in the K reactor seismic evaluations will provide timely input to the upgrading of safety studies.

An in-depth review of the technical bases for nuclear safety criteria and of the administrative system to assure the continued safe operation of the reactors have been prompted by curtailment of the technical sup- port effort at a time of evolving nuclear safety philosophy and corre- sponding advances in technology. An aggressive program is being continued to assure safety of the reactors.

Emphasis has been placed on investigation of flocculation of K reactor effluent water as a method for removal of radioactivity. flocculation test is being prepared. Manpower and funds are Limiting.

A one-tube

Extensive measurements were made of the geological, hydrological, geo- chemical, and seismic parameters which influence the rate and direction of radioactive materials displacement in the environment. The knowl- edge gained has led to the development of predictive capabilities for determining the consequences of accidental release of radioactive roaterials.

RL 6-01 Page 31

Data obtained from samples, hydrological tests and geophysical logs made i n t h e deep borehole a t Hanford, ARH-DC-1, ind ica te t h a t t he b a s a l t flows a re e f f ec t ive b a r r i e r s t o v e r t i c a l groundwater movement. Certain of the basa l t flows have natural r ad ioac t iv i ty leve ls t ha t appear t o be cha rac t e r i s t i c and which may prove usefu l i n ident i fying and co r re l a t ing s t ra t igraphy through the use of na tura l gamma logs.

Samples co l lec ted from t h e Rocky F l a t s f i re a r e being studied t o char- ac t e r i ze the par t ic le -s ize d i s t r i b u t i o n and nature of plutonium aero- sols t o a i d i n evaluating the consequences of accidents.

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W A S T E MANAGEMENT PROGRAM

The research and development work under this program will provide technology whereby the wastes (gaseous, liquid, solid) generated in the chemical processing of irradiated fuels and targets may, at low cost, be solidified and confined for long-term surveillance and con- trol, or safely dispersed in the environment.

Objectives and Incentives

The purpose of this activity is to provide long-term high-integrity confinement and control of hazardous nuclear waste and to reduce the amount of radioactivity released to the environment to the lowest practical value.

The incentives for solidifying the bulk salt waste as opposed to con- tinued in-taak storage are both safety (immobilization of fission products) and economics (estimated savings have a present worth of $60 million as compared with continued storage in tanks having a 20-year life).

The incentive for improved treatment of low-level wastes is safety in keeping with the AEC anti-pollution campaign aimed at reducing signif- icantly- the mounts of radioactivity released to the environment.

The incentive for removing the long-lived heat emitters for separate high-integrity storage is near-term safety ting isotopes, strontium-90 and cesium-137, are to be removed by sol- vent extraction and ion exchange, respectively, and packaged in high-integrity containers. ified in underground storage tanks in a few years instead of in several decades.

Progress During Report Period - Reactors

The long-lived heat emit-

"he residual bulk wastes will be solid-

-.

Reactor Waste Management

A plan has been developed to establish requirements and methods for ultimate return of the reactor operating areas to the original state and to define safeguard requirements consistent with these methods. Also involved are the definitions of special facilities and systems, removal techniques, disposal sites and transportation methods,

RL 6-01 Page 33

development of costs and schedules and radiological mapping, Preliminary concepts for disposing of N Reactor waste products have been sufficiently completed to submit a budget data sheet outlining requirements for the necessary facility.

Progress During Report Period - Chemical Separations Low- and Intermediate-Level Waste

Evaluation of a prototype monitoring system on the Purex Plant cooling water line continued. ating satisfactorily. In laboratory tests, a plastic scintillator was 2 to 5 times more sensitive to beta radiation than present detectors e

The monitoring system has been oper-

Methods for fabricating sample cells of mar (trade name--Perm Walt Corporation) with cast-in-place plumbing adapters are being devel- oped. In early models, the Kynar slabs were not adequately fgsed, and the cell split in half when the adapters were machined out of the cell.

A reverse osmosis test unit successfully removed salts and radio- active materials, but not tributylphosphate, from Purex process condensates.

Tributylphosphate and normal paraffin hydrocarbon were found in a Purex process condensate stream (1UD) at concentrations of 1.15 and 1.97 @;/gal, respectively. present. Plant tests indicate that at these concentrations 1UD can be used as process water in the uranium extraction column without affecting plant performance.

Other unidentified acidic material is also

A column of AW-500 zeolite (trade name--Union Carbide Corporation) was successflilly used in the laboratory to remove cesium-137 from 241-A and AX tank farm gondensates. ceeded 100; cesium-137 concentrations in the effluent were less than 0.1 of the maximum permissible concentration for water (MPCw)

Preliminary tests by Idaho Nuclear Corporation personnel indicate that a fluidized bed unit can be used to immobilize 2 Plant waste organics while calcining 2 Plant high salt aqueous waste. Purex and B Plant solvents may also be suitable for burning in a fluidized bed unit.

Decontamination factors ex-

RL 6-01 Page 34

Organic materials present i n 2 Plant high salt waste can be e f f ec t ive ly removed by sorp t ion on act ivated carbon, s i l i c a g e l and alumina-type calcine. Regeneration of the carbon bed using acetone o r l o w temperature steam w a s only pa r t i ca l ly successful.

S a l t Wrzste Processing

Molecular weights of ai-( 2-ethylhexyl) phosphoric acid (HDEXP) were determined as a funct ion of i r r a d i a t i o n conditions. Some polymer- i za t ion of HDEPH occurred under all conditions tes ted. i za t ion w a s h ighest f o r HDEPH i r r ad ia t ed while i n intimate contact with 1 M n i t r i c acid. I n i t i a l scouting t e s t s were successful i n removing KDEPH from i r r a d i a t e d B Plant solvent by using a weak base macroreticular ion exchange r e s i n (Amberylst , t rade name--Rohm & Haas Company), phosphate and normal pa ra f f in hydrocarbon t o prepare B Plant sol- vent, demonstrated strontium ext rac t ion capabi l i ty comparable t o uni r rad ia ted solvent.

The polymer-

The recovered HDEPH, when mixed with f resh t r ibu ty l -

Testing of var ious f loccu la t ing agents to increase the s e t t l i n g r a t e of water-leached sludges continued. Dow Chemical Company) w a s successful i n increasing s e t t l i n g r a t e s of water-leached sludges i n tanks 104-A and 106-~.

Only Purif loc A-21 ( t r ade name--

The Hanford s l u i c e r has been used t o successfully remove sludges from underground s torage tanks 101-A and 104-A- a corrosion i n h i b i t o r (Rhodine A) was e f f ec t ive in softening and/or dissolving a hard sludge l a y e r encountered during removal of sludge from tank 105-A. ac id i n t h e center of t h e tank, t h e sludge i n the t e s t aFea w a s removed successf'ully by s lu ic ing .

Sulfuric acid containing

A f t e r spraying about 750 gallons of 1 M su l fu r i c

Modifications of t h e €3 Plant pH measurement systems have permitted considerable improvement i n pH measurement accuracy and r e l i a b i l i t y .

An e lec t ron ic u n i t capable of measuring n i t r i c acid concentration i n the Purex Plan t ex t rac t ion column waste stream (HAW) w a s developed and i s ready f o r p lan t t e s t i n g .

Use of cathodic pro tec t ion t o reduce corrosion of waste tanks has shown encouraging r e s u l t s . A laboratory-scale tank protected at 10 milliamperes per square foot has not f a i l e d a f t e r 49 weeks' contact with simulated bo i l ing Purex a lka l ine waste. and a protected tank (1 mil l impere/square foot ) f a i l e d a f t e r 5 and

An unprotected tank

8 weeks of exposure, respect ively.

RL 6-01 Page 35

Recovery of strontium from current ac id waste (CAW) so l ids was attempted i n t h e laboratory by contacting adjusted CAW, which con- t a ined the usual chemicals added i n preparing B Plant solvent ex- t r a c t i o n feed, w i t h B Plant solvent. t i u m was removed from the so l id s . Plant t e s t s have shown that t h e caustic-carbonate metathesis process f o r removing strontium from CAW s o l i d s is preferred t o t he N-hydroxyethylethylenediaminetriacetic ac id leach process.

Only 24 percent of the stron-

Laboratory r e s u l t s indicated t h a t rare ear ths (RE) may be subs t i tu ted f o r lead as a strontium c a r r i e r i n t h e sulfate precipi ta t ion flow- sheet , w a s precipita+,ed at a pH of 0.5 t o 1.0 by using 1 - M su l f a t e and

About 95 percent of t h e strontium i n Purex ac id i f ied sludge

0.01 - M RE* Early laboratory data show t h a t strontium can be pur i f ied using a z e o l i t e 13X (trade name--Union Carbide Corporation) ion exchanger when metal ions i n the feed a r e complexed w i t h sodium n i t r i l o t r i - ace t a t e . The process appears t o be a t t r a c t i v e as an a l t e rna t ive t o pu r i f i ca t ion of B Plant strontium product by second solvent extrac- t i o n cycle and magnesium hydroxide p rec ip i t a t ion steps.

Waste So l id i f i ca t ion i n Tanks

Thermal conductivity of Redox sludge from tank 108-s~ was measured. The value obtained [O.llO Btu/hr n2 (OF/f t ) ] agreed reasonably w e l l with synthe t ic Redox sludge L0.121 Btu/hr f t 2 ( O F / f t ) ] .

Water leaches of tank 108-SX Redox sludge removed 2.2, 2.6 and 19.2 percent , respect ively, of the plutonium, strontium-90 and cesium-137 a f t e r 2 1 leaches i n which f r e s h water o r 0,1 sodium n i t r a t e solu- t i o n w a s vigorously mixed w i t h t h e sludge during each leach cycle. Leachates were passed through a s o i l column t o demonstrate adsorption of t h e leached radionuclides by the soil. column measurements were made with a salt cake sample from tank 116-y~ ( t h e in-tank s o l i d i f i c a t i o n u n i t ) more readf ly dissolved from t h i s sludge sample, and loading on soil was less e f f i c i e n t ,

Similar leaching and s o i l

Plutonium and cesium were

The e f f e c t of s o i l moisture on thermal conductivity of t he s o i l w a s determined i n laboratory experiments. The one-dimensional computer program for calculat ing temperature gradients i n underground waste ta&s and surrounding s o i l w a s rewr i t ten and debugged a f t e r e r ro r s were found,

RL 6-01 Page 36

Waste Packaging

The cesium chloride and strontium fluoride flowsheets were verified in the laboratory by using both simulated and radioactive plant feeds. Test capsules of nonradioactive cesium chloride a.nd strontium fluo- ride were prepared for container compatibility studies, Based upon initial compatibility results ( 3 months) with nonradioactive cesium chloride, HasteUoy C-276, was equal to or better than Hastelloy C. Preliminary results indicated that 316~ stainless steel may be a good alternative to Hastelloy C for csntaining strontium fluoride.

Immersion of the capsules in boiling water was found to be effective in removing most of the radioactive contamination from the metal surfaces

Pilot-scale equipment for conversion of nonradioactive strontium nitrate to strontium fluoride followed by filtering and sintering of the strontium fluoride was successfully used to demonstrate the var- ious process steps. The impact consolidation protctype equipment was used to encapsulate nonradioactive strontium fluoride to about 85 percent of theoretical density. A thermosiphon evaporator and downdraft condenser were used to concentrate the nonradioactive cesium-ammonium carbonate product from the zeolite ion exchange purification process. was acidified with hydrochloric acid and evaporated to dryness. The nonradioactive cesium chloride formed was then melted and trans- ferred by vacuum into a series of stinless steel containers. Studies indicated that relatively low amounts of cesium are volatilized dur- ing the evaporation, drying and melting steps; and that conventional. off-gas equipment can trap the volatilized cesium.

A computer program was written to determine steady-state temperature profiles in doubly encapsulated sources when stored either singly or in various arrays.

Six Hastelloy C capsules containing nonradioactive strontium fluoride were welded with a Gas-Tungsten-Arc welding system. observed with helium leak tests, and the capsules are undergoing nondestructive testing. sules for containment of cesium chloride are under way. The ability to measure weld penetration ultrasonically was established with developental weld samples,

The resulting cesium carbonate concentrate

No leaks were

Welding tests on 3 1 6 ~ stainless steel cap-

RL 6-01 Page 37

A cell mockup of B cell was constructed as presently designed for the Strontium and Cesium Packaging Facility, arrangements were tried to obtain assured remotability and opera- bility of a l l equipment.

Various equipment

Salt Waste Volume Reduction

Miniature centrifugal contactors were fabricated to test the flow- sheet for reducing plutonium( IV) to plutonPum(II1) with hydrazine- stabilized uranium( IV) reductant. demonstrate the use of uranium(1V) as a replacement for the reductant ferrous sulfamate are being completed.

Preparations for a plant test to

Further laboratory tests show that uranim(V1) and plutonium(1V) can be reduced with hydrazine catalyzed by platinum. lated 2DF feed (feed to the second uranium cycle uranium-plutonium column) and hydrazine, when passed through a bed of platinum pellets, yielded uranium( IV) concentrations proportional to the hydrazine con- tent up to about 30 percent 7 ghydrazine nitrate and 70 percent of 2 DF feed. of uranium(1V) formed which agreed with theory. More hydrazine was consumed than theory would predict.

Mixtures of shu-

Acid consumption vas determined to be 1.5 moles per mole

Use of ultraviolet light to stimulate reduction of plutonim(1V) to plutonium( 111 ) by tributylphosphfte was concluded to be unsuitable for plant operation. Organic soluble nitrite suppressors were tested; phenylhydrazine and phenylurea were most effective in destroying nitrite ion,

Macroreticuiar ion exchange resins were scouted for possible use in treatmer;t of &rex solvent. Amberlyst 15 (trade name--Rohm & Haas Company) successfully removed fission products (zirconium, niobium and ruthenium) and the prin- cipal solvent radiation degradation product dibutylphosphate (DBP ? from used -ex solvent.

A mixture of Amberlyst A-21 and

- A modified falling-film evaporator was successfully used to concen- trate Z Plant high salt aqueous wastes by factors of 1.3 t o 2,4.

Plutonium can be recovered from 2 Plant sump wastes by Dowex-50 resin (trade name--Dow Chemical Company) withou% reducing plutonium to the trivalent stateo Some unidentified cation was loaded on the resin with the plutonium and restricted the resin capacity for plutonium.

RL 6-01 Page 38

Alterna+,f-se Sol id i f ica t ion Methods.

An engineering study t o evaluate several a l ternat ive processes f o r sol idifying Purex high heating and low heating wastes has been s t a r t ed ,

Tests conducted by Idaho Nuclear Corporation personnel have shown t h a t Z Plant high sal t aqueous wastes can be calcined i n an in-bed combustion, f lu id ized bed uni t . A t e s t rim using ac id i f ied Purex coating waste w a s unsuccessful as the high sodium content caused p a r t i c l e agglomeration and lumping,

Evaluation of Ef for t

Process Developrent expenditures f o r the f i s c a l year t o date are 43 percent of t h e FY 1970 budget fo r Waste Management, The minor program f o r reactor waste management has been limited, t o date, t o def in i t ion of t he requirements i n the reactor area.

Increased e f f o r t has been placed upon developing the technology f o r reducing armual quant i t ies of radionuclides released t o the environ- ment i n low- m d intermediate-level l iquid castes t o l e s s than 400 b e t a cur ies and less than 10 grams of plutonium. Progress i s satfs- fac tory towaxd developing sensi t ive radiation cletection instruments f o r use i n diver t ing contaminated wastes t o tanks fo r l a t e r treatment, toward f inding economical methods f o r recycling o r removing radio- nuclides from intermediate-level l iquid wastes , md toward treatment of organic gaseous and s o l i d wastes for removing radionuclides and noxious chemicals. The development schedde should allow f a c i l i t i e s t o be i n s t a l l e d t o a t t a i n the Eibove release quantit ies by the end of FY 1975-

The salt w a s t e processing and waste sol idif icat ion i n tanks sub- programs have been reduced i n level of e f fo r t since B Plznt and the in-tank so l id i f i ca t ion un-$ts are now opera thg successfully centered xpon improving flowsheet and equipment operation and i n de- veloping a l t e rna t ive flowsheets f o r increasing safety a d reducing cos t s 0

Work has

A major development program for definfng and demonstrating methods '10 encapsulate fission-product strontium and cesfw~ has been progressing s a t i s f a c t o r i l y i n support of a plant s ta r tap i n FY 1974.

RL 6-01 .Page 39

E f f o r t s t o reduce t h e salt content of Purex xastes have increased because of t he progected waste management cost savings and concomitant improved strontium recovery from current acid wastes. velopment work has been completed, and uranium(1V) w i l l be tes ted i n t h e plant as a replacement f o r ferrous sulfamate.

Laboratory de-

Reduction of Z Plant aqueous waste t o a sol id calcine has been demon- s t r a t e d i n a f luidized bed system.

RL 6-01 Page 40

REFERENCES

I,

2.

3 .

Douglas United Nuclear, Inc. , Semiannud Research ani! Development 11

Report", by J. W. Riches, December 31, 1969, DUN-6548, RL 6-21.

Atlant ic Richfield Hanford Company, Semiannual Report, Process If

Development , May 1 , 1969 through October 31, 1969", by R . E. Isaacson and R. E. Tomlinson, DeceEber 31, 1969, ARH-1515, RL 6-11.

"Richland 02 R&D Program, Annual Report Summasy", by Richland S t a f f , June 30, 1969, RL 5-0,

RL 6-01 Page 4 1

DISTRIBUTION

RICHLAND OPERATIONS OFFICE

1. D. G. W i l l i a m s 2. H. E. Parker 3. 0. J. Elger t 4. W. Devine, Jr.

5.-8. M. R. Schneller 9. C. L. Robinson

10 . w. E. Lo t2

11. I. D. H a r t m a n 12. R . L. Plum 13. M. H. Arndt 14. E. B. Jackson 15. J. T. Christy 16, 0. W. Rathbun 17. Record Copy

ATLANTIC RICHFIELD HANFORD COMPANY

18. S . J. Beard 19. D. J. Brown/G. L. Ritter 20. B. M. Dobbs/F. W. Gates 21. R. P. Corlew/J. H. Warren 22. J. B. Fecht 23. B. B. Field 24. M. K. Harmon 25. W. M. Harty 26. 0. F. Hill 27. R. E. Isaacson

28 0

29 0

30 4

31 * 32 33 34 4 35 36 e 37 0

BATTELLE-NORTHWEST

R . W. McCullugh G. C , Oberg L. M. Richards H. P. Shaw A. E. Smith/H. H. Hopkins P. W. Smith W. G. Smith M. J. Szul inski R . E. Tomlinson ARHCO Fi l e s

38. P. W. Albaugh 39. E. L. Alpen 40. J. M. Batch 41. A. L. Bement 42. C. A. Bennett -. 43. R. E. B u r n s 44. E. D. Clayton 45. F. G. Dawson 46. r), R. deHalas 47. E. A. Eschbach

48. R. F. Foster 49. B. M. Johnson, Jr. 50. J. 1. Judy 51. E. M. Parker 52. A. M. P l a t t 53. W. D. Richmond 54. C . A. Rohrmann 55. B. Wolfe 56, C , J. Touhill/M. H. Karr 57. BPJW F i l e s

RL 6-01 Page 42

J

58, G. O*, 59. P o A. 60. G. C. 61. D. H. 62. R . E. 63. G. C. 64. R. G. 65. R. W.

67. H. W.

69. R, T .

66. c. D.

68. c. w.

DOUGLAS UNITED NUCLEAR, I N C .

Amy C a r l s o n Coleman Curtiss Dunn Fullmer Geier Hal le t , Jr . Harrington He ac o c k KUhlmEUl Jessen

J. M. Lutton W. M. Mathis W. S. Nechodom D. W. Peacock R. W. Reid J. W. Riches 0. C. Schroeder R . H , Shoemaker J. T, Str inger W. K. Woods DUN Fi l e s

O F F S I T E

81.-85. F. P. Baranowski, AEC-HQ 86. W. L. Ginkel, AEC-ID 87. S. R . Sapi r ie , AEC-OR 88. I?. Stetson, AEC-SR