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R L-2-17
Cover Sheet for a Hanford Historical Document Released for Public Availability
Released 1994
Prepared for the US. Department of Energy under Contract DE-AC06-76RLO 1830
Pacific Northwest Laboratory Operated for the U.S. Department of Energy by Battelle Memorial Institute
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.
*
* RL-2-17
Hanford Code C-44
Date 6- \ -66
RICHLAND FIVE-YEAR
02 R&D P R O G R A M
HIGHLY ENRICHED FUEL PROGRAM
RICHLAND OPERATIONS OFFICE DOUGLAS UNITED NUCLEAR, INC.
GENERAL ELECTRIC COMPANY - HAPD ISOCHEM INC.
BATTELLE - NORTHWEST
R L 2-17
Hanford Code C-44
R I C H L A N D F I V E - Y E A R
0 2 R & D P R O G R A M
H I G H L Y E N R I C H E D F ' U E L P R O G R A M
RICHLAND OPERATIONS O F F I C E
DOUGLAS UNITED NUCLEAR, INC.
GENERAL ELECTRIC COMPANY - HAPD ISOCHEM, INC.
BATTELLE MEMORIAL INSTITUTE
P L E A S E S I G N P s R s
B E F O R E R E A D I N G Files
R o u t e To: , Nos , L o c a t i o n . R o u t e D a t e . Signature and D a t e /
&.€&+ 8/&2 h \
AUG 8 1966 3 b d n 3 321, 1
- J -
R L 2 - 1 7 Page 2
R I C H L A N D F I V E - Y E A R
0 2 R & D P R O G R A M
H I G H L Y E N R I C H E D F U E L P R O G R A M
PRINCIPAL CONTRIBUTORS
W. S. Nechodom - GE H. C. Rathvon - ISO R. H. Meichle - WN . ,
c
R L 2 - 1 7 Page 3
R I C H L A N D F I V E - Y E A R
0 2 R & D P R O G R A M
2.
SUMMARY
B A S I C PRODUCTION MISSION
COPRODUCT MISSION
RL 210
RLI 2-1
RL 2-2
INTEGRATED SITE OPERATION RL 2-3
NOH-DEFENSE PLUTONIUM MANAGEMENT RL 2-4
TRANSPLUTONIUM PROGRAM
PU-238 PROGRAM
U-233 PROGRAM
RL 2-5
RL 2-6
RL 2-7
co-60 PROGRAM RL 2-8
PO-210 PROGRAM RL 2-9
HIGH POWER DENSITY FUEL P R O G W RL 2-10
REACTOR MIDERNIZATION
FUNDAMENTAL AND GENERAL STUDIES
NUCLEAR SAFETY PROGRAM
COLUMBIA RIVER STUDIES
WASTE' MANAGEMENT
OTHER ISOTOPES
HIGHLY ENRICHED FUEL PROGRAM
RL 2-11
RL 2-12
RL 2-13
RL 2-14
RL 2-15 7
RL 2-16
RL 2-17
R L 2 - 1 7 Page 4
F
TABLE OF CONTENTS
SUMMARY
INTRODUCTION INCENTIVES SCOPE AND OBJECTIVES PROGRESS DURING REPORT PERIOD BUDGET PERIOD PLANS AND EXPECTED RESULTS
GENERAL ELECTRIC COMPANY
SCOPE AND OBJECTIVES BUDGET PERIOD PLANS AND EXPECTED RESULTS
DOUGLAS UNITED NUCLEAR
SCOPE AND OBJECTIVES PROGRESS DURING REPORT PERIOD BUDGET PERIOD PLANS AND EXPECTED RESULTS
ISOCHEM, INCa
SCOPE AND OBJECTIVES PLANS AND EXPECTED RESULTS
i
Page
5
5 5 7 ' 13 13
17
17 22
26
26 26 26
30
30 30
: r
R L 2 - 1 7 Page 5
HIGHLY ENRICHED FUEL PROGRAM
SUMMARY
INTRODUCTION
As t he mi l i ta ry needs f o r plutonium decline, t h e Commission's production- oriented reactor complex should develop and demonstrate pure non-defense operating modes t o serve non-defense needs, The a l te rna t ive is production reactor shutdowns coupled with needless expense i n producing necessary products elsewhere o r doing without them, Savannah River already has secured such pure non-defense capabi l i ty , at l e a s t from a technical f e a s i b i l i t y s tandpoint , i n t h e i r fully-enriched, isotope-producing loadings and i n t h e i r Pu-burning, p a r t i a l loadings, Although t h e use of f u l l y enriched fue l (partial . o r complete loadings) fo r t h e Richland production reactors was once practiced, renewed study and research t o rees tab l i sh p rac t i ca l incentives, demonstrate technical f e a s i b i l i t y (including reactor sa fe ty) , and f i r m up product capabi l i ty must be undertaken ,
It is unquestionably t r u e tha t fully-enriched uranium can be u t i l i z e d as fue l fcr t h e Hanford reactorso It is the purpose of t h i s mission t o develop the conditions under which t h i s may be accomplished safe ly and productively, t o determine the necessary plant modifications required t o achieve t h i s processing capabi l i ty , t o disclose the inherent plant l imitat ions t o such processing, and t o evaluate t h e incentives f o r operating on f'ully enriched uranium fuel ,
INCENTIVES
The purpose of most of the Hanford R&D programs i s t o develop methods of diver t ing t h e productive capacity of t h e Hanford plant to '- the production of non-weapons products; i o e o s products other than weapons-grade plutonium and t r i t i u m , To accomplish t h i s , t he reactors must be fueled so as t o produce an excess of neutrons, beyond those required t o sustain the f i s s ion process, which' are'= absorbed i n U-238 o r i n Li-6, and straightforwardly accomplished by eliminating these absorbers from the reactor core, and subs t i tu t ing ta rge t atoms which w i l l produce more desirable products upon neutron absorption, For complete diversion of production t o non-defense products, fully-enriched fue ls , e i t h e r U-235 o r plutonium, must be u t i l i zed ,
T h i s i s most Peadily
R L 2 - 1 7 Page 6
A var ie ty of neutron-irradiation-produced isotopes have been found t o be of i n t e r e s t t o those engaged i n developing thermal-electric power sourceso These isotopes ranee from the f i s s ion products such as strontium and cesium, t o the transpluton$.um isotopes such as curiump They include several which are produced by simple neutron capture i n t a rge t atoms, such as cobalt-60, plutonium-238, thallium-204, polonium-210, and thullium-170,
The degree of i n t e r e s t i n these isotopes var ies , depending on t h e poten t ia l application being studied, and perhaps the state-of-the-art i n a pa r t i cu la r applications areao Studies of production methods also vary; i n somewhat d i r ec t proportion t o t h e i n t e r e s t evidenced, these s tudies range from simple estimates of production capabi l i ty t o scoping s tudies of possible i r r ad ia t ion techniques t o fu l l - sca le design s tudies and tes t i r rad ia t ions . It appears inevi table , and not undesirable, that several o r a l l of these isotopes w i l l ul t imately f ind the i r way i n t o production programs, s ince t h e var ie ty of power source missions f o r which they are required w i l l r e s u l t i n finding t h a t each isotope i s optimum fo r cer ta in spec ia l applications . If one makes the (hopefully cor rec t ) assumption t h a t tihe production complex w i l l be faced with an unstable and f l ex ib l e product demand, it i s only log ica l t h a t a single , o r a few, modes of reactor fuel ing be developed which w i l l su f f i ce t o produce any product mix i n a near-optimum manner, It does not appear log ica l , and would impose an unrea l i s t i c demand on fue ls fabr icat ion and processing f a c i l i t i e s , t o attempt t o accomplish more than t h i s .
Set t ing up t en ta t ive c r i t e r i a :
a) Elimination of weapons-grade plutonium production,
b) I r rad ia t ion of a number of d i f fe ren t t a rge t isotopes,
c ) Possible production of f a s t reactor feed grade plutonium plus transplutonium isotopes, and
Production of e l e c t r i c power ( i n t h e case of N-Reactor), d)
One concludes t h a t . f u e l systems of e i ther U-235 o r some plutonium isotopic m i x should be investigated, Should these then prove infeasil j le, t he m a j c i i n m feasible concentration of U-235 i n uranium, and of plutonium mixed w i t h a resonance absorber, should be sought, Therefore, t r u s t i n g t h a t su i tab le target isotopes w i l l be provided when required, these s tudies w i l l be undertaken
SCOPE AND OBJECTIVES
c
1, Reactors I
A basic assumption i s made t h a t other programs w i l l develop the need fo r , and the technology of producing, t h e target materials, T h i s '
mission is designed t o prevent "reinventing t h e wheel" f o r each t a rge t proposal, However, one must recognize t h a t t he t a rge t s must be put somewhere - i r r ad ia t ion space must be provided, which leads inevitably t o the question of how mucho Since t h e only reasonable premise is; "as much as possible," t h e program must begin on t h i s noteo The goal of developing "high power density" fue l i s another mission - unfortunately, it is not p rac t i ca l t o a t tack the problem s e r i a l l y within the t i m e a l l o t t e d (assumed t o be four years) , (commonly known as a guess) i s therefore made t h a t one-third of t he i r r ad ia t ion volume can be devoted t o t a rge t space, while operating t h e remaining two-thirds at a power density which w i l l del iver t h e reactor design power leveld (This i s not necessarily r a t iona l - there m u s t be some optimum combination of target-fuel r a t i o and power level. ) Variations i n i r r ad ia t ion volume and reactor power l e v e l are therefore excluded from the scope of t h i s study, The scope of t he study w i l l be confined t o :
An a rb i t r a ry engineering judgment
a) Physics behavior
1 ) ''Prompt I' e f fec t s 2) Total r eac t iv i ty e f f ec t s 3) Transient r eac t iv i ty .effects
b )
c ) Reactor effects
Fuel materials (shared with High Power Density Mission)
1) 2)
Moderator temperature and l i f e considerations Zirconium tube temperature and l i f e considerations
a) Fuel handling problems - c r i t i c a l mass considerations (shared w i t h Fuels Preparation and Reprocessing Sec t ioss ) , * +
1,1 Physics Behavior
1,1,1 Reactor Control - I'Prompt" Effects
Once the U-238 resonance absorber i s eliminated from t h e fue l composition, t he benefi t of t he doppler e f f ec t as an a i d i n control l ing reactor excursions i s l o s t (except fo r t he e f f ec t of t he Pu-240
R L 2 - 1 7 Page 8
resonance i n plutonium fue ls , which w i l l be explored), t he e f f ec t of delayed neutrons is (never fear) s t i l l available, I f one recognizes that the fue l heating t i m e constant i s not zero i n present fue l s , it is conceivable t h a t t h e l o s t doppler could be replaced by t h e rapid scram act ion of control elements; yielding an adequately control led ' system, some combination o f t (a) engineered safeguards t o inh ib i t large, rapid additions of r eac t iv i ty , (b) instrumentation f o r ear ly detection of an incipient excursion, and ( c ) a reasonably rap id scram system, w i l l solve t h i s problem,
However, ,
It seems l i k e l y thalj
l,lo2 Reactor Control - Total Reactivity Effects
Highly enriched fue l loads w i l l be characterized by large r eac t iv i ty changes which are generated as t h e f ie1 burns, by (possibly) la rge s t a r tup t r ans i en t r eac t iv i ty e f f ec t s , and (possibly) by adverse posi t ive r eac t iv i ty changes i f coolant is lost , The .burnup r eac t iv i ty changes can be made manageable by: ( a ) Possessing or i n s t a l l i n g a very la rge capacity control system - hopefully being imaginative enough t o fabr ica te it so as t o produce useful products when irradiated, (b ) Charging t h e h e 1 i n timed increments so t h a t only a s m a l l par t of t h e burnout t rans ien t i s suffered between charging, This, of course, must be adjusted t o be within the avai lable control capacity, Differences i n channel power inh ib i t t h i s ac t iv i ty . ( c ) progressesa One can imagine e i ther changing t h e density of a pa r t i cu la r t a r g e t isotope, o r per iodical ly changing t o a new t a rge t isotope, depending on product demand, (d) Uti l iz ing "burnable" t a rge t isotopes (which, of course , they are anyway) ,, T h i s confines the target candidates t o reasonably high cross sect ion materials, probably d i lu ted w i t h i n e r t materials, so t h a t j u s t t he r igh t f rac t ion of targets w i l l burnout t o counterbalance the f u e l burnup, Tricky but possible, ( e ) A combination of a l l of these - t o some extent inevi table , wandering around i n permutations forever, and make a s tu t e choices based on survey s tudies ,
Periodically removing par t of t he t a rge t poisons as burnout
One must guard against
i
lolo3 Reactor Control - Transient Effects
Transient r e a c t i v i t y e f f e c t s and t h e i r control a r e d i f f i c u l t t o define, s ince they tend t o be very sens i t i ve t o f u e l and t a r g e t compositions and or ien ta t ion , and t o f u e l exposures , Calculatidns have disclosed that, when t a r g e t isotopes a re lumped so as t o be thermally black, and the t a r g e t pieces a re placed i n channels separate from the dr ivers , t h e temperature t r ans i en t s a r e l a rge and negative, T h i s is a l s o t r u e i f t he t a r g e t is placed concen- t r i c a l l y ins ide t h e f u e l element, However, i f t h e t a rge t is placed immediately outside t h e f u e l , it can be made t o e f fec t ive ly i s o l a t e t h e f u e l from the spec t r a l e f f ec t s of moderator temperature changes, and pos i t ive r eac t iv i ty t r ans i en t s w i l l r e s u l t , Thus, a t l e a s t one mechanism is ava i lab le t o t a i l o r t he r eac t iv i ty e f f ec t s t o control- l a b l e values, already i n s t a l l e d at t h e older Hanford reactors , and a re under development f o r t h e N-Reactor, These mechanisms can be 'arranged t o compensate f o r l a rge swings - up t o 5-7 percent k - i n reac t iv i ty ,
Supplemental r eac t iv i ty control mechanisms a r e
The r eac t iv i ty change resu l t ing from reactor coolant l o s s can be pos i t ive o r negative, depending on the pa r t i cu la r reactor load configurationc provided t h a t t h e response of t h e control system is fast enough t o overcome the pos i t ive r eac t iv i ty ramp, and t h a t t h e worth of t h e control system i s grea te r than the dry excess r eac t iv i ty , The l a t t e r requirement may r e s t r i c t t he loading designs t o those w i t h higher f i s s i l e material concentrations , and consequently t o "blacker" t a r g e t systems, Calculations ind ica te that the worth of control elements and t a rge t s does increase markedly with coolant loss due t o t h e increased neutron migration area i n the absence of coolant, By l'lumging" t h e t a r g e t i n t o well-spaced "control elements" t h i s e f f ec t may be enhanced, and should provide adequate control strength, Designs which depend on t a r g e t control worth under coolant loss conditions a lso r e s t r i c t t he t a rge t design t o materials which w i l l remain i n t a c t under these high temperature conditions,
A pos i t ive r e a c t i v i t y change is perfec t ly acceptable,
1,2 Fuel Materials I
Many 0-f the problems which must be considered under t h i s subject may be assigned t o t h e H i g h Power Density Program, However, some problems would not a r i s e a t a l l but fo r t h e attempt t o i r r a d i a t e ful ly enriched fue l s , and $re sponsored by t h i s program,
E
R L 2 - 1 7 Page 10
Oralloy fue ls have t r ad i t i ona l ly consisted of a l loys of uranium and aluminum which are readi ly fabr icated and reprocessed. it is not at all cer ta in t h a t these al loys w i l l be compatible w i t h modernized f i s s ion product confinement philosophies; cer ta in ly , s tudies of f i s s ion product re tent ion i n melting metal l ic uranium would not apply,
However,
Certain fundamental c r i t e r i a w i l l govern t h e choice of f u e l materials; ( a ) substances, else f i s s ion density and aCcOmPWYing exposure-dependent e f f ec t s could not be to le ra ted , Dilutions by a fac tor of 20-100 are expected t o be required, . (b) The diluent must have a low neutron capture cross section, The resu l t ing f u e l a l loy must not pass through phase changes involving volume differences between shutdown and operating points, The f u e l a l loy must resist swelling with exposure; such changes w i l l upset the thermal-hydraulic balance of multi-coolant-channel elements and fur ther , may endanger the i n t e g r i t y of the f u e l cladding, must r e t a in f i s s i o n products f o r a t i m e , or t o an extent, which satisfies c r i t e r i a r e l a t i v e t o the release of f i ss ion products t o t h e environs, (f) The fue l must lend i t s e l f t o mass-production fabricat ion and canning, cast ing and canning, e tc , Fabrication processes which inherently generate l a rge amounts of f i n e f u e l pa r t i c l e s are less desirable, because of c r i t i c a l mass control problems, (g) The fuel m u s t be reprocessable i n sone exis t ing p lan t , a t a r a t e compatible with large-scale reactor i r rad ia t ions , c r i t i c a l mass control are considered inevi table , and separate from t h e fue ls material problems, However, it is t o be hoped t h a t t he choice of fuel materials can be guided by ex is t ing chemical technology i n t h e dissolving and pur i f ica t ions s teps ,
The ac tua l f iss i le material must be d i lu ted many-fold by i n e r t
( c )
[d)
(e ) Upon coolant loss, t h e f u e l assembly
Plant modifications t o insure
1,3 Reactor Effects
1,3,l Moderrartor Effects
A small f rac t ion of t he energy of each f i s s ion i s released as gamma radiat ion, This energy is then deposited i n each posi t ion of the l a t t i c e according t o t h e gama absorption cmss section of each material, and i t s density, It follows t h a t t h e removal of the uranium438 f’rom the f u e l grea t ly reduce6 t h e m a s s ; the gamma energy deposition w i l l be redis t r ibuted, and a la rge share w i l l f a l l t o t h e graphite moderator, Since t h e heat generated ‘must then be conducted back t o the coolant heat s inks, it follows t h a t the moderator temperature w i l l rise u n t i l a new equilibrium heat balance i s
R L 2 - 1 7 Page 11
establ ished, be compatible with minimizing graphite growth o r contraction phenomena, o r with graphi te burnout l i m i t s i n cases where oxygen o r oxides of carbon a re present as par t of t h e reac tor atmosphere,
The temperature of t he moderator may or may not
About 30 percent of t h e gamma energy generated i n standard metal l ic uranium fue ls is "prompt" gamma released upon neutron. capture i n uranium-238, These neutrons w i l l now be captured i n some s u b s t i t u t e t a r g e t mater ia l , I n some cases, t h e prompt energy re lease w i l l not be i n t h e form of gamma rad ia t ion , but w i l l appear as r e c o i l energy i n alpha p a r t i c l e s , energy i s confined t o the d i r e c t l y cooled t a rge t , and t h e f r ac t ion of t h e f i s s i o n energy deposited i n t h e graphi te moderator is reduced,
I n these cases, t h e r e c o i l
1,302 Zirconium Tube Effec ts (N-Reactor)
If t h e net heat t r a n s f e r from the graphi te moderator t o t h e coolant streaui is increased, t h e temperature r i s e across t h e zirconium process tube w i l l increase proportionately. A s ign i f i can t increase (more t h b , sw, 10 percent) may cause the outer surface of t h e tube, at some points , t o reach temperatures incompatible with t h e prevention of gas phase hydriding w i t h t he ex is t ing reac tor atmosphere Research i n t o i n h i b i t i n g atmospheres w i l l probably be an e s s e n t i a l p a r t of t h i s program i f a "universal" f u e l system, compatible with a l l possible t a r g e t s , is required,
Both of t h e above problem areas a r e "local" e f fec ts , r e l a t ed t o l o c a l power density conditions, not gross reac tor power, magnitude of t h e problem may therefore be reduced by simply reducing t h e r a t i o of maximum t o average power density, accomplished f o r years i n t h e r a d i a l reac tor dimensions by se l ec t ive f u e l loadings, Flat tening i n t h e a x i a l dimension has not been s ign i f i can t ly explored, and may prove t o be t h e simplest solut ion, The degree of a x i a l f l a t t en ing possible tends t o be inh ib i ted by t h e tendency of these l a r g e reactors toward f lux cycling due t o cyc l ic var ia t ions i n xenon poison concentration, Such tendencies a r e a lso sharply dependent on t h e temperature coefgicients of r eac t iv i ty , and appl icat ion of f?ux f1atter;ling could probably be accomplished only i n those reactor l oads ' ' which exhibit ne t negative temperature coef f ic ien ts of r eac t iv i ty ,
The
This has been
R L 2 - 1 7 Page 12
1,4 C r i t i c a l M a s s and Fuel Handling Problems
Cri t ical m a s s considerations w i l l generate handling problems throughout t h e f u e l handling complex, can be contained has , however, been demonstrated at other si tes, which process f u l l y enriched fue l so
That these problems
A s igni f icant a l lev ia t ing feature i n t h e processing of fu l ly enriched fue ls is the reduced volumetric throughput rate, With p a r t i a l l y enriched uranium fue l s , severely low c r i t i c a l masses are accompanied by high mass and volume throughputs, since only modest f u e l exposures are sought, Fully enriched fue l s w i l l o rd inar i ly be aimed t o achieve a high percentage burnout of f i s s i le material., which w i l l g rea t ly reduce throughput. For example, using t h e crude rule of thumb t h a t t h e f issi le material input concentration w i l l remain t h e same as i n metal l ic fue l , but t h a t 50 percent burnout w i l l be achieved, t h e residence t i m e of the f u e l w i l l be increased by a fac tor of about three. F'urther the m a s s of fuel handled w i l l be reduced by roughly an order of magnitude due t o the removal of t h e uranium-238, changes w i l l not eliminate the f u e l handling problems, they w i l l become more manageable,
While these
It is conceivable t h a t spec ia l fuel-target designs can be u t i l i z e d t o reduce the c r i t i c a l mass management problem throughout much of t h e f u e l cycle, By building an in t eg ra l driver-target fue l , the c r i t i c a l values can be reduced t o those of common, metal l ic uranium f u e l u n t i l i t is necessary t o separate t he components. probably be accomplished j u s t before dr iver reprocessing, eliminating c r i t i c a l m a s s problems inherent i n t he w e t s torage of irradiated fuel. Many t a rge t systems w i l l not be amenable t o t h i s treatment, which is , therefore , not a Cure-alJ., It is undoubtedly t r u e t h a t considerable equipment design w i l l be required t o provide geometrically favorable arrays and handling equipment,
This could
I . * 2, Reprocessing Plant
The scope and objectives of s tudies of t h e reprocessing plant are by-and- l a rge implici t i n those outlined f o r t h e reactors. However, i n order t o prevent a s e r i a l i z a t i o n of t h e e n t i r e mission, o r " ta i lgat ing" the reactor s tudies by t h e reprocessing s tudies , it is necessary t h a t t h e la t ter be approached first from t h e standpoint of inherent plant and process l i m i t a - t ions , r a the r than "go - no go" s tudies of reactor fuel ing proposals,
R L 2 - 1 7 Page 13
Two maJor considerations guide the study of reprocessing plant capabi l i t i es ; ( a ) Chemical compatibility of possible fue ls with t h e plant flow sheet, and (b) C r i t i c a l i t y prevention a t a l l points 'in t h e stream,
Basic technology f o r reprocessing a l imited number of types of f u e l ' i s avai lable; these include metal l ic uranium clad with aluminum or zirconium, uranium oxide, thor ia , mixed oxides, etc,
It is qui te possible t h a t f u e l models developed t o sa t i s fy reactor require- ments w i l l f i t within one of these reprocessing schemes, f u e l investigations w i l l be guided by the avai lable reprocessing technology so tha t , i f possible, an extensive research program io t h i s area 'can be avoi de de
I n i t i a l reac tor
PROGRESS DURING REPORT PER1 2 Highly enriched fue l processing has only recently been designated a separate research and development mission, t o begin i n m-1967; thus, no progress toward the goals and objectives out l ined has been made during FY-1966,
Work spec i f ic t o t h i s mission is scheduled
BUDGET PERIOD PLANS AND EXPECTED RESULTS
1, FY-1967
1,l Reactors
Work during FY-1967 w i l l be first directed toward establ ishing the basic f e a s i b i l i t y of reactor operation w i t h f u l l y enriched fue l , and ident i fying spec i f ic l imi ta t ions imposed by the inherent reactor design charac te r i s t ics , As such, t h i s work w i l l be i n i t i a l l y con- f ined to .phys ic s s tudies , assumptions out l ined under Scope and Objectives', designing b series of reactor fuels , varying (a ) the f i s s i le material concentration, while compensating the r eac t iv i ty changes w i t h added target con- centrat ion, (b ) t h e fuel ing geometry, from in tegra l elements containing both fue l and t a rge t t o nsupercell" arrays where fuel and t a rge t s are placed i n separate channels, and ( c ) f u e l burnouto Calculations w i l l be performed t o determine I ( a ) Reactivity changes with fue l exposure, and accompanying fue l channel r e l a t i v e power changes, i f "continuous" charging i s assumed,
These s tudies w i l l follow t h e basic
(b) Reactivity
R L 2 - 1 7 Page 14
changes with reactor temperaturep ( c ) Reactivity changes upon loss of coolant. systems, t h e i r a b i l i t y t o control t h e reactor under all normal and accident conditions , or t h e amount of addi t ional control capacity required, (e ) The k ine t ic response of t h e reactors ' t o t he m a x i m u m credible pos i t ive r eac t iv i ty ramps, assuming ,
t h a t avai lable control systems are scrammed, and any required engineered safeguards t o l i m i t the m a x i m u m credible ramp. Assuming tha t t he i n i t i a l physics s tudies reveal a feas ib le method of u t i l i z i n g f u l l y enriched fue l s , s tudies i n the la t ter p a r t of FY-1967 w i l l be d i rec ted toward defining a fue l material which w i l l s a t i s f y reactor c r i t e r i a with regard t o i r r ad ia t ion behavior, f i s s ion product confinement, and f ab r i cab i l i t y , As t h i s work progresses, coordination with t h e reprocessing s tudies w i l l permit t he adaptation of ex is t ing technology t o the proposed fue ls , o r development of new technology,
(d ) The worth of t h e avai lable reactor control
1 2 Reprocess i n 4 +
Work during FY-1967 w i l l be first directed toward determining t h e inherent l i m i t s of t he ex is t ing separations plant i n t h e processing of f u l l y enriched fue l , Fuel models w i l l not ye t have been defined by reactor s tudies; i n i t i a l reprocessing s tudies w i l l therefore be la rge ly confined t o : (a) assessing the c r i t i c a l i t y r e s t r i c t i o n s , based on a series of assumed f i ss i le material throughput assumptions, ( b ) of exis t ing flow sheets, thus providing assis tance t o reac tor s tudies of possible f u e l materials choices, ( c ) determining plant capab i l i t i e s i n those posi t ions where such capac i t ies w i l l be r e l a t ive ly independent of t h e fue l models f i n a l l y chosen, i o e o , reduction of separated products t o f i n a l form, separation of products after dissolut ion, and (a) determining t h e extent by which t h e plant capacity may be enhanced through modifications t o provide c r i t i c a l l y safe equipment, improved instrumentation systems, and other engineered safeguards.,
making preliminary s tudies
2, FY-1968 '
2,1 Reactors
1
Reactor fuel ing systems which w i l l develop from t h e FY4967 studies w i l l probably be qu i t e dissimilar from any previous reactor loadings, While the development of sophis t icated ana ly t ica l models has enhanced confidence i n the accuracy of calculated r eac t iv i ty parameters, it
R L 2 - 1 7 Page 15
w i l l be necessary t h a t some experimental ver i f ica t ion be obtained. One o r two loading schemes w i l l be chosen which w i l l provide "landmark" tests of t he ana ly t ica l techniques Experiments w i l l be performed t o determine the l a t t i c e parameters of t h e loading koO, thermal u t i l i za t ion , and f lux d is t r ibu t ion , and t h e worth of t he reactor control systems i n these loadings, these tests, in-reactor i r rad ia t ions w i l l be performed t o ver i fy fue l performance charac te r i s t ics , fue l burnout predictions, and l a t t i c e productivity, materials and fabricat ing fue l and t a rge t s , w i l l consume most of fl-1968; t e s t i n g w i l l be i n i t i a t e d l a t e i n IT-1968,
Design of the prototypic tes ts should be completed by March 1968, It is assumed t h a t a ful l - reactor i r r ad ia t ion mission, i d e n t i e i n g a spec i f ic product and quantity, w i l l have been defined, on these data, design of a f u l l reactor i r r ad ia t ion campaign w i l l be i n i t i a t e d , This design w i l l be completed by mid FY-1969, and a Hazards S&ary Report w i l l be prepared by about March 1969, The re su l t s of t h e prototypical i r rad ia t ions , which w i l l be started i n about June 1968, w i l l become avai lable about June 1969, thus providing t h e experimental ver i f ica t ion required fo r ful l - s ca l e loading approvalo
I n p a r a l l e l with
Planning and design of these tests, obtaining
Based
2,2 Reprocessing
A spectrum of reactor f u e l candidates will have been obtained during ~'Y-1967, during FY-1968 f o r these fuels. equipment and materials requirements t o provide the required plant capacity will be defined, P i lo t plant investigations t o demonstrate flow sheet f l e x i b i l i t y w i l l be performed, By t h e end of E-1968 su f f i c i en t d e t a i l w i l l be developed t o permit budget s tudies fo r f a c i l i t y modifications and additions ,
Separations process d e t a i l s w i l l be investigated i n depth As flow sheets are developed,
R L 2-17 Page 15a
S ta t i s t i ca l Summary Schedule
Fiscal. Year I
1
Dollars ( i n Thousands )
General Electric (N-Reactor)
Douglas United Nuclear
I sochem
Total-
M a n Years
General Electric (N-Reactor)
Douglas United Nuclear
Isochem
Tot a1
Equipment
General Electric (N-Reactor)
Douglas United Nuclear . I
Isochem
T o t a l
35 35 100 175 250
30 loo 140 170 250
158 191 200 250 300
223 326 440 595 8oo
05 0 5
0 163
1.6 2,2
2,1 4.0
- -
0 0
5 10
0 0 -I
5 10
i
R l C H L A N D RESEARCH AND DEVELOPMENT STATISTICAL SUMMARY
SlON NO. AND T I T L E 2. B U D G E T ACTIV ITY NO.
CON T R AC TO R
5 Y E A R CURRENT BUDGET YEARS * FY IS 67 FY 1s 68 FY 1969 FY is 70 FY 1 9 7 1 T O T A L
5. MAN-YEARS
A) SCIENTIF IC AND TECHNICAL
B) O T H E R
T O T A L
6. COSTS ($000’5)
AI D I R E C T L A B O R (INCL. COST O F SERV.
B)
C ) SPONSORED WITH BNW
MATERIALS, TECH. SERV., SUB CONTRACTS
T O T A L DIRECT
D) I N D I R E C T EXPENSES
T O T A L OPERATING COSTS
E) EQUIPMENT (NEW) I
04 0 4
01 01
14 13 e , 17 55 - 93 30 28
5 ‘I
F) IRRADIATION UNIT COSTS
7. COMMENTS:
.
D A T E W o S, Nechodom * AGREES WITH SCHEDULE 189 PREPARED BY
R L 2 - 1 7 Page 17
GENERAL ELECTRIC COMPANY
MISSION NO, 17, HIGHLY ENRICHED FUEL PROGRAM
SCOPE AND OBJECTIVES
The objective of t h i s mission i s t o devise one o r a few highly enriched f u e l models fo r t h e N-Reactor which w i l l serve as neutron sources for t h e i r r ad ia t ion of a broad spectrum of ta rge t materials, Studies w i l l be made of t he use of U-235 and plutonium isotopic mixes as fuel, Density) w i l l undertake t o define t h e f rac t ion of t he reactor i r r ad ia t ion space which m u s t contain fue l , it w i l l be i n i t i a l l y assumed t h a t one-third of t h e space must be reserved f o r t a rge t material, and t h a t t h e remaining two-third must operate t h e reactor a t ful l design power,
Since another program (High Power
The scope of t h i s study w i l l include:
a) Physics Behavior' of highly enriched fue ls
1) "Prompt" k ine t ic e f f ec t s 2) Total r eac t iv i ty e f f ec t s and control system worths 3) Transient r eac t iv i ty e f fec ts
b ) Fuel Materials
c ) Reactor Effects
1) Moderator temperature 2) Zirconium tube temperature
d) Fuel Handling - c r i t i c a l mass considerations,
1,l Physics Behavior
l o X O l 'lPrompt" Kinetic Effects '
The N-Reactor Mark I fue l load possesses'& prompt ne ia t ive temperature coeff ic ient of r eac t iv i ty , made up of a doppler e f fec t due t o fuel heating, and a net r eac t iv i ty defect due to coolant heating, 'The reactor power i s therefore self-damping, and accidental excursions are l imited by t h i s e f f e c t e fuel loads would not possess a doppler defect, However, the negative coolant temperature coeff ic ient of r eac t iv i ty would be retained i n most models, I n addition, t h e major f ac to r
Fully enriched
R L 2 - 1 7 Page 18
contributing t o adequate reac tor control is t he existence of delayed neutrons, which minimize t h e reac tor power response t o small changes i n r eac t iv i ty , While t h e k ine t ic behavior of a l l proposed models must be evaluated, it is l i k e l y t h a t t he combination of negative coolant temperature coeff ic ient of r eac t iv i ty , delayed neutron fract ion, sens i t ive scram instru- ,
mentation, and accelerated safe ty rods w i l l provide adequate reactor control
10102 Total Reactivity Effects and Control System Worth
The worth of t he N-Reactor primary control system is limited t o 6-8 percent k, depending on t h e f u e l model and reactor condition, The secondary backup system contributes an addi t ional 5-7 percent ko These systems must control the r eac t iv i ty swings of s t a r tup t r ans i en t s , including xenon poison buildup, leaving l i t t l e capacity f o r control l ing f u e l burnup r eac t iv i ty t r ans i en t so Studies of f u e l loads which w i l l be subJect t o la rge burnup t rans ien ts w i l l include evaluation of t h e need f o r supplementary control. devices , incremental f u e l charging t o minimize t h e r eac t iv i ty swing, arranging the target isotopes as burnable poisons and intermit tent t a rge t element replacement w i t h lower cross sect ion materials.
1,103 Transient Reactivity Wfec t s
Reactor s t a r tup net r eac t iv i ty t rans ien ts and t h e r eac t iv i ty t rans ien t associated with coolant loss can be e i ther negative o r posi t ive, depending on the loading model, The t r ans i en t s can be adjusted t o some extent by t a r g e t design and placement, control system must compensate f o r both t rans ien t and f u e l burnup r eac t iv i ty e f fec ts , these problems in te rac t and t h e t o t a l r eac t iv i ty swing determines the required control capacity,
Since the
102 Fuel Materials
The jdesign of t h e N-Reactor, u t i l i z i n g the confinement concipt of f i s s i o n product release control., requires t h a t t h e f u e l element perform as a f i s s ion product container at least through the l i m i t e d t i m e period while the primary loop is undergoing depressurization following coolant supply f a i lu re . The current, conservative, hydraulic analysis of t h i s accident indicates t ha t f i s s ion products m u s t be re ta ined within the fuel. f o r at l e a s t 220 seconds after coolant is l o s t , I n the case of t he Mask I fue l , t h i s is' accomplished automatically, s ince t rans ien t analyses indicate t h a t
R L 2 - 1 7 Page 19
the f u e l does not exceed t h e uranium melting point u n t i l at least 220 seconds have elapsed, A similar resis tance t o gross fuel failures w i l l have t o be provided f o r other f u e l models., This could again be accomplished by choosing a high melting point f u e l matrix, o r by insuring t h a t t h e Fuel cladding w i l l remain in t ac t f o r t h e required t i m e ,
,
1,3 Reactor Effects i
1,3,1 Moderator Effects
The N-Reactor graphite temperature, when the reactor is loaded with Mark I fue l , ranges up t o 1150 F, The m a x i m u m graphite temperature is l imited primarily by considerations of graphite burnout rates i n the presence of oxidizer agents, and t h e l i m i t i s therefore a function of t h e concentration of these agents, It is found t o be necessary t o maintain a ce r t a in minimum concentration of water vapor i n t h e reactor atmosphere t o prevent hydriding of t he zirconium process tube; t h i s concentration i n t u r n establ ishes a graphi te temperakure l i m i t The current practical. l imi t ing case occurs w i t h a water vapor concentration between 200 and 600 ppm, a zirconium tube temperature l i m i t of 700 F, and a graphite temperature l i m i t of about 1550 F,
The graphite temperature i s a function of t he heat generation rate i n t h e moderator, and the conductivity of t h e heat t r ans fe r . paths from t h e graphite t o t h e reactor coolant streams, The con- duct ivi ty is not upwardly adjustable; helium is used as a heat t r ans fe r medium, and possesses the best conductivity of available gaseous mediums, not lowered, by modiqing the gas composition, rate i n the graphi te is a function of reactor (ac tua l ly regional) power, and of t he f u e l composition, Heat is generated by moderation of neutrons and by capture of gamma radiat ion, i n f u e l models can r e su l t i n changing the r a t i o of water t o graphite 'volume, thus changing the r e l a t i v e amowit of neutrgn moderation, and can a l s o r e s u l t i n changing t h e r e l a t ive amount of gamma radiat ion absorbed i n t h e fue l and moderatoro s ince par t of t he gamma rad ia t ion results from radia t ive neutron capture i n t a rge t materials, t he t o t a l gamma energy generation may change as a re su l t of changing fue l models. I n general, it i s found that lighter f u e l weights result i n higher graphite temperatures, while changing the target material from U-238 t o an alpha-particle- producing material w i l l reduce graphite temperatures Each highly enriched fue l model w i l l therefore have t o be evaluated i n combination
The graphi te temperature could be raised, but The heat generation
Changes
In addition,
w i t h several t a rge t materials, t o establish a range of heat generation rates as a function of power leve l ,
R L 2 - 1 7 Page 20
The most d i rec t method of changing t h e maximum graphite temperature with any f u e l model is t o adjust t h e m a x i m u m loca l power l e v e l o of enriched, "spike," fue ls , o r by control rod manipulation, It has been found t h a t the Mark I f u e l load can readi ly be operated with a r ad ia l peak-to-average tube power r a t i o of about l,25, However, it is found tha t attempts t o reduce t h e - axial r a t i o t o t h i s va lue ' r e su l t i n xenon cycling i n s t a b i l i t i e s , Thus, t h e axial r a t i o i s maintained at about 1,4 o r greater, resu l t ing i n an overa l l peak-to-average power r a t i o of greater than 10-750 The xenon cycling constraint w i l l not necessarily apply t o all loadings; those which exhibit strong negative temperature coef f ic ien ts of r eac t iv i ty w i l l not cycle, and may be f l a t t ened as much as economically feasible , An overa l l r a t i o of about 1,55 o r less may be possible, fuel loads w i l l therefore be biased i n t h e direct ion of negative r eac t iv i ty coef f ic ien ts i n order t o es tab l i sh t h i s f l e x i b i l i t y .
This may be accomplished by se lec t ive placement
Designs of fu l ly enriched
1,3.2 Zirconium Tube Temperature Effects
The N-Reactor zirconium process tube outer surface temperature m u s t be controlled below a l i m i t of about 700 F t o prevent zirconium hydriding. The maximum tube surface temperature with Mark I fue l ranges up t o 660 F , being establ ished by t h e operating coolant temperature and t h e rate of heat t r ans fe r from t h e moderator t o t h e coolant stream (and hence, t h e heat generation r a t e i n the graphi te) , generation rate i n t h e graphite, t h e discussion i n t h e preceding section a l so applies t o zirconium temperature control, the current operating point is found t o be c loser t o t h e technical l i m i t , I n addition t o attempts t o control within the ex is t ing l i m i t , it may be necessary t o undertake s tudies designed t o in- crease the l i m i t , either by a more accurate and precise experimental programe o r by devising inh ib i t ing reactor atmospheres which permit operat ion at higher temperatures c.
Since the tube temperature is a function of heat
Here, however,
1,4 Fuel' Handiing - C r i t i c a l Mass Considerations i
While l i t t l e o r no new research w i l l be required, application of ex is t ing c r i t i c a l mass technology t o the present plant process w i l l require an extensive and thorough design and equipment development e f fo r t
R L 2 - 1 7 Page 21
Current N-Reactor c r i t i c a l i t y control c r i t e r i a require t h a t all f i s s i l e mater ia l arrays be assumed optimumly moderated, f o r purposes of evaluating c r i t i c a l i t y poten t ia l , Th i s is f eas ib l e w i t h t h e r e l a t ive ly low f u e l enrichments which have been u t i l i zed , However, experience indicates tha t fue l handling l i m i t s w i l l become excessively r e s t r i c t i v e at enrichments i n excess of 2-3 w/o U-2350 It w i l l be necessary tha t c red i t be taken f o r t he absence of moderator, and tha t t he process be designed w i t h t h i s end i n mind, i n order t o e f f i c i e n t l y process f u l l y enriched fuel . This i s current ly standard p rac t i ce at other sites,
C r i t i c a l mass problems w i l l be p a r t i a l l y a l l ev ia t ed when processing f u l l y enriched fue l by the reduced volume of f u e l requiredo f u e l w i l l be i r r ad ia t ed t o much higher f i s s i l e mater ia l burnups than is t h e current prac t ice with meta l l ic uranium fuels . be expected t h a t t h e f u l l y enriched f u e l volume throughput w i l l be roughly a f ac to r of t h ree l e s s than t h a t o f current f u e l models, In addition, 'the m a s s of f i s s i l e f u e l handled w i l l be reduced by roughly an order of magnitude due t o t h e removal of t h e U-238, Both of these fac tors w i l l tend t o make eas i e r t he solut ions t o c r i t i c a l mass handling problems
The most d i f f i c u l t problem foreseen, from the standpoint of modifying ex is t ing systems (it is assumed a d i f fe ren t fue l fabr ica t ion plant w i l l be u t i l i z e d ) is the i r r ad ia t ed f u e l handling and s torage system, batcheso The ex is t ing f u e l handling and storage arrangement is incompatible w i t h e f f i c i e n t handling of f u l l y enriched fue ls , system w i l l have t o be modified t o handle t h e discharged f u e l more rapidly, i n s m a l l quant i t ies or a continuous stream, and t o package and s t o r e t h e f u e l i n safe grrrxys, A system design which i s s a f e by design and geometry, not by procedure, i s considered a necessi ty where a moderating medium i a always presento
Such
It can
For reac tor eff ic iency, f u e l m u s t be discharged i n l a r g e
The
I 9 PROGRESS DURING I O F t P PE-RIOD
Mission 17 is now being i n i t i a t e d , and has progressed only i n t o t h e plarmiw s t a t e .
R L 2 - 1 7 Page 22
BUWET PERIOD PLANS AND EXPECTED RESULTS
1,
The bas ic physics f e a s i b i l i t y of N-Reactor operation with f u l l y I
enriched fue l s w i l l be es tabl ished, Parametric calculat ions w i l l be performed, varying: (a) f i s s i l e material concentration, (b) fuel element geometry, from i n t e g r a l elements containing both f u e l and t a r g e t t o "supercell" arrays, and ( c ) fue l burnout, w i l l determine: with f u e l burnup, ( 2 ) r e a c t i v i t y changes with reactor temperature and power, (3) r e a c t i v i t y changes w i t h coolant loss, (4 ) worth of t h e avai lable cont ro l systems, and amount of required supplemental control , and ( 5 ) t h e k ine t i c response of t h e reac tor t o rapid r eac t iv i ty inser t ions ,
The calculat ions (1) r e a c t i v i t y and f u e l r e l a t i v e power var ia t ions
Assuming t h a t f eas ib l e methods of f u l l y enriched fue l ing a r e defined, s tud ies i n t h e l a t , t e r pa r t of FY-1967 w i l l be directed toward defining f u e l materi-al(s) which w i l l satisfy c r i t e r i a f o r i r r ad ia t ion behavior, f i s s ion product conf'inement , and f ab r i cab i l i t y , Canversely, incentives may be revealed f o r engineering modifications t o t h e N-Reactor confine- ment system t o make it compatible with an otherwise des i rab le fuel model, Coordinatim.with the reprocessing s tudies is an essential feature of t h i s effart ,
2. FY-1968
Reactor fue l ing systems which w i l l develop from t h e FY-1967 s tudies w i l l probably be q u i t e d i ss imi la r from any previous reactor loadings, While the development of sophis t ica ted ana ly t i ca l models has enhanced confidence i n t h e accuracy of calculated r e a c t i v i t y parameters, it w i l l be necessary tha t some experimental ve r i f i ca t ion be obtained, One o r two load- schemes w i l l be chosen which will provide "landmark" tests of the ana ly t i ca l techniques, t h e l a t t i c e parameters of t h e loading - koo, thermal u t i l i z a t i o n , and f lux d i s t r ibu t ion , and t h e worth of t he reac tor control systems i n these loadings, I n p a r a l l e l w i t h these t e s t s , in-reactor i r r ad ia t ions w i l l be performed t o ver i fy f u e l performance cha rac t e r i s t i c s , f u e l burnout predictidns, '&nd l a t t i c e product ivi ty , tests, obtaining mater ia ls and fabr ica t ing f u e l and t a r g e t s , w i l l consume most of m-1968; t e s t i n g w i l l be i n i t i a t e d l a t e i n M-1968,
Experiments w i l l be performed t o determine
Planning andl design of : these
R L 2 - 1 7 Page 23
Design of t h e prototypic t e s t s should be completed by March 1968. It is assumed t h a t a full-reactor i r rad ia t ion mission, identirying a spec i f ic product and quantity, w i l l have been defined during FY-1.967. Based on these data , design of a f u l l reactor i r r ad ia t ion campaign w i l l be i n i t i a t e d , 1
and a Hazards Summary Report w i l l be prepared by about March 1969. The results of the prototypical i r rad ia t ions , which w i l l be s t a r t e d i n about March 1968, w i l l become available about June 1969, thus providing the experimental ver i f ica t ion required for ful l -scale loading approval
This design w i l l be completed by mid m-1969,
i
L
9
Schedule of Activ i t ies
Establish feas ib le reactor loading design,
Define fue l material,
Establish a full-reactor target irradiation mission,
I n i t i a t e PCTR testso
Complete PCTR tests,
In i t ia te prototype reactor irradiations,
Complete full-reactot 'test design,
Complete f'ull-reactor hazards summary report e
Complete and analyze prototype irradiations ,,
Obtain fu l l - sca le loading approval,
i
R L 2 - 1 7 Page 24
January 1967
March 1967
June 1967
September 1967
March 1968
March 1968
January 1969
March 1969
June 1969
July 1969
' ?
. RL-401 ( 8 -66) APC-IL RICWLAMD. W1SH
U. S. ATOMIC ENERGY COMMISSION
RICH L A N D I RESEARCH AND DEVELOPMENT STATISTICAL SUMMARY 02 PROGRAM HAN- OPERATIONS O F F I C E
1- MISSION NO. AND TITLE No, 17, H i g h l y Enriched Fuel Program 2. B U D G E T ACTIVITY NO.
MISSION SUB-PROGRAM T I T L E
3. CONTRACTOR AND A. Douglas United Nuclearp Inca AT( 45-1 I1857 4. PERSON IN CHARGE: To W, Ambrose R , W, Reid
_11-
E. I N V E S T I G A T O R CONTRACTOR NUMBER C. INVEST1 G A T 0 R
F I V E Y E A R F O R E C A S T CURRENT BUDGET YEARS *
5 Y E A R FY is 67 FY 19 68 FY 1s 69 FY 1s 70 FY 19 71 T O T A L
5. MAN-YEARS
A) SCIENTIFIC AND TECHNICAL 0 l o 3 E ) O T H E R 0 0
0 103 T O T A L
6. COSTS (IOOO'S) s $ b $ $ s A) DIRECT LABOR (INCL. COST O F SERV. 0 26 E ) MATERIALS, TECH. SERV., SUB CONTRACTS 12 C) SPONSORED WITH BNW 28 50 70 90 120
T O T A L DIRECT 28 80 0 ) I N D I R E C T EXPENSES 2 12
T O T A L OPERATING COSTS
E ) EQUIPMENT (NEW) - 30 100 140 170 250
5 10
F) I R R A D l A T t O N UNIT COSTS
7. COMMENTS:
_. ..
D A T E G o Fe Wsley * AGREES WITH SCHEDULE 189 PREPARED BY
R L 2 - 1 7 Page 26
DOUGLAS UNITED NUCLEAR
MISSION NO, 17. HIGHLY ENRICHED FUEL PROGRAM
SCOPE AND OBJECTIVE
Pr ior t o 1955 highly enriched fue l i n the form of uranium-aluminum al loy was used as spike enrichment and as reac t iv i ty support f o r a tritium production load, of unresolved questions concerning safe ty of the loadings and economic disadvantages e
The use of highly enriched fue l was discontinued because
The objective of t h i s mission i s t o invest igate t h e safe ty and economic f e a s i b i l i t y of f u l l and p a r t i a l highly enriched fuel-target loadings i n t h e f ive production reactors operated by Douglas United Nuclear, Inc, The ava i l ab i l i t y of high speed computer codes, improved experimental methods and f ac i l i t i e s ' , and proven calculat ional techniques w i l l allow resolution of t he questions on the safe ty of highly enriched fuel-target loadings
The stimulus f o r such loadings is t h a t e f fec t ive ly no Pu-239 is produced and a l l production can be diverted t o other isotopes,
The predominate charac te r i s t ic of t h e highly enriched fuels which requires sa fe ty analysis is t h e lack of U-238, which r e su l t s i n decreased negative doppler Soeff ic ient and the l o w melting temperatures of uranium-aluminum al loys ,
PROGRESS DURING RECENT PERIOD
This mission has j u s t been established, and hence, there is no progress t o report
BUDGET PERIOD PLANS AND EXPECTED RESULTS
Reactor engineering and physics analysis computer codes w i l l be used t o - determine f u e l design and operating charac te r i s t ics , l a t t i c e physics parameters, temperature coef f ic ien ts , water l o s s reac t iv i ty , and c r i t i c a l mass data, Reactor k ine t ic codes w i l l be used t o determine operating safety limits and melting t i m e s especially as concerned w i t h "speed-of-control''
R L 2 7 1 7 Page 27
and "confinement" c r i t e r i a . Modifi st ions necessary i n safe ty system, fuel fabr ica t ion f a c i l i t i e s and handling, storage, and shipment f a c i l i t i e s w i l l be determinedo Calculated y ie lds and operating chazac ter i s t ics w i l l be used t o obtain preliminary cost estimates, From t h e calculated data a preliminary Hazards Summary Report 'can be prepared, I
If t h e preliminary Hazards'Summary Report shows f e a s i b i l i t y of highly enriched f u e l , experiments w i l l be performed t o confirm calculat ions and in-reactor t e s t loads w i l l be designed t o obtain data on f u e l element and t a r g e t fabr icat ion and separation, reac tor performance, and yieJd data , These da ta w i l l then be used t o design full and p a r t i a l reac tor loadings
,
i
R L 2 - 1 7 Page 28
Schedule of Activit ies
Calculate reactor physics and engineering parameters - January 1961 DUN and BNWo
Determine modification8 t o exist ing f a c i l i t i e s - DUN, Preliminary Hazards Summary Repart - DU3,
March 1967
June 1967
Physics and Engineering tests - BNW. In-reactor demonstration load design and irradiation - DUN, March 1968
September 1967
Separations and analysis - Isochem.
Define f u l l and partial reactor load parmeters - DUN. June 1968
July 1968
i
4, . . . . .
I I RL-401 ( 9 - 6 0 ) AX-I1L RICHLIWD. WASH U. 5. ATOMIC ENERGY COMMISSION
RlCHL AND I RESEARCH AND DEVELOPMENT STATISTICAL SUMMARY
I . MISSION NO. AND T I T L E No, 17, Highly Enriched Fuel Program 2. B U D G E T A C T I V I T Y NO.
MISSION SUB-PROGRAM T I T L E
3. CONTRACTOR AND Isochem, Inc, AT(45-1)1851 4. PERSON I N C H A R G E A-
INVEST1 G A T 0 R
NUMBER C. I N V E S T I G A T O R
CONTRACTOR B.
F I V E Y E A R F O R E C A S T
5. MAN-YEARS
A) SCIENTIFIC AND TECHNICAL
B) OTHER
T O T A L
6. COSTS (BOOO'S)
A) DIRECT L A B O R (INCL, COST O F SERV. )
8 ) MATERIALS, TECH. SERV., SUB CONTRACTS
C l SPONSORED WITH BNW
T O T A L DIRECT
D) INDIRECT EXPENSES
T O T A L OPERATING COSTS
E) EQUIPMENT (NEW)
F) IRRADIATION U N I T COSTS
CURRENT BUDGET YEARS * 5 Y E A R
FY 19 67 FY 19 68 F Y 19 69 FY 19 70 FY 19 71 T O T A L
1,6 202 0 s B B I I
19 27 25 45
LO3 100 135 175 200
147 17 5
P.
7. COMMENTS:
.. .r.
g!a c (D m r
D A T E H, C b Rathvon * AOREES WITH SCHEDULE 189 PREPARED BY
R L 2 - 1 7 Page 30
ISOCHEM, I N C ,
MISSION NO, 17, HIGHLY ENRICHED JWEL PROGRAM
SCOPE AND OBJECTIVES
Detailed separations requirements f o r reprocessing f u l l y enriched fue ls o r plutonium fuel I w i l l depend on: dimensions, U-235 or (Pu) content; and (b) The throughput requirement, Presumably, incentives f o r maximum throughput w i l l be great i n order t o assure accommodation of other Hanford reprocessing programs, Thus, t he general requirements t h a t govern the de ta i led scope and objectives include t h i s assumptiono
(a) The fuel description---materials,
The scope of the Isochem program w i l l include the chemical reprocessing s teps commencing with the receipt of i r r ad ia t ed f u e l s D basic separations processing , t h e product streams including neptunium values, i f any, w i l l be pur i f ied and t h e waste streams w i l l be disposed of i n accordance with t h e requirements of acceptable waste management pract ikeso w i l l be converted t o a form acceptable fo r recycle at Hanford, o r for any other indicated use,
Following the
The UNH ( o r Pu n i t r a t e ) product from the separations process
The Isochem research and development objectives for such a program would be t o provide the technology required by the program as scoped. technology f o r reprocessing cer ta in types of these fue l s is now available. The avai lable technology and experience would provide a basis f o r the R&D required f o r application at Hanford, Because of the current s t a tus of Kanford reactor and f u e l technology, it is ant ic ipated tha t i n i t i a l programs would stress engineering s tudies of separations processes t h a t would complement s tudies by General E lec t r i c and DUN,
Basic
During t h i s period, laboratory R&D e f fo r t s would be of a preliminary exploratory nature e f fo r t would be expected t o accelerate , as the program needs become better defined,
Subsequentlyo a t about m i d FY-1967, t h e laboratory
P
PLANS AND EXPECTED RESULTS
'During FY-1967, R&D e f f o r t s w i l l i n i t i a l l y include engineering s tudies t o define the most promising approaches f o r Isochem reprocessing e f fo r t s Processes consis tent w i t h t h e program scope and objectives
R L 2 - 1 7 Page 31
would be included. To t h e extent compatible w i t h reactor s tud ies , preliminary s tudies of (a) flow sheets , (b) c r i t i c a l i t y requirements ( c ) sa fe ty , (d ) equipment design, and ( e ) mater ia ls would be included. Laboratory programs w i l l i n i t i a l l y be exploratory i n nature. s tudies w i l l be concerned with (a) dissolut ion of cwceptuttl f u e l types and mater ia ls ; (b) dissolver concepts; ( c ) reduction of the ,
separated products t o a useful form; and (d ) plant instrumentation requirements,
These '
As t h e r e s u l t s of the engineering s tudies become available, de f in i t i on of the supporting laboratory programs w i l l improve, and more spec i f i c program needs w i l l be indicated, process se lec t ion and flow sheet def in i t ion can be made f o r t h e simpler candidate f u e l forms; and the more complicated fue l materials w i l l require considerable laboratory and engineering support t o develop su i t ab le processes of high r e l i a b i l i t y and f l e x i b i l i t y ,
By the end of FY-1967, preliminary
During FY-1968, t h e b e t t e r de f in i t i on of f u e l elements w i l l permit t h e invest igat ion of separations process d e t a i l s i n depth, Flow sheet problem areas w i l l be invest igated and equipment and mater ia l needs w i l l be defined, Supporting p i l o t plant invest igat ions t o provide process demonstration f o r advanced f u e l forms w i l l be performed. By t h e end of a-1968, Isochem requirements f o r the Hanford program can be defined on a prac t icable , working bas i s , t o permit t h e preparation of budget-s tuaies t o support necessary planning f o r f a c i l i t y modifications and/or additions
3. Post FY-1968
Subsequent R&D programs w i l l be aimed at (a) process optimization, and (b) t he invest igat ion of promising a l t e rna t ive processes. Information necessary f o r design study and f a c i l i t i e s scoping w i l l be completed,
i ' I
R L 2 - 1 7 Page 32
Schedule of Activit ies
Establish process and flow sheet for DUN and GE primary fuel candidates e
Complete p i l o t plant process demonstration.
Complete planning.of f a c i l i t y modifications,
(Have modifications i n place)
September 1967
June 1968
January 1969
June 1970
i ' 9
\' I
2 -17 Pa,Ge 33
DISTRIBUTION
RICHLAND OPERATIONS OFFICE
1-23. D. G. Williams
DOUGLAS UMITED N u C m , IN!. _..
24. T. W. Ambrose 29. W. M. Mathis 25. P. A. Carleon 30. R. Nilson 26. R, G. Geier 31. G. F. Owsley 27. C. D. Harrington 32. R. W. Reid 28. C, W. Kuhlman 33. 0. C. Schroeder
34. J. T. Str inger
GENERAL ELECTRIC COMPANY - HAPD
35. R. L. Dickeman 41. W. S. Nechodom 36. D. W. Leiby 42. G. W. Nicholaus 37. M. C. Leverett 4 3 . 4 4 . J. W. Riches/D. L. Condotta 38. M. Lewis 45. R. H. Shoemaker 39. J, S. McMahon 46. R. E. Trumble 40, J. Milne
ISOCHEM INC.
47. R. F, Campbell 53. A. J. S c o t t 48. R. A. Connell 54. H. P. Shaw 49. J. B. Fecht 55. A. E. Smith 50. 0. F. H i l l 56,-57. R. E. Tomlinaon 51. J. N. Judy 58. J. H. Warren 52. P. E, Reed
BATTELLE -NORTHWEST , I !
59. F. W. Albaugh 60. C. A. Bennett 61. J. J. Cadwell 62, F, G. Dawaon 63. D. R. de Ealaa 64. E. A. Eechbach 65. J. J. Fuquay 66, H. H m t y
67. E. R. I r i b h 9
68, A. R. Keene
70. R. S. Paul 71. A. M. PLatt 72. C. A. Rohrmsnn 73. E. E. Volland 74. M. T. Walling, Jr. 75. E. C. Worlton
69. G. A. Last
