.-

Cover Sheet for a Hanford H istorkal Document Released for Public Availability

Released 1994

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

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

DISCLAIMER

This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, make any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof.

DISCLAIMER

Portions of this document may be illegible in electronic image products. images are produced from the best available original document.

THIS M A T E R I A L C O N T A I N S 1NFORMATlON A F F E C T I N G

T H E NATIONAL D E F E N S E OF THE U N I T E D S T A T E S

W I T H I N THE MEANING O F THE E S P I O N A G E L A W S ,

T I T L E 18, U . S. C . , S E C S . 793 A N D 794, T H E T R A N S -

M I S S P O N OR R E V E L A T I O N O F WHICH IN A N Y MANNER

TO A N UNAUTHORIZED P E R S O N I S P R O H I B I T E D BY

T. W. Evans

I?. K, Kratzer and

MAR2 1952 R ~ I U K I \ I 18

1

I

LAW.

I I I I

I I I I

I I

54-3000-340 0 5 7 ) AEC-GE RIC*CAND. I (CLASSIFICAT-ION)

D E ~ l ASSIFIEU DISTRIBWIOM

l., Fw Albaugh 2, GS Allison 3. TW Ambrose 4. Go Amy 5. JA Ayres 6, LV Barker T o Fs B e l l 8, Bolliger 9. JH Brown 10, SH Bush 11, JJ Cadwell 12" RTJ Call 13. AC Callen 34, PA Carbon 15, MA Clinton 1.6, RL Dickeman 17. DR Dickinson 18, EL Dillon 19. RV Dulin 20. '1Iw Evans 21. EAmms 22, WS Frank 23, Gcz Fublmer 24. w14 Gilbert 25. SM Gill 26, JW Goffard 27'. SM Graves 28, OE Greager 29, AB Greninger 30, RO Gunprecht 31. 6TL Guthrie 32, AK Hardfn 33. EN Heck

340 35. 369 3 7. 38. 3 9. 40. 41, 42. 43 0

44, 45 0

46, 470 48, 49, 502 5;. 5 20 53 0

540 55 0

56. 5 7. 58, 5 9" 60. 61. 62, 63 0

64, 65 0

PC Jerman €?I Jessen E3 Kemper Fax Kratzer I& Kusler GA L a s t CG L e w i s 6;8 Mansius JE Minor MR Moon SL nelson

R Beidner JW Nickolaus

R Milson Kw NGIWOOd

RW Reid JW Riches OC Schroeder RPJ Shields HG Spencer DR Stenquist

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J!E Stringe J C Tobin

FW Van Wormer JW Weber RG Wheele D6 Worlto FE Young IFF Z u h r 300 Fi les Record Copy

Thfs document c lass i f ied by

Thfs document consists of

7

February 14-, 1962

KER LOOP FUEL 'EESTIMG PROGW AND SCHEDULF: - cy4962

INTRODUCTION

The in te res t s of s e v e r d departments at Hanford are involved in the planning, execution and evaluation of the results of the KER loop tes t ing e f fo r t i n support of the NPR f u e l program, ing groups m u s t be well-integrated i f effective use of our l imited tes t ing

The varied in te res t s and ac t iv i t i e s of the participat-

HW-72680 Page 2

capabili ty is t o be made. integration by surmnarizfng the current thinking on the goals of the MPR f u e l tes t ing program and by presenting the current loop schedule,

The purpose of t h i s report i s t o help achieve t h i s

I. Testing Objectives

The NPR first load fue l design i s fim, the fabrication process is essent ia l ly established, and the fuels plant is comyl2te m d production has begun. Ideal- i s t i ca l ly , our greatest need at present i s f o r knowledge of the behavior of the product of the fuels plant under anticipated NPR operating conditions, i n quantit ies great enough t o have s t a t i s t i c a l significance, Because of the low capacity of the loops, however, s t a t i s t i c a l performance data cannot be obtained u n t i l after N reactor, i t s e l f , begins operation. The loop tes t ing e f f o r t w i l l only be able t o turn up design, material o r process deficiencies which are serious enough t o be detected by irradiat.ion of s m a l l quantit ies of test elements, these needs, are discussed below,

Realizable program needs and the scope of t e s t designs t o settfsfy

A, Characterization of "First Load" Behavior -- The first load f u e l design w a s f ixed before any appreciable, direct 'ly pertinent t e s t ing had been accomplished, The greatest need has been, and w i l l continue t o be u n t i l reactor startup, f o r t e s t s which w i n . provide assurance that the first load w i l l be reasonabley adequate o r tha t w i l l indicate where changes must be made t o assure adequacy, s m a l l capacity of the loops imposes serious limits on our a b i l i t y t80 do t h i s and makes it mandatory that considerable care be given t o the planning and interpretation of the t e s t s . characterization of the behavior of the first. load follows.

The

A l ist of those elements t ha t con.stitute

1. Metallurgical Behavior

The behavior of uranium in tubular geometry, irradiated at. temperatures from 200-456C t o exposures somewhat i n excess of 2000 W I T , i s not well known but is f e l t t o be the most important single fac tor i n determining the behavior of the I P R f u e l element, Information is needed on the magnitude of swelling as a function of temperature and exposure, the dimensional changes tha t occur t o accommodate swelling, the extent t o which w a r p , growth and bumping OCCUT, the nature and severi ty of the cracking of the uranium and the tendency f o r cracks t o propagate in to and thpough the claddhg,

The behavior of the cladding i s of concern from the points of view of corrosion, hydrogen pickup, crud &positdon and mechanical be- havior under creep loading conditions w i t h strafn-cycling and i r radiat ion damage superimposed,

In addition, the brazed closure process resu l t s i n phase transforma- t ions i n both the Uranium and the Zircaloy-2 and i n thickening of the clad-to-fuel bond. The e f fec t of the associated metallurgical changeg on the f u e l element performance should be evaluated, The thermal incompatibility of uranium and zirconium may lead t o fa i lure of the cap-to-fuel bond Which, if it occurs, may result i n cladding failure at the cap-fuel interface, Evidence bearing on the l i k e l i - hood of t h i s event is required,

2, Design Testing

In addition t o the concern over the iniiependent mat.eri&l prope:rties discussed above, the gross, over-all fuel element behavior needs t o be determined, The f u e l element cons1st.s of fuel, cladding, c:Losupes, spot-welded supports of two types, one of which has iron covers (shoes), and a f ixing device, needs a considera5le amount of ;-oof~~testlng, f u e l dimensions have been picked t o sat isfy, o-eher things, heat s p l i t require- ments, Measuremnt. of water channel temperatures w i l l be required t o determine whether the heat s p l i t s are within the allowable :range. Trouble is not expected t o arise from any ~f these problem, purpose of the tes t ing effcrt i s t o demmttrate whether t h i s confidence is just i f ied. It is presumed that if" the t e s t s indicate the need f o r design o r process changes tha t the changes required w i l l be of a minor nature s o that they m y be made without appreciable disruption of the fuels plant. o r process,

!The behavior of the whole assembly

The

30 Fuels Plant Quakity Monitoring

If the materials of" the fuel element behave web& and i f the design - of the element i s found zo be adeq-ate them is s t i l l the very important pract ical problem of mass-producing the element i n a f a i r l y economical manner. While economy is not an important consi&ration f o r the first load, every inspection step requires a quantitative c r i te r ion of acceptabili ty which, t o a Large extent now, 18 determined by judgment based on on.4ly l imited data, While it is t rue that t h i s will a:bways be the case, t o sorie degree, the t e s t ing e f f o r t should continuczlly provide more information on which t o base these jtidgments, although the s t a t i s t i c a 2 aspect w i E be missing,

4. surface C o n t a n a t i o n

Rupture monitor reliabl. l i t :y i s important In the MPR because of the possible consequences of fuel. element fa i lure . Early, posit ive dei;ection of failures appears now t o be essential . sources.of gama. ac t iv i ty i n the primary loop of the reactor w i l l reduce the sens i t i v i ty of" the rupture monitors, and may a l s o increase personnel exposures signif fcantly, Contm.ination of the fue l element surface i s possible and work on instrumentation t o detect surface contamination i s i n progress, Quantitative i n f o m t l o n i s neecied on a l l aspects of" t h i s potential. problem - how much contamination is probable, how much can be detected with the existing instrwneni;ation and how much does it take t o interfere with the detection of fa i lures .

Extraneous

5. Restrain% of Warp

Because of the relat ive slenderness of the two fuel components,, the inner tube is f e l t t o be the m o r e waq-prone, inside the outer tube by supports, the crushing strength of which i s adJustable t o some extent by design, tha t w i l l a r i s e should the inner tubes warp shodd be determined by the extent t o which the supports and the outer tube are capable of restraining warp.

lt i s held i n posit ion

The magnitude of the problems

Information on t h i s point is required.

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6* Frequency and Severity of Warp

The need f o r clarifying the consequences of warp of the inner tube was stated above. t o occur. mined w i t h confidence only a f t e r reactor s ta r tup when significant

’

numbers of f u e l elements are irradiated. A considerable effor t , has been expended on study of f u e l :ALpiF--ctes l i k e l y t o cause w a r p during fabrication. Determination GP the re lat ion (if any) between w a r p tendencies d u r h g fabrication and warp 6uring i r radiat ion is needed,

In addition, we need t o know the amount of warp l i ke ly This falls in to the category of things that can be d.eter-

E.

7. Cladding and Closure Corrosion

%e corroslon ra tes of Zr-2 and 219-5 Be have been studied extensively and are believed t o be more t h m adequate f o r the anticipated service conditions, months, the main problems haw been assumed t o be w i t h the central elements i n a. process tube and the tp-tine conditions adfusted accordingly, ou t le t temperature*

under conditions appropriate f o r corrosion tes t ing.

Although prototy-pe t e s t ing has been i n progress f o r s ix

This requires coolanr terqexatures some 50°C below the NPR Corrosion i s expected t c be most rapid i n the

&wnstream. elements and prototype tests have not yet been conducted

8. Fuel Element inz zits

While t,he NPR f u e l element is currontly f e l t t o be adequate f o r use under the anticipated operating conditions, knowledge of the f u e l operating l i d t s is desirable f o r a number of purposes, S t r i c t l y speaking, large scale t e s t ing is required f o r an accurate statement of f u e l limits but much of value can be determined from small scale test- ing - information that should be usefid i n determining the direction of the fuels development program.

Testing of Process and Design Immovements

Characterization of the behavior of the firs+, load of fuel. i s curre:ntly the most important Job f o r the KER t e s t loops, of process and design improvements is beginning t o grow and w i l l continue t o do so. The leve l of e f fo r t on the development of the first load design and process has been so high as t o prevent- much at tent ion being paid t o process a l ternates Until. recently. Some work has been done, howeve:r, and the leve l of work w i l l increase, As promising al ternate approaches reach an appropriate stage of development, t e s t s w i l l be scheduled i n the loops.

The need f o r the t e s t ing

4

1. Alternate Closures

Several, a l ternate closures have been proposed and two currently look promising after a moderate amount. of development work. The mechanical pressure bonded closure i s being developed by HLO and the ga.s-p:ressure diffusion bonded closure is being developed by Fw). Both are expected, t o be ready f o r t e s t ing within the next year, a bonded closure without heating the f u e l element in to the uranium gamma phase, as is done i n the current brazing process.

Both processes produce

Hw-72680 Page 5

2.

3.

4.

5.

. .".. . Single Tube Fuel Design

Use of a single tube f u e l element with an i ne r t flow divider offers economic advantages over the tube-tube design. the higher tewera tures expected i n a single tube has been the great- e s t deterrent t o consideration of t h i s geometry, needed on the swelling rate i n the 500-600°c temperature region and the nature of the result ing f u e l deformation, limits of t h i s design are requi=d t o fur ther develop the incentive f o r an extensive t e s t ing e f fo r t ,

The swelling r a t e at

Information is

Some knowledge of the

Fluted We1

Shaping of the outer surf'ace of a tubular element, s o tha t o d , y compressive and bending s t r a ins occur i n the claddLng when the fue l swells, has been proposed as a means of extending f u e l l i f e i f the clad-thinning f a i lu re mechanism becomes limiting. tube-tube element i s currently deemed unlikely except, possibly, i n association w i t h an attempt t o reduce the cladding thickness. Use of the f luted concept i n the single tube design seems more l i ke ly because the swelling threat i s greater i n t h i s design,

Application t o the

Reduction of Clad Thickness

If, as is predicted, the tube-tube element turns out t o be reasonably trouble free, development programs aimed at cost reduction, by chipping away a t the conservatism i n the design, w i l l be insti tuted. Elimination of the brazed closure is one possibi l i ty , the single tube design is another, and reduction of cladding thickness is s t i l l another, The clad thickness of the first load element has been se t somewhat a r b i t r a r i l y and conservatively (we believe ). t i on i s obtained on fabrication capability, in-reactor f u e l performance, and swelling r a t e reduction poss ib i l i t i es through optimization of composition (see below), a b e t t e r basis f o r establishing cladding thickness should emerge,

As more informa-

Optimization of Fuel Coiaposition

Small alloying additions t o uranium can change the behavior of the metal during c o e x t m i o n and the result ing f u e l quali ty (e. g, , the Fe-Si additions now being made t o NPR fuel) , ing r a t e may be affected by the presence of allaying elements. B r i t i s h experience i n par t icular indicates tha t Fe-Al additions may lower the swelling rate appreciably, The e f fec ts of small alloying additions are, i n the case of both fabrication and swelling behavior, expected t o be incremental, Thus it appears tha t large scale tes t ing would be required t o allow selection of the optimum composition. Work is i n progress now t o improve the fabr icabi l i ty by a l te r ing the amout of Fe-Si additions and the heat t rea t ing schedule, improvement i s found, the appropriate process changes w i l l be made ahd test elements f o r the KXR loops w i l l reflect. the changes as soon as possible so tha t i r radiat ion tes t ing can be s ta r ted i n order t o check whether the f u e l performance is affected i n any way. s l i gh t composition changes w i l l improve the coext.ruded product, these changes would be made, on the assumption t h a t the swelling r a t e would not be increased. unduly.

In addition, the swell-

If significant

That is, if

C.

The problem of optimization may be approached from the other crirection. That is, we could aim t o f ind a f u e l composition tha t would minimize swelling,, and then check it l a t e r from the fabrication point of view. However, at the present t i m e , swelling does not appear t o be EL problem, but fabrication does, Thus the e f fo r t is being directed mainly toward the other approach,

The single-tube fue l element o f t a r s economic gains as a replac:ement fo r the double-tube element, SwellLng ts regarded as the greatest poten%ial problem with t h i s geometry because the mimum f u e l tempera- t.ure ture w i l l be about 100°C greater. Thus, the establishment of f e a s i b i l i t y w i l l involve determination of the swelling rate arid the consequences of swelling i n single tube elements i n the higher tempera- ture domain, If excessive swelling is encountered, then optindzation ,

of the composition (for swelling) would, possibly, make use of the single tube element feasible,,

The e f fo r t t o reduce the cladding thickness would be affected beneficial- l y by f u e l composition optimization f o r swelling also because a re- duced swelling ra te w i l l push the exposure l i m i t f o r a given cladding thickness out further.

Proof Testing of F i r s t Load Alternates

Consideration must always be gfven t o a l ternate materials and proc:esses when the success of a project depends on the sat isfactory completion of development work during the courrSe of the project, coextrusion appears t o be available, there are several backups t o other aspects of the f u e l element, proof tes ted depends on how badly you feel you may need it and on whether loop-space is available f o r the testing,

While no backup t o

!!%e extent t o which a backup should be

1,

-

2,

Backup Closures

The brazed closure has been regarded as one of the c r i t i c a l aspects of the first load element, of various end closures has l e f t something t o be desired, There is a need f o r a demonstration t e s t of a l ternate closures which may be regarded as a backup f o r the brazed closure presently specified f o r the first. load, The brazed closure has performed w e l l under i r radia- t i on i n the prototype tests completed t o date, s o the need fo~: backup tes t ing i n the closure area is diminishing but, nevertheless, s t i l l i s present,

Our a b i l i t y t o predict the behavior

Alternate Fuel Material

The use of unalloyed uranium a e l i n the NPR has been a ma t t e r of some concern because of (1) core corrosion of f a i l ed f u e l elements (2) swelling and (3) fabrication, The uranium-2$ zirconium a l loy has been considered as a backup f u e l m t e r i a l because it has 2%

lower corrosion ra te than uranium and because a be t t e r coextrudad product can be obtained with it. It appears t o have a higher swell- ing rate but lMt .ed data suggests t h a t the cladding uniformity is suf f ic ien t ly improved t o offset. t h i s and produce a longer-lived fuel. I r radiat ion performance data i n prototypic geometry would be desirable.

D.

E.

Phase P I and Phase P I 1 Fuel %st ing

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As time goes on, the emphasis on first load t e s t ing w i l l diminish and s h i f t toward t e s t ing of improvea Phase I elements, A t the same time, consideration should be given t o the t e s t ing of Phase I1 and Phase I11 f u e l elements.

'

Phase I element only s l i gh t ly so the tes t ing e f f o r t w i l l be a minor one. Phase 111 operation seems t o be a lqng way i n the future but the design c r i t e r i a w i l l be quite different from tnose f o r the Phase I element so that a major t es t ing e f fo r t w i l l be required, Further, because the goal exposure f o r the Phase 111 element w i l l be high, very long te:rm tests w i l l be required i f significant information is t o be obtainetl. Although Phase 91 and Phase I11 element tests w i l l be minor e f fo r t f o r the near future, the emphasis i n t h i s direction eventually w i l l increase.

The Phase II element is expected t o d i f f e r from the

Miscellaneous Tests

1"

2.

3.

4.

Water Quality Studies

Zircaloy-2 is a very corrosion-reeistant cladding material i f the cooling is adequate. Crud deposits w i l l r a i se the cladding tempera- ture and may, i f heavy enough, lead t o ear ly f a i lu re of the chdding by corrosion, performed t o date have shorn that crud should not be a problem under normal operating conditions, Tests are needed yet t o determine whether t h i s is equally true during periods of abnormal operation such as t ha t period immediately following a decontamination,

Crud deposit.ion tests with thermocoupled elements

Development of a Crud Monitor

For the reason staked i n the preceding paragraph, it has been decided t h a t several crud monitors shall be i n the NFR a t a11 times f o r the first year of operation. This requires that an adequate crud monitor be developed and tes ted t o an extent suf f ic ien t t o demonstrate a high probabili ty t ha t the crud monitors adopted f o r NPR usage w i l l operate without fa i lure ,

I r radiat ion of Defected Elements

I r radiated elements are needed f o r the PRTR rupture loop program., A t present, the only proven scheme f o r "push-button" failure o:? an i r radiated element requires tha t a defected element, with a cover welded over the defect, be irradiated i n the KER loops t o the desired exposure leve l and then transported t o the P E R f o r the rupture t e s t . Thus, space must be macle available i n the loops f o r i r radiat ion of defected elements,

Development work is i n progress on ways t o defect a sound irradiated element. If t h i s work is successful, all elements from HER loop t e s t s would be potent ia l candidates f o r f a i lu re tests and the need f o r the specially prepared elements would be great ly redaced OjY eliminated,

Service Irradiations f o r Other Programs

When loop space is avaflable tha t would otherwise be wasted, irradia- - - t ions i n support of other progsams may be carried out.

HW- 72680 Page 8

Generally, a separate t e s t is not required f o r each of the objectives l i s t e d i n the previous section. In the water quali ty t e s t s , f o r instance, highly specialized t e s t s are called f o r , but, u s ~ a l i y , a given f u e l element t 'est w i l l contribute t o several of the problem areas, In some cases, (ee g., fue ls plant quali ty mosaitorbg) it is almost inrposstble tc design a t e s t t h a t w i l l provide meaningful data, In other C ~ E ' C (e res t ra in t of w a r p ) , it, should be possible t o devise a good test, but t h i s has not been accomplished yet.

A. Tests i n Progress

The tests described below have been developed t o provide information bearing on one cr more of %he problem meas Listed i n Section I.

1, Prototype Tests

The term "prototype testing", as used i n t h i s discussion, refers t o the i r radiat ion o f M fuel. components, o r f u e l elements reasonably close t o them i n dmmnsions, composition, support hardware, e t cetera, under temperature, power and eqosure conditions typical of those which are expected i n the cent ra l region of a f la t tened zone process tube with the reactor operating a t a power of bo00 MW and w i t h a goal exposure of 1400 IWDIT. (Peak exposures am: outer tube - 2300 IWD/T, inner tube - 1700 W I T , average - 2100 W / T ) . Because of equipment l imitations, only the central 5 t o LO f ee t of the KER loops are charged w i t h fuel , Thus, exposure quotations i n the loop t e s t s refer t o the peak value rather than ts a t,ube-average value and should be compared t o the a 0 0 IWD/T figure above rather than t o lL00 MwD/T.

Prototype t e s t ing began i n November, 1960, when the i r radiat ion of braze-closed Km3, N I N l and NEE1 elements w a s started, KSE3 elements are thin-walled, large diameter tubes, and as such, are regarded as a stand-in f o r the actual N outer tube, whichsannot now be irradiated i n the KER loops, component ( N I E I ] only i n that they contain natural rather than enriched (Oo947$) uranium and they do not contain the iron and s i l icon additions (about 100 and 50 ppm respectively), the same power i n the KER loops as the peak power of the inner compo- nent i n the NPR' element produces s ignif icant ly more power i n the KER loops than it w i l l i n the NgR and, therefore, fuel temperatures are high i n :N IE1 tests i n the KER Loops,

NINL elements d i f f e r from the actual NPR inner

NINl elenents produce about

Without the shielding of the outer tube, the NU1

On successful completion of the current KER-2 i r radiat ion, t e s t s of KSE3 and NINI-NIEl elemezts t o roughly 1000, 2000 and 3000 MKD,/T wila have been completed. Prototype tes t ing w i l l s h i f t t o KE3-3 and KER-4 exclusively on completion of CGI-839 (the loop modification project) , Full size, complete N elements, fabricated by the approved process, and intensively charactepized, w i l l be i r radiated t o average exposures of" '700, 1400 and 2106 MWD/T, are 1000, 2000 and 3000 MWD/T. infomiation i n most of the categories under A (Characterization of "First Load" Behavior),

The corresponding peak exposures This ser ies of tests w i l l provide

RW-72680 Page 9

2, Crud Monitor Evelopxent Tests

Crud is not expected t o be a problem under normal operating conditions i n the NPR. Hcwever, a number of tranaient conditions may lead t o crud formation, Crud on the f u e l element heat t ransfer swfactes w i l l induce auto-catalyt IC cladding corroslon, Therefore it has beten decided tha t crud monitore shoidd be used i n the NPR. The current approach is ts measure the central temperature of a f u e l sample, using the behavior w i t . h tinie t o iilnicate whether anything is happening that would cause the f u e l temperatuses t o increase, a suitable design f o r the crud monitor is i n progress. The first design was found through in-reactor t e s t ing t o be inadequate so a new design is being studied ~ Q W wfth a KER loop t e s t now under way i n KER-I. t o demolistrate that %he design and fabrication process w i l l be adequate f o r B Reactor usage

Development of

Subsequent t e s t s w l i l be acheiiuled t o an exsent suff ic ient

3. Crud Measurement, St;ud;,es

Study of water chemistry changes a x the e f fec t on crud deposition can proceed i n pa ra l l e l with the cmd monitor development t e s t s t o some degree, The sens i t iv i ty requirement; f o r the I P R crud monitor is not as great, as desired f o r study of water chemistry e f fec ts so the lat ter w i l l suffer t o some extent, We do not. yet know the sens i t iv i ty of the present crud monitor design, so the degree of incompatibility of the zwo objectives cannot. be assessed at t h i s t i m e ,

Tests proposed f o r t h i s aspect of the Loop psogrm include: a) study of crud deposition a f t e r a decontanination, b) measurement of crud deposition at pH values below 10 (%o aid i n se t t ing operating :Limi ts ) and c ) evaluation OF ammonia as a pH control m t e r i a l (instead of lithium hydroxide). the sens i t iv i ty is determined, KERr.1 w i l l be decontaminated anti the themocouple readings taken over a, period of several weeks, If no temperature increase is noted and Bf the thermocouple continuers t o operate, the pE w i l l be lowered t o about 9 and the ~'hsmoeoup1e readings followed, If no increase i n temperatuxe is noted, the pH w i l l be lowered further, e tc , wit.h pH-7 being the hwer l i m i t , The primary indicator of crud deposition i n these t e s t s w f l l be the temperature behavior with post-srradiation examination as a backup.

When an operating crud monftor is i n place, and

4, Tapered End Cap Test

The c r i t i c a l aspects of the braze closure rzm f e l t t o be: it be mass-produced w i t h high yields and i n good, quality? the cap-to-fuel bond stand up under irradiation?

Backup closure designs then f a l l into two categories - backups f o r brazing pes se and backups t c the present brazed closure design, '

The best backup of the former var ie ty is f e l t t o be a simple welded (unbonded) closure, However, the f la t end cap design with a fltush end cap has been shown t o have a short l ifetime (the thel-mai2incompati- b i l f t y of uranium and zircolifum leads to shear fa i lure of the <:ladding at the cap-fuel Juncture) and use of a gap between the cap and fuel appears t o be unaesirable from the point of view cJf handling damage. Thus the best welded closure i s currently f e l t t o be one i n which the

1) Can 2) W i l l

m7-72680 Page 10

cap and fue l m e essent ia l ly flush but the contact, surface is V-shaped i n cross-secrional view, This shape is expected t o Lead t o a reduced fa i lure potent ia l by extension of the length of &ladding over which the thermal incompatibility operake, This type of closure has not been tested,

If the braz-hg process is found t o be satisfactory, but a f a i lu re problem develops as a resu l t OF P r a c t m of the cap-to-core bond followed by cladding failure by the shear mechanism, then it appears that the next, best, design is a brazed closure s i t h the same shape of cap and fue l discussed above. a brazed closure is t o redme the shear s t r a i n on the cap t o file1 braze tha t results from the thermal incompatibility of uranium and zirconium.

The effect of the tapered end cap i n

Tnis type of cILosure also has not been tested,

A test has been prepared involving ten M%L elements with tapered end caps, half of the charge being unbonded ( w i t h no gap) and half being brazed ( w i t h Zr-5 Be). !The objectives of the test are t o demonstrate the capabili ty of the two backup closures t o overcome the potent ia l problems just noted and t o provide information on the behavior of end caps of these two designs.

5* Warp Test

The decision t o use straightened f u e l elements i n the NPR is a source of some concern, Qur understanding of the causes and mechanisms of warp - during fabrication and also during i r radiat ion - is not w e l l enough developed t o provide posit ive assurance tha t straightening w i l l not lead t o a w a r g problem, Thus the recent KER-1 t e s t was designed: straightened f u e l t o w a s p , m d 2) t o provide sone information on the correlat.ion, i f any, mong straightness, warp during heating, and in-reactor warp. Fuel elements are included i n the t e s t which were far beyond what is currently regarded as accepta3le f o r straightening and elements which were acceptable i n all respects but came from extrusions which were very crooked, scheduled as development studies indicate tney are warranted.

1) t o provide some information on the propensity of

Further warp tests may be

B. Tests Cur-ntly Under Consideration

The t e s t s l i s t e d below are being act ively ccnsidered f o r CY-62, have not Seen scheduled yet and preparatioE of the t e s t elements has not yet been s tar ted, and the settlement of one o r more de ta f l s of the % e m design must be established o r fur ther development work m u s t be done before work on the t e s t elements can be started,

Thiey

The need f o r the zests has essent ia l ly been esta,blished

1, Tests of New Closures

Development work is i n progress on several new closure processes. When and i f each of these programs reaches the point where reactor grade material can be produced, and t,he closure i n question holds suff Scient promise of cost reduction and/or product improvement re la t ive t o the present desigr,, tests w i l l , be scheduled i n the KER loops. ready u n t i l l a t e t h i s calendar year,

It appears now that no t e s t s of new closure designs w i l l be

2. Alternate Fuel Material Test

It w i l l be some time before we have posit ive assurance t h a t unalloyed uranium is an adequate fue l from the point of view of failure behavior. Although use of U-2 Zr f u e l would increase the cost of separation of plutonium from the i r radiated fuel , the corrosion resistance of t h i s a l loy makes it an a t t rac t ive al ternate o r backup f u e l material. It is desirable t o i r rad ia te some elements of ITIN1 geometry using;U-2 Z r f u e l during the next year, The objective would be t o determine the dimensional behavior and t o provide i r radiated elements f o r IRP and PRTR rugture loop t e s t s ,

3. Water Quality Tests

In the one successful t e s t w i t h a thermocoupled element t o date, no evidence of crud deposition was observed during nomal operation (at pH-10, maintained with LiOH additions), The t e s t scheduled next fo r KER-1 w i l l involve study of crud deposition a f t e r a decontamina- t ion, i n addition t o being the first t e s t of the new design f o r the N reactor crud monitor, There is some concern over potent ia l corro- sion problems w i t h LiOH as a pR6 control additive. desirable t o t e s t another approach. a f t e r the currently scheduled test probably w i l l be carried out using ammonia as the pH control additive,

Thus it appears The next water quali ty test

4. High Temperature Swelling Data

The swelling data obtained t o date indicates t h a t swel l ing should not be a problem i n the first load element. o f fe r economic advantages over the tube-tube element, but extrapolation of the swelling data we have obtained i n the prototype t e s t s suggests t ha t it may be sederely l imited by swelling, we w i l l want t o explore these limits during the next f e w years., Tests of actual single tube designs must. be carried out i n KER3 and KER-4 (because of the process tube s ize) but these loops w i l l not be available f o r t h i s purpose f o r a year o r more. I n the meantime, our swelling data could be extended nearly to 550°C by i r radiat ion of KSE-3 elements i n the smaller loops with the loop out le t tempera- ture s e t at o r near the maximum permitted,

A single tube element would

It is expected that

5. C r u d Monitor Development Tests

The next crud test w i l l probably determine whether the new deaign is adequate. If it is not, another design w i l l be tried. Af't'er completion of a successful t e s t ("success" w i l l be determined by (a) did the monitor element survive without f a i lu re and (b) w a s the sens i t iv i ty found t o be adequate) a fabrication process w i l l be devised which w i l l economically produce the required number of monitors (currently estimated at about 20 for the first year). It i s deemed necessary a t t h i s time t o undertake something resembling volume tes t ing of the design selected. t o the best way t o accomplish th i s objective. without thermocouple leads as heater elements f o r l a t e r crud t e s t s currently is favored.

Consideration is being given Use of crud rconitors

111.

Iv.

Pr ior i t ies

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Stated very broadly, the order of pr ior i ty f o r the various tes t ing ac t iv i t i e s i s f e l t t o be as follows:

1. Prototype t e s t s 2. 3. Crud monitor development t e s t s 4, Water quali ty tests 5. 6 -

Proof tes t ing of first load alternates

Testing of process and design ilnprovements Testing of Phase I1 andphase I11 fue l designs

On an individual t e s t basis, the next crud and crud monitor test is regarded as being of higher pr ior i ty than the backup closure t e s t (tapered end caps). However, generally speaking, the above list re f lec ts the current feeling on pr ior i t ies . As an indicator of our current thinking, the following estimate of p r io r i t i e s one year from now is given:

1. 2. Prototype tests 3. Crud monitor development t e s t s 4, Water quality t e s t s 5. Testing of Phase I1 and Phase I11 f u e l designs

Testing of process and design improvements

I

I Schedule

T e s t loop schedules must be regarded with som degree of skepticism. Past experience has demonstrated our inabi l i ty t o predict precisely when certain tests w i l l be performed o r how long they w i l l remain i n the reactor. With some reservation, then, a schedule is presented (Figure 1) which represents the course w e now f e e l w i l l offer the best chance of achieving the tes t ing objectives mentioned ear l ier .

Loop Modif ication

XER-3 and HER-4 are currently being modified under CGI-839. tha t XER-1 and KER-2 be removed from service f o r modification soon a f t e r XER-3 and KER-4 resume operation. and KER-4 w i l l be r e d y f o r resumption of tes t ing before the middle of March. t ion after the current crud t e s t (mid-April) and KER-2 after completion of the high exposure NIEl t e s t currently in XER-2 (April),

It i s desired

It is currently anticipated tha t KER-3

Thus, it is now planned tha t KER-1 w i l l be released f o r modifica-

n

st 3 Y

r! d u

. .

V. Reconciliation

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The KER loop tes t ing objectives and the t e s t s i n progress and planned have been discussed and the schedule has been presented, the par t icular t e s t s i n progress contribute t o achieving the s ta ted objectives and whether all the objectives are covered is i n order.

Some discussion of how

The prototype t e s t s with N @ l elements ~ l m n e i for KER-3 and KER-4 cover those par ts of I A en t i t l ed Metallurgical Behavior, Design Testing and C l a d d i n g Corrosion quite well. They cover Fuels Plant Quality Monitoring, Surface Contamination, Frequency and Severity of Warp and Fuel Element Limits as w e l l as we f e e l we can do now, The item, Restraint of W a r p , is essentially ignored.

Alternate Closure Tests w i l l be scheduled when t e s t elements are available. High Temperature Tests of KSE3 elements should provide data of use i n assess- ing the limits of the single tube fuel design, In l i e u of t e s t & of actual elemezts. Composition are ignored here and must be considered a8 part of next yelar's program,

The Tapered End Cap Test constitutes our e f fo r t on F i r s t Load Alternates, Alternate Fuel Material being ignored. Phase I1 and Phase I11 test ing has not been included here except insofar as characterization of the first load design and investigation of 1Wts f o r the first load element bear on assess- ing the potential of 2s-2 clad, coextruded uranim elements f o r Phase I1 and Phase I11 application.

Fluted Fiel , Reduction of Clad Thickness and Optimization of Fuel

The Water Quality Studies and Development of a C r u d Monitor are covered w e l l by the tests planned although volume tes t ing of the crud monitor is not, as yet, a tes t is planned f o r the ETR t h i s spring of the device f o r defecting irradiated elements in-reaktor,

Four irraiated, defected elements are available fo r PRTR tests and

I

- W. &A&+- KO Kratzer T, w. mans Reactor Fuels Unit Process and Reactor Development Research and Engineering Section IRRADIATION PROCESSING D E T A R W N T

Reactor Fuels Unit Process and Reactor Development Research and Engineering Sectdon IFWDIATION PROCESSING DEPARI'MEN!P