J.E.N.454Sp ISSN 0081-3397

STATUS OF IVO-FR2-Vg7- EXPERIMENTFOR IRRADIATION OF FAST REACTOR

FUEL RODS

by

Otero de la Gándara,J.L.*Kummerer, K.*Bojarsky ,K . *Elbel.H.»López Jiménez,J.*

Junta de Energía Nuclear, División de MetalurgiaKernforschungszentrum Karlsruhe, Instituí fürMaterial-und Festkorperfor schung.

JUNTA DE ENERGÍA NUCLEAR

MADRID,1979

CLASIFICACIÓN INIS Y DESCRIPTORES

B25; E23FUEL RODSIRRADIATIONFR-2 REACTORFAST REACTORS

Toda correspondencia en relación con este traba-jo debe dirigirse al Servicio de Documentación Bibliotecay Publicaciones, Junta de Energía Nuclear, Ciudad Uni-versitaria, Madrid-3, ESPAÑA.

Las solicitudes de ejemplares deben dirigirse aeste mismo Servicio.

Los descriptores se han seleccionado del Thesaurodel INIS para-describir las materias que contiene este in-forme con vistas a su recuperación. Para más detalles consultese el informe IAEA-INIS-12 (INIS: Manual de Indiaa-ción) y IAEA-INIS-13 (INIS: Thesauro) publicado por el Or-ganismo Internacional de Energía Atómica.

Se autoriza la reproducción de los resúmenes ana-líticos que aparecen en esta publicación.

Este trabajo se ha recibido para su impresión en

Julio de 1. 979.

Depósito legal n° M-28745-19"9 1.5.B.N. 84-500-3320-9

Informe sobre el Seminario celebrado en elCentro Nacional Juan Vigón CMadrid) el día21 de Septiembre de 1978, organizado porla Junta de Energía Nuclear en colaboracióncon el Centro de Investigaciones Nuclearesde Karlsruhe ( K F K ) , República Federal Alema-na, sobre el Programa de Irradiación conjuntode barras combustibles para reactores rá-pidos.

En el marco del Convenio de Colaboración existente entreel Centro de Investigaciones Nucleares de Karlsruhe (KFK)y la Junta de Energía Nuclear (JEN) se ha establecido unprograma de irradiación de combustibles para reactoresrápidos reproductores. En su etapa presente, este Progra-ma, comporta la irradiación de 12 varillas combustiblesde óxidos mixtos en el reactor experimental FR-2 deKarlsruhe.

Participan, por parte alemana, el Projekt SchnellerBrüter (PSB) y el Institut für Material-und Festkorper-forschung (IMF), y por parte española, la División deMetal urgí a.

Aunque el Seminario comprendió temas relacionados conel desarrollo e implantación de los reactores rápidosen el mundo, y el ciclo del combustible, este informerecoge aquellas intervenciones más directamente ligadasal Programa de Irradiación conjunto IV0-FR2-Vg7.

Contents

1. Presentation of the Metal 1 urgí cal División Activities(J.L. Otero de la Gándara, JEN).

2. Research and Development Work in Fast Reactor Fue!Development (K. Kummerer, KFK).

3. Possibilities for Fast Breeder Fue! Pins Irradiationin the Karlsruhe Nuclear Research Centre (E. Bojarsky,KFK) .

4. The Irradiation Experiment IV0-FR2-Vg7.(H. Elbel , KFK).

5. Design of IV0-FR2-Vg7-Experiment(J. López Jiménez, JEN).

PRESENTATION OF THE METALLURGICAL

DIVISIÓN ACTIVITIES

by

J.L. Otero de la GándaraJUNTA DE ENERGÍA NUCLEAR

División de Metalurgia

-1-

In 1951 was created the Junta de Energía Nuclear (JEN) -and four years later it started the construction of the NuclearCentre Juan V i g ó n , situated in the north-west periphery of M a d r i d ,Uni versi ty C a m p u s .

The objectives of the JEN were to supply the sufficienttechnological support to the Country in the field of Nuclear --Energy: Research and D e v e l o p m e n t of Nuclear R e a c t o r s , Fue! Cyclewith the creation from Pilot P l a n t s , Safety and Nuclear M e d i c i n e ,the production of I s o t o p e s , Basic R e s e a r c h s , e t c . , and the trai-ning of especialists in this área.

In this short introduction to our Seminar I will try topresent the principal activities of the Metallurgy División anda historial short view.

Between 1951 and 1 9 6 0 , the Metallurgy D i v i s i ó n , with thehelp of other Divisions of J E N , -especially the Materials Divi-sión- concentrated on materials and uranium compounds and uraniumm e t a l , and in addition to acquiring k n o w l e d g e , it trained person-nel in specific technical fields of Metallurgy related to nuclearmaterials as well as creating the infrastructure of experimentalmeans and s e r v i c e s . Between 1960 and 1970 it completed t e c h n o l o g ical studies which led to the manufacture of UOp pellets , uraniumcarbide and uranium metal ingots. Of the l a t t e r , 55 tons were prp_duced for the Vandellcis reactor as well as it was obtained an alu_minum urani um-oxide cermet of interest for manufacturing fue! ele_ments for research r e a c t o r s . In addition of these studies of rawm a t e r i a l s , it dedicated time to all stages of fue! element manu-facturing in the specific field of research r e a c t o r s , obtainingexperiences for manufacturing them with sufficient quality guaran_tee. With the help of an own fabrication p r o c e s s , it has been pp_ssible to supply fue! elements for the JEN-1 reactor as well asfor the Venezuelan and Chilean r e a c t o r s .

- 2 -

• a

General view of JEN

-3-

In the decade between 1970 and 1980, in which we find our.selves at present, the nuclear fuel program is directed towardsa closer cooperation with the National Energy Plan, which in theoriginal versión includes a tremendous increase (a factor of 10)in the installed nuclear power in the next 6-8 years. This programsould contribute to the increase of national participation in --the future nuclear power plants. In this regard, a work programhas been started in collaboration with ENUSA, which will permitthe manufacture of prototypes of light water fuel elements for -pressurized reactors as well as for boiling water reactors. Forthis purpose, a UOp pellet manufacturing plant, as well as over90% of the means required for the fuel element assembly line isalready available. The manufacture of these prototypes w.i 11 be -complemented by three-circuit f1uid-dynamic studies, which are -workable at present, one for ordina.ry pressurized water, anotherfor heat cycling and a third for testing on oversized scale mo-dels for studies with speed measurements and other equipment, --which due to this size, could distort current lines in standard-scale models.

Studies have been carried out in the Division's Hot Celis,in col 1aboration with Unión Eléctrica (Zorita Reactor) and Wes-tinghouse, regarding performance of the fuel elements used in -the said plant and it is hoped that the study, started 10 year-ago, will be completed in 1979.

With the assistance of General Electric and Nuclenor, --this year we hope to start a work program related to corrosión -problems in the reactor. We also hope that this will familiarizethe Division's technical personnel with the auxiliary techniqueswhich are required in observation, measuring, sample-taking offuel elements made simultaneously with the recharge operations -in the power reactor's operations ponds.

- 4 -

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Ti. SsSS !ri

%

11

\

Piant of fuel píate s manufacture

-5-

At Westinghouse's request, we will prepare irradiation -capsules of the steels to be used in carrying out performance --studies of those which make up the reactor's vessel , which willbe manufactured in Spain in the ENSA factory. In this same fieldof activtty, we hope within the next 2 years to increase the --services in the Metallurgy Division's Hot Cells in the Juan Vi--gón Centre in order to carry out Tensite strength, fatigue and -resilience tests which are required for studying property modifi_cations due to irradiation as a result of the performance of --steels used in the vessels.

At present, the Metallurgy División Organization is shownon the attached Qrganigram» which groups the different activitiesin clearly-defined sections.

With reference to the new Centre, which will probably beset-up in Soria, the Division's mission will be the establishmentof bases for three projects: 1) Hot Cells for examination and --evaluation of the heat elements in the National Energy Plan's po_wer reactors,2) Fue! element manuf acturing plant for the new cen_tre's research reactor and 3) Mixed oxides Laboratory.

Using the Division's means and services, design and cons_truction of the different equipment required for the work programhave been carried out. In addition, non-destructive test equipmenthas been built, especially in the field of ultrasound, automatedwelding equipment, auxiliary equipment for hot cells, etc.

At present time, a lot of concrete works are in cours, --for example Aluminium-clad graphite elements for JEN-1 and Chileanexperimental reactors; boral plates for the precisión power adjus_tment systems and for storage modules in the irradiation fue! ele_ment transport equipments, etc.

- 6 -

•

m

" . " • "

ü

-7 -

F i n a l l y , we w i s h t o p o i n t o u t t h a t s e r v i c e s t o c o m p a n i e sw i t h n o n - n u c l e a r a c t i v i t i e s and w h i c h r e q u i r e s t e p s o r s t u d i e s ,w i t h i n t h e D i v i s i o n ' s m e a n s , a r e e a s i l y c a r r i e d o u t .

I n t h e f a s t r e a c t o r f i e l d , we a r e t r y i n g t o c a r r y o u t - -s t u d i e s w i t h a n i n t e n s e a n d l i m i t e d s c o p e , by m e a n s o f p a r t i c i p a _t i n g i n i n t e r n a t i o n a l c o o p e r a t i o n p r o g r a m s . T h e w o r k c a r r i e d o u tb e t w e e n JEN a n d KfK a d a p t s i t s e l f t o t h e s e i d e a s . D u r i n g t h e l a s tf e w y e a r s , a g r e a t j o i n t e f f o r t h a s b e e n c a r r i e d o u t i n s t u d y i n gp e r f o r m a n c e o f s t e e l s f o r c l a d d i n g a n d s t r u c t u r a l m a t e r i a l s a g a i n s tS o d i u m a t t a c k i n t h e ML-1 a n d ML-2 c i r c u i t s , a n d t h e y w i 1 1 c o n t i -n u é i n t h e c o t n i n g y e a r s t o m o d i f y w h a t i s p r e s e n t l y c a l l e d t h e -M L - 3 c i r c u i t . T h e s t u d i e s a r e b a s i c a l l y c a r r i e d o u t w i t h r e g a r dt o c o r r o s i ó n a n d t o t h e m e c h a n i c a l p r o p e r t i e s 9 p a y i n g s p e c i a l - -a t t e n t i o n t o c r e e p a n d f a t i g u e t e s t s . T h e E n g i n e e r i n g a n d M e t a -l l u r g y D i v i s i o n s c o l l a b o r a t e i n t h i s w o r k .

W i t h i n t h e s c o p e o f t h e J E N - K f K c o l 1 a b o r a t i o n , t h e IVO -P r o j e c t - w h i c h w i l l b e d i s c u s s e d i n g r e a t e r d e t a i l d u r i n g t h i s -S e m i n a r - d e a l s w i t h s t u d i e s r e l a t e d t o i r r a d i a t i o n o f m i x e d o x i -d e f u e ! r o d s . T h i s a s s u m e s p r i o r t h e o r e t i c s t u d i e s t o d e f i n e e x -p e r i e n c e p l a n n i n g , m a n u f a c t u r e o f t h e c a p s u l e s a n d f e r t i l e m a t e -r i a l s » i r r a d i a t i o n a n d e v a l u a t i o n i n h o t c e l l s o f m a t e r i a l p e r -f o r m a n c e . W i t h o . u t g o i n g i n t o t h e p a r t i c u l a r a s p e c t o f t h e w o r k -p r o g r a m ' s s p e c i a l c h a r a c t e r i s t i e s , o r t h e c o n v e n i e n c e o f s e l e c -t i n g t h e b a s i c e x p e r i m e n t a l c o n d i t i o n s , we w i s h t o m e n t i o n i t s -c o n c r e t e i n t e r e s t f r o m t h e D i v i s i o n ' s p o i n t o f v i e w . T h i s e x p e r i _m e n t a l l o w s u s t o e x p e r i e n c e , f o r t h e f i r s t t i m e , w i t h d i r e c t -p a r t i c i p a t i o n i n a w o r k p r o g r a m , p l a n n i n g o f i r r a d i a t i o n e x p e -r i e n c e s a n d t h e m a n u f a c t u r e o f c a p s u l e s a s w e l l a s p r o b l e m s r e -l a t e d t o i r r a d i a t i o n i n s i d e a n e x p e r i m e n t a l r e a c t o r .

T h e d i f f e r e n t a s p e e t s o f i n t e r e s t i n f u e l e l e m e n t t e c h n o -l o g y , s u c h a s d e s i g n a n d p r e p a r a t i o n o f c o d e s f o r c a l c u l a t i o n -a n d t h e m a n u f a c t u r e o f p r o t o t y p e s a n d o f i r r a d i a t i o n c a p s u l e s , -

-8-

c h e G k i ng of performance i n the r e a c t o r , hot ce 11 measurements andevaluation of performance in subsequent comparative studies withthe design b a s i c s , give us a complete picture of the problem. -The study of structural materials also claims our a t t e n t i o n , dueto its decisive importance not only in fue! e l e m e n t s , but also -in the reactor's basic c o m p o n e n t s .

These aspects have beentreated with different levéis ofi n t e n s i t y , but always trying to be realistic in evaluating theirrelative i m p o r t a n c e . In this r e g a r d , our aim is to be useful w i -thin the scope of the National Energy P l a n , to contribute expe-rience and means in joint p r o j e c t s , support and service to manu_facturers of nuclear power plant equipment or with those firrnswhich opérate these p l a n t s .

FABRICACIÓNY

MONTAJE

TECNOLOGÍADE EQUIPOS

DISEÑO DECOMBUSTIBLES

MATERIALESCERÁMICOS

PREPARACIÓNDE PROTOTIPOS

EVALUACIÓN DECOMBUSTIBLES

TÉCNICASFUNDAMENTALES

PROPIEDADESTECNOLÓGICAS

CIRCUITOS DEENSAYO

ENSAYOSNO DESTRUCTIVOS

ENSAYOSMECÁNICOS

CORROSIÓN

METALURGIAF ÍS ICA

METROLOGÍA

ENSAYOS ESTÁTICOSY DINÁMICOS

EVALUACIÓN FRACTURAY CORROSIÓN

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-10-

1• METALLURGY DIVISIÓN2. Administrative Services3. Fuel Elements4. Prototype testing5. Metallurgic technology support6. Structural materials7. Manufacturing and assembly8. Equipment technolog9. Fuel design

10. Ceramic materials11. Prototype preparation12. Fuel evaluation13. Basi c techni ques14. Technological properties15. Test circuits16. Non-destructive tests17. Mechanical tests18. Corrosión19. Physical metallurgy

20. Metrology21. Dynamic and static tests22. Evaluation of fracture and corrosión23. Personnel

-11-

Research and Development Work inFAST REACTOR FUEL DEVELOPMENT

by

K.R. Kummerer

KERNFORSCHUNGSZENTRUM KARLSRUHE

Instituí fur Material-und Festkorperforschung

Contents:

1. Performance Requirement2. Structure of R+D Work3. Problems Identification4. Methods and Procedures5. Irradiation Experiments6. Present Status7. Future Development Trends

-12-

Research and Development Work in

FAST REACTOR FUEL DEVELOPMENT

C o ni p a c t

of a Lecture, held on September 21, 1978, at

Junta de Energia Nuclear Madrid

by

K.R. Kummerer

KERNFORSCHUNGSZENTRUM KARLSRUHE

Institut für Material- und Festkorperforschung

The- aim of the research and development work is

to provide knowledge and experience on design andbehaviour of fuel elements for prototype and coiranercialfast reactors and

to provide knowledge and experience for all steps ofthe fuel cycle.

In this lecture we make a selection out of the very broadfield: We concéntrate mostly on oxide fuel and consider mainlythe irradiation performance of fuel pins and bundles. We dothat within the following chapters:

1. Performance Requirements2. Structure of R+D Work3. Problems Identification4. Methods and Procedures5. Irradiation Experiments6. Present Status7. Future Development Trends

- 13 -

1. Performance Requirements

During the lifetime of a fuel element we have scheduledoperation phases and non-scheduled situations.

Scheduled operation phases are:

start-upsteady state at different power levéisoperational power changepower ramping

Non-scheduled situations are:

reactor scrampower transientsloss of coolantoperation with failed pins

The steady state operation performance requirements mainly stemfrom basic economic assumptions as follows:

The linear rod power of the fuel pin and the specific power(in thermal kW per kg fissile material) influence the coreinventory. At high specific power valúes the fissile inventoryin the core is reduced.

The fuel burnup directly acts on the fuel cycle costs, becauseat a higher burnup the specific reprocessing and refabricatingeffort is reduced.

The máximum tolerable ciad temperature is a decisive figurefor coolant outlet temperature and, henee, for the thermo-dynamic efficieney of the electricity generation.

In the following Table the nominal standard operational andperformance data for oxide and carbide fuel pins are compiled.As the power and temperature distribution in the core is by farnot uniform (mainly due to radial and axial flux variation andthe axial temperature gradient in the cooling channels), thenominal máximum valúes are quoted:

Linear rod power (W/cm)

Ciad midwall temperature ( C)

Fuel burnup(MW days/kg heavy metal)

Oxide

450

600

80

Carbide

800

600

70

It should be held in mind that - at an actual fuel designthese nominal figures are superposed by hot spot calculations,the results of which being then confronted with the possibledesign limits of the materials involved.

- 14-

2. Structure of R+D Work

Our development work is oriented on actual and future fastreactor core design. In the case of oxide fuel these designsare determined for:

KNK-II Test ZoneSNR-300 Mark laSNR-3OO Mark IISNR-2

In the case of carbide fuel the design work aims to:

KNK-II Test BundleHigh Performance Prototype Core

All the development activities can be devided in the followingsepárate sections:

FuelCladding MaterialPinBundle

3. Problems Identification

The problems in the different sections of development areidentified and outlined in a short manner.

F U E L :

As a background, let us consider the basic properties ofdifferent fuel types:

Melting point (°C)

Theoretical density(g/cm3)

Heavy metal density'''(g/cm3)

Thermal conductivityat 200 °C (W/Km)

Metalloid content(w/o)

Metal

U

1132

19.04

19.04

28

Pu

641

19.82

19.82

12

0

Oxide

uo2

27 60

10.96

9.66

7

Pu02

22ao

11 .46

10.11

6

11.8

Carbide

ÜC

2400

13.63

12.97

20

PuC

16542)

13.62

12.96

8

4.8

at room temperature; 2) melts peritectically

_ 15 -

The main fuel problems can be categorized as follows;

Design

Fabrication

IrradiationBehaviour

Fuel FormFuel StructureDensityStoichiometrySolubility

MethodsSpecificationsQuality Control

SwellingRestructuringFission Gas ReléaseU-Pu-SegregationFission Product MigrationO/M-Shift

Which máximum burnup is reachable ?

C L A D D I N G

At first we consider potential cladding materials. They areenlisted and discussed in the light of requirements as follows:

Ferritic Steels

StabilizedAusteniticStainless Steels

Nickel-BaseAlloys

RefractoryAlloys(e.g. Vanadium)

MechanicalStability

up to 500°C

up to 500°C

up to 7CX)°C

very good

CorrosiónResistance

good

good

good

intemalattack

FastNeutrónAbsorption

low

low

higher

low

Irrad.Behaviour

good

someswelling

someswelling

good

AvailabilityFabrication

very good

very good

good

peor

Presently only austenitic stainless steels are chosen for actualdesign. The problems which have to be dealt with are:

- 16-

Mechanical Properties:

Corrosión:

Fast Neutrón Influence

StrengthDuctilityCreep

Na-CorrosionFuel-Clad Compatibility

Volume IncreaseEmbrittlement

P I N and B U N D L E :

At pin and bundle design the problems can be identified asfollows:

Pin Design:

Bundle Design:

Diameter OptimizationFuel-Clad GapPlenum Position and Length

Spacer, Type and DistanceTemperature GradientsHot Spot CalculationsSwelling Accomodation

Pin and Bundle Performance at Given Requirements

4. Methods and Procedures

The methods and procedures to be applied in the development workcan be characterized schematically as follows:

DoandLook

Experimental Work

Considerand -*- Theoretical WorkThink

FabricationQuality ControlIrradiation ExperimentsPost Irradiation Examination

Design of ExperimentsEvaluationModeling

Fabrication experiments are necessary for fuel. Quality controlmethods are to be developed for fuel, ciad,pin and bundle. Theirradiation experiments are designed for a single featurequestion, for parameter variation or as a performance test.

- 17 -

For the post irradiation examination hot cell facilities andextensive experience in operating such facilities have to begained.

The theoretical work is based on experiments and their evaluation.In the modeling of fuel and pin, the design, operation andmaterials parameters are combined into mathematical relation-ship within a computer code system leading to"load guantities"which have to remain within failure limitations:

COMPUTER

CODE

SYSTEM

íí L0AD \\QUANTIT1ES

FAILURE -^ • •LIMITS

5

The parameter groups are detailed as follows:

- 1 8 -

DesignParameters A

OperationConditíons B

MaterialQuantities Mof Fue! and Ciad

LoadQuantities L

FaiiureLimitations S

Materials CompositionState of FabricationExternal Geomeiryinterna! Geometry

etc.Linear Rod PowerNeutrón FiuxCooiant Temperature

etc.

Meciianicai PropertiesThermai ConductivityHeat Transition Fuel/CiadSweüing BeitavtourPore MigrationFission Gas Reléase

etc.

Temperature DistributionStresses and StrainsMaterial DistributionCorrosión Attack

etc.

Aiiowabie Máximum of— Temperatures— Stresses and Strains— Burnup

etc.

- 1 9 -

5. Irradiation Experiments

Irradiation experiments are carried out for fuel, ciad, pin andbundle investigation. They simúlate steady state condition,startup, load follow operation or also abnormal conditions.

An irradiation experiment normally lasts several years from thefirst idea to the final evaiuation of the results. One candistinguish 6 phases as f oll.ows:

PHASE

1

2

3

4

5

6

IDENTIFICATION OF STEPS

Definiíion of Task and Objectives, Theoretical Anaiysisof Expectations

Conceptual Design, Speciíication of Test Sampie,Developmení of Irradiation Tool

Fabrication of Samples, Final Control

Irradiation in the Reactor, Evaiuation of IntermedíateOperational Data

Non-destructiveDestructiveSpecial

Post Irradiation Examination

Evaiuation of Ail Results, Documentation, Interpretationby Modeling Calculations

Let us consider at first the KfK oxide irradiation program.There are special fuel experiments (e.g. for creep and swellingbehaviour) and also special cladding material irradiations. Welimit here the discussion on oxide pin irradiation experimentsaccording to the list:

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Reactor

FR-2KarisruheF.R.Germany

BR-2MolBeigium

DFRDounreayScotland/UK

RapsodieCadaracheFrance

Identificationof Experíment

Capsute-Vg. 4a

Capsuie-Vg. 4b

Capsuie-Vg. 5a

Capsule-Vg, 5b

Loop-Vg. 3

Loop-Vg. 5

Moi-7A

Mol-7B

Mol-7D

Mo!-8A

Moi-8B

Mo!-8C

Moí-SD

Mol-16

DFR-304

DFR-350

DFR-435

DFR-455

RAPSODIE 1

RAPSODIE II

NumberofPins

28

35

9

18

34

10

7

18

19

2

2

10

12

14

3

23

6

60

2x34

19

Objeciives

Pin performance at high linear rod power

Pin performance up to high burnup

Correlation between fuel density and restructuring

Performance of small diameter pins

Startup and short term behaviour

Pin behaviour at power cyciing

Smali bundie performance

Pin performance at hoi spot conditions

Performance of finned tubes in a 19-pin-bundie

Pin performance

Pin performance, in-pile fission gas pressure

High burnup pin performance, in-piie f.g. pressure

Thermal behaviour as function of Bup, fuei central T in-pile

Chemical interaction between fuel and ciad

Pin performance in fast flux

Pin performance within a 77-pin- subassembly

Pin performance after pre- irradiation in DFR-350

Pin and bundie performance in fast flux

Performance of two bundles in fast flux

Performance of bundie in fast flux

Out of this program soiue interesting features are demonstratedin the following.

At the irradiation test group Vg. 5a in the Karlsruhe thermalreactor FR 2,separated regions within a fuel pin were loadedwith different fuel densities. The restructuring of the fuelafter an irradiation of 17 MWattdays per kg heavy metal wasdistinctly different. Not only the diameter of the central

- 21 -

-07 -

90

% th D

% th D

93

% th D

87

% íh D

_ 1

2.5

2.0

1.5

1.0

0.5

K

S i

z .,

.

t

— : __J* — u

8/i 87 90 93-•— Fabrication Density {%T. D.i

channel (Z), but also thecolumnar grain zone(S) and theregión of uniform grain growth (K]showed linear correlation to thefuel density.

In the course of the experimentM0I-8C in the Belgian reactorBR 2,10 pins with fuel, blanketand gas plenum regions wereirradiated up to high burnup.The fission gas pressure buildupwas experimentally determinedin-pile as demonstrated in thediagram - together with thecalculated reléase rateas follows:

_ 40 —

¡ J

304

400 •

50 S

Finí! 6umu{) f

A«enijt Bunnp [ H W d k í MI100

22 _

Also the typical fuel restructuring and the migration of Pu(a-autoradiography) and fission products (B-y-autoradiography)were made visible at the post irradiation examination:

Pin Cross Seclion a-Autoradiography S-y-Auloradiograptiy

1 ram

The chemical interaction between fuel and ciad led to areaction zone of about 150 yin:

-23 _

Another experiment in theBR 2 - the bundle irra-diation Mol-7A - showede.g. a distinct axialcesium migration. In theblanket pellets adjacentto the fuel, a remarkablelocal cesium enrichmentis detected by g- y ~autoradiography.

The integral gamma profileof the- whole pin demon-strates cesium peaksaccordingly:

Cs-137Cs-134

Caramooraphy /3,^'AutoradiOBraphy

Local Cesium Enrichment

Cs-137,Cs-134

Cs-137

Cs-137i Cs-134

Blanket Fue! Blanket Plenum

Cesium Migration in the Mol-7A Demonstrated by the IntegralGamma Profile

_24 -

In the Dounreay Fast Reactor a bundle of 77 pins wereirradiated up to 55 MWattdays per kg M. The so-calledDFR-350 test pins contained fuel, blanket and gas plenumthus simulating real prototype fuel pins:

Blanket

A special feature in the fast neutrón flux was the diameterincrease during irradiation:

„ 6.08-

1 6.07-

6.05-

6.04-

6.03-

6,02-

6. 1-

6.00-

5.99

Clad 1.4961

Test Pin NS G 49

Ad= 1.30 %T

x Average before Irradiation• Average after Irradiation

6.03-

6.02-

6.01-

6.00-

5.99-

5.98-

Clad 1.4988

Test Pin N i G16

0 100 200 300 400 500[mm]

End Plug Fuel Blanket Plenum

- 25 -

This phenomenon is caused by the fast neutrón induced swellingof stainless steel. Different steel types showed differentswelling rate.

Finally the attention is drawn to the KfK carbide irradiationprogramme. A lot of fuel and pin experiments include manyimportant objectives:

Reactor

FR-2KarisruheF.R.Germany

BR-2MolBelgium

DFRDounreayScotland/UK

Identificationof Experiment

FR-2/73K

FR-2/100

Capsule-Vg. 6aCapsule-Vg. 6cCapsule-Vg. 6dCapsule-Vg. 6e

Loop-Vg. 4ALoop-Vg. 5K

Mol-12

Mo l - I 1 /K2Mol-11/K3Moi-11/K4

Mol-15

DFR-330/1DFR-330/2DFR-330/3

Bonding

(unciad)

(unclad)

HeArHeNa

He.ArHe

(unciad)

HeNaHe

Na

NaHeNa

No. ofSamplesor Pins

14

24

6633

810

2

344

4

777

Objectives

Creep behaviour

Free and restraint swelling

Pin performance atdifferentparameters

Pin startup andcyciing behaviour

Creep behaviour

Pin perfjn-pilefission gas press.and fuel temp. measurement

Influence of M2C3 on corrosión

Performance of 7-oin-bundles

- 26 -

According to the experimental results the irradiation-induced creep in carbide fuel is lower by an order ofmagnitude compared to oxide:

1200 IODO 800 700 S00 500 400 T°C10-"

10"

io- 6-

l&ireiil cretp

Alt ¿iti «nwl&ií tiF= I i 1 0 " l / c s 3 i

c r = 2 0 MU/»2

io-e

UN-Bnicklicliir. Ziiwr 'ose

UG-Cliugü

0,6 0.8 1.0 1.2 1.4 1.6 1/T 10"3 K"

Another feature is the carburization of the cladding whichis measured by the microhardness, the amount of which beingdependent on steel type:

"= 300-i

500

= ¿00-

200-

25 50 75Wall-Thickness !%]

6. Present Status

As far as the present status of fast reactor fuel developmentis concerned the following statements can be given:

Many additional experience is available throughout theworld, especially in France, Great Britain, USA, USSRand Japan.

The requirements for steady state operation can be fulfilledup to high burnup.

- 27 -

Fabrication of fuel and closing the fuel cycle is stillproblematic to some extent.

Non-steady state and transient operation phases needadditional experiments and theoretical work.

There is still discussion on some main design variableslike fuel density and pin diameter.

But in spite of some limitations in knowledge and experiencea safe and reliable design of oxide fuel elements is possible.^ e design principies are:

No gross redistribution of fuel

Cladding'tight and strong

Pin geometry unchanged

According to these principies the fuel element for the DEBENEprototype reactor SNR-3OO in Germany is outlined in thefollowing drawing;

-»H . 2,8 *«t-

HEAD MIXING DEVICE SPACER T1E RODFUEL PIN BUNDLE WRAPPER TUBE

SPARK HONEYCOMBERODED SPACERSPACER

J=OOT

FUEL PIN LENGTH 2475

0 6

!-AAA"Vv'Ai_J

400

! " * i

950

í-r-ií-Ttí-rO:- ~.~ -ítT-fe-O

400

I •vi____ _ _ _ . _ _ ^ ..

!

^ H i!' ' ' 'I I \ \

AXIAL BLANKET ACTIVE ZONE AXIAL BLANKET FISSION GAS PLENUM

1400 —CORE MID PLAÑE

J 1750

| FUEL PIN

ELEMENT

i

LENGTH

LENGTH

2475

3700

Finally a synopsis of different fast reactor fuel pin designis given:

• • • >O S G H) XJ ni

SUPERPIIENIX

4

DFRMk B-f«i

5~

CFR-I

OREA! 8RIIAIN

KNK-n

8

SNR-300Mlih-U I M*fti~D

SNR-J

GERMANY tDENELUX

PEC

HA IV

BR-5

13 14 15

«M-400

16

EBR-S

18USA.

JOYO

20"

MONM

21 22INDIA

r-ooo

Fast Reactor Fuel Pins— Synopsis of the Different Designs

- 29 -

7. Future Development Trends

The future development trends are highly influenced by fuelcycle considerations, e.g. the fuel cycle costs and thereprocessing behaviour. Another incentive may be the guestionof U-utilization. Major future development trends are:

Increase pin diameter

Burnup -> 100 MWd/kg M

Increase fuel density

Fuel homogeneity

Total solubility of fuel

Further open questions refer to Pu-losses in the fuel cycleand security measures for protecting and safeguarding Pu-fuel.Also the performance of the breeder blanket elements shouldbe investigated more extensively.

-30-

Possibilities for Fast Breeder Fuel Pin Irradiation

in the Karlsruhe Nuclear Research Centre

by

E.Bojarsky

KERNFORSCHÜNGSZENTRUM KARLSRUHE

Institut für Material- und Festkorperforschung

Contents:

1. Introduction

2. The Karlsruhe FR2-Reactor

3. The FR2 Capsule Irradiation Device in General

4. Double Walled Na/PbBi Capsule

5. Gas Gap Double Capsule

6. Single Walled NaK-Capsule

7. Fuel Creep and Swelling Capsules

8. Examination Techniques for Fuel Pins and Capsules

- 3 1 -

1. Introduction

I'11 try to give some informations about technical equipments and

possibilities for fast breeder fuel pin irradiation experiments in the

Karlsruhe Nuclear Research Centre in order to obtain a general view and

not to learn many technical details.

We use of course not only our own reactors in Karlsruhe, the FR2, KNK and

M2FR, but also some different reactors in other countries, for example in

Belgiuin the material testreactor BR2, in the Netherlands the high flux

reactor HFR, in France the Rapsodie and CABRI and in Great Britain the Proto-

typ FastReactor PFR. But because our common IVO-experiments will be installed

in our FR2-reactor in Karlsruhe I will 'confine myself to this reactor.

2. The Karlsruhe FR2-Reactor •

Our good oíd FR2-reactor is a so-called all purpose research reactor of the

vessel type (Fig. 1). Heavy water serves as the moderator, coolant and neutrón

reflector. There are nuiaerous vertical and horizontal experimental channels

and .also a thennal graphite coluinn for irradiations and beamhole experiments,

as well as a rabbit system for small samples.

The photo of figure 2 gives an impression of the reactor hall with the reactor

block, some physical beamhole experiments, the refuelling machine ect.

The reactor core is relatively large with approximately 2.2 m diametre and

2.16 m height (Fig. 3). It is operated at a thermal power of 44 MW, generating14 2

a máximum thermal neutrón flux of approximately 10 n/cm s.

-32-

We use slightly enriched UO with 2% U-235 arranged in the fuel elements in

a 7 pin bundle surrounded by a hexagonal Zry-shroud tube. The fuel pin outer

diameter is 13 mm. The mean temperature of the coolant is about 50 to 60 C.

The core is normally loaded with around 150 fuel elements, and there are

another 41 intermediate lattice positions. As our irradiation capsules fit

into all these positions we have enough freedom for lo.cating our commen fuel

pin experiments according to the necessary neutrón flux.

3. The FR2 capsule irradiation device in general

Let us now turn to our capsule irradiation device. As you see on figure 3,

it is a long slim thing, extending from the reactor top shield down to the

coolant entrance píate.

The photo of figure 4 shows the capsule rig in clean and fresh condition

hanging from the eran in the reactor hall. You can imagine the necessary

aecuracy in straightness for example.

In principie the capsule irradiation device is composed of three elements

(Fig. 5) : the irradiation capsule proper - about 3 m long - with the .specimen,

the upper part and the coolant guiding unit. The upper part serves mainly as

a shielding plug and for transmission of the measuring leads. The capsule rigs

are equiped with an activity monitoring system of the cooling water for capsule rigs

leakage detection and numerous thermocouples . Easiness of disassembly allows

the repeated use of the upper part and the coolant guiding unit, and they

both can be used with any type of capsule. Since more than ten years we

developed and used about ten capsule types or modifications. Let us have a

short look at some of them.

4. Double walled Na/PbBi capsule

The first type we built in larger quantities was a double walled Na/PbBi

capsule (Fig. 6 ). In this case the fuel pins are enclosed in an inner capsule

filled with sodium or sodium/potassium for a good heat transfer, and this is

enclosed for safety reasons in a second outer capsule filled. with eutectic

-33-

olead bismuth alloy ( melting point 125 C ).

5. Gas gap double capsule

Another type is a gas gap double capsule for fuel testing without a normal

cladding tube (Fig. 7). Here the fuel is enclosed in a thick walled very

strong molybdenum capsule which is surrounded by an outher capsule in stainless

steel. In between there is a defined gas gap, in which the desired temperature

gradient arises.

6. Single walled NaK - capsule

Now we come to that type we'll use for our common experiment (Fig. 8). It is a

single walled NaK-capsule. In this case the fuel pin is immersed in a heat

transfer médium which is always liquid and gives no problems with shrinkholeso

ect. The eutectic NaK has a melting point of -11 C. The liquid metal is

separated from the cooling water by only a single wall, of which obviously the

stability and intrgrity have to be proved and guaranteed with a high degree

of confidence. In the NaK space an intermediate tube is provided which has two

functions: Firstly, it prevenís any significant convection in the liquid metal

and secondly the fuel pin temperature can be adjusted within relativly wide

limits by changing the material of the intermediate tube and its wall thickness.

For our conmon irradiation experiment we 11 use two capsule modif ications: the

first one for relatively high rod power and modérate cladding temperature with

an intermediate tube of Zry 2 and a thin walled anticonvection tube, and the

second one for low rod power and high cladding temperature with an intermediate

tube system of two stainless steel tubes with a gas gap in between.

7. Fuel creep and swelling capsules

At the end of this FR2~capsule review let us have a short look at three of our

fuel creep and swelling capsule types. In the first one the fuel sample stack

is immersed in liquid sodium and compressed by a defined axial load (Fig. 9).

The linear movement of the compressing pistón can be measured by a carefully

calibrated inductive system with high accurary.

-34-

In the second case the fuel column is pressed into a very strong molybdenum

cylinder and the axial swelling-is measured continously (Fig. 10). This type

is designed for high fuel temperatures, that is between 1000 and 2000 C.

The last type is a high gas pressure capsule (Fig. 11), in which the fuel

swelling under triaxial compression at high temperatures will be investigated.

The layout pressure of 500 bar pose some material and fabrication difficulties.

8. Examination techniques for fuel pins and capsules

The very important quality control of the fresh fabricated fuel, cladding tubes

and capsule components normally is carried out and documented in the factories

and workshops according to our specifications. In Karlsruhe we have a final check

of the fuel pins by x-raying in order tq inspect the correct arrangement of

the interior and the quality of the end plug qeldings. The profile of the fresh

fuel pin surface - the diameter, ovality, straightness and length -can be measured

in our hot cells with high accuracy. This offers the possibility to compare it

with the post irradiation'data later on. During and after the capsule assembly

there is a series of leak tests,pressure tests, x~ray tests, electrical test

of the thermocouples etc. And what about the intermediate and post irradiation

examination.

There is for example the neutrón radiography for nondestructive testing of the

capsule. This can be done in front of the thermal column of the FR2-reactor.

This technique has been developed especially for highly radioactive objects. The

principie is to modulate the intensity of a collimated neutrón beam by the

structural, absorption and scattering properties of the object. Behind the object

there is a dysprosium foil, which absorb a latent picture by radioactivation. After-

wards this information has only to be transfered to an ordinary x-ray film.

Inside our hot cells there are facilities for x-raying and betatrón radio-

graphy as valuable tools for more detailed nondestructiv post irradiation

examination of the capsules and the test fuel pins proper. Especially the high

energetic betatrón radiation with 18 MeV peak energy is best suitable for the

examination of radioactive objects, that means to make visible the interior

structur. So it is possible to detect the condition of the cladding and the

fuel column, the shape of the central channel in the fuel, cracks and density

differences etc. This is necessary in order to determine the different cuts

and other destrucktive and special examinations.

-35-

In addition of course all the necessary mechanical, physical and chemical

techniques are available, for example to take Y~scanning in order to detect

the axial fission product distribution, or to take a- as well as S/y -auto-

radiographs of a distinct fuel zone, or to evalúate the free and entrapped

and dissolved fission gases, or last not least to make radiochemical burn up

analysis. That is ónly to give some.examples. But as you certainly know, because

post irradiation examination is very time consuming and expensive in every

case one has carefully to reflect and to decide on the best suitable examination

program.

- 3 6 -

Fig. 1

The FR2 - Reactor

in karisruhe

Fig. 2

A view of FR2

- 3 7 -

KAPSEL-VERSUCHSEINSATZ

COREH=2160mm0=24OOmmfth max =10l4n cnv'

He-LOOPEINSAT.

Fig.3

FR2-reactor withcapsule irradiationdevices.

Fig.4

A fresh irradiationcapsule hanging fromthe crane in the reactorhall.

-38-

1~

* • -V* * y • ••\ í \ * - •

../s v• '/•"•'A'"--

: : • • • • • ' / :> ' - " :

6380

3080

ii i7

26

Elektrische SteckverbindungElectrical connection

.ReaktordecketReactor cover

.OBERTEILUPPER PART

AktivitátskontrolleFission product detection

KupplungCoupling

.KUHLWASSERFUHRUNGCOOLANT GUIDING SYSTEM

.BESTRAHLUNGSKAPSELIRRADIATION CAPSULE

_BrennstabFuel pin

•D2O

FR2- Kapselversuchseinsatz

FR2 Fuel Pin Irradiatibn Rig

- 3 9 -

3080

Elektrische SteckverbindungElectrical connection

KupplungCoupling

10 Thermoelemente10 Thermocouples

Blei-WismutLead -Bismuth

Inneres KapselrohrInner capsule tube

NatriumSodium

ÁuBeres KapselrohrOuter capsule tube

BrennstabFuel pin

KühlwasserführungsrohrCoolant guiding pipe

Fíg. 6Na/Pb Bi- Doppelkapsel

Na/Pb Bi Double Capsule

-40-

ThermoelementeThermocouple

.GasspaltGas gap

BrennstoffpelletsFuel pellets

Innere Molybdan-KapselInner Molybdenum capsule

ÁuBere Edelstahlkapsel 220

Outer Stainless Steel capsule 22mm(

_KühlwasserführungsrohrCoolant guide pipe

F¡g.7Gasspalt- Doppelkapsel

Gas Gap Double Capsule

- 4 1 -

NaK

DoO

DurchführungsstopfenPenetration plug

ThermoelementeThermocouples

ZwischenrohrIntermedíate tube

BrennstabFuel pin

NaK-SpalteNaK-gaps

Kapselrohr 024/27 mmCapsule pipe

WrmVlIMFUL

Fig.8: Einwandige NaK-Kapsel für den FR2Single walled NaK capsule for the FR2

- 4 2 -

Outerpressure capsule 30/340

Inductivedisplacement transducer

Compressiontransmission pistón

Insulating gas gaps

Fuel sample stack

Thermocouples

-43-

c

OC

Outerpressure capsule 30/34

Inductivedisplacement transducer

Compressiontransmission pistón

Insulating gas gaps

Fuel column

Molybdenum cladding

Thermocouples

-44-

To pressure supplyand pressure transducer

Filter packet

Insulating piece

insulating gas gaps

Fuel sample stack

Intermedíate tube (mo)

Pressure capsule

Pressure sleeve

Thermocouple

Outercapsule tube 34/38

-45-

THE IRRADIATION EXPERIMENT IVO-FR 2 - V g . 7

by

H. E lbe l

KERNFORSCHUNGSZENTRUM KARLSRUHE

Institut für Material- und FestkSrperforschung

C o n t e n t s :

1. Objectives

2. Fuel Pin Design

3. Irradiation Conditions

4. Irradiation Equipment

4.1 Reactor

4.2 Irradiation Device

4.3 Instrumentation

5. Irradiation Behaviour

5.1 Consequence of the Thermal Neutrón Flux

5.2 Influence of the Fabrication Tolerances

-46-

1. Objectives

The solubility of Fast Breeder mixed oxide fuel depends primarily on the

fabrication process of this fuel. Because the structure of the fuel changes

during its burnup in the reactor, for example by cracking, partial densifi-

cation, grain growth, plutonium redistribution and embedment of fission pro-

ducts, the original solubility can be more or less modified.

The purpose of the experiment IVO-FR2-Vg7 is, therefore, to show to what extent

the solubility of the fuel can be influenced by the operation conditions. By

the choice of two different kinds of fuel which differ in the plutonium enrich-

ment it is intended to study, in addition, the influence of the fabrication

parameters with the aim to deduce criteria for the specification of an optimized

fuel.

Another aim of the experiment is the verification of a fuel pin concept with

fuel of high density.

2. Fuel Pin Design

For the experiment a fuel pin concept of the Mark II type was chosen, which

corresponds in its main features to that of the second core of the KNK II, the

Mark II core of the SNR 300 or the core of the SNR 2 (see Table 1).

The fissile material of the fuel is only plutonium according to the conditions

in the future commercial Fast Breeder reactors.

The experiment is planned as test irradiation. The length of the fuel column

does not play an important role. Short fuel pins are, therefore, sufficient

for the purpose of the experiment (Fig.1).

The length of the fuel column was determined on the basis of the following

criteria:

1. The column should be, on the one hand, so large that sufficient fuel

would be available for all the planned post-irradiation examinations

besides the reguirement that "end effects" could be neglected.

2. On the other hand, the column length is limited by the requirement

to provide the linear rod power as constant as possible along the pin

under the given neutrón flux conditions of the test reactor.

-47-

3. Irradiation Conditions

The irradiation conditions of the test pins have been chosen in a way which

allows to simúlate the operation conditions of a Fast Breeder fuel pin at the

position of the máximum of the linear rod power as well as the coolant temperature<

Fig. 2 shows schematically the axial distribution of linear rod power and coolant

temperature of a Fast Breeder fuel pin.

The position of the maximal linear rod power is equivalent to the position of

the maximal radial temperature gradient in the fuel. The position of the maximal

coolant temperature coincides with the minimum of the linear rod power.

The reference valúes which have been chosen for the two positions are for the

linear rod power 450 W/cm and 200 W/cm and for the ciad surface temperature

520 C and 600 C, respectively.

4. Irradiation Equipment

4.1 Reactor

The experiment will be performed in the FR2, the test reactor of the Karlsruhe

Nuclear Research Centre. This reactor operates with thermal neutrons. The reasons

to use this reactor are:

1. A sufficiently large number of appropriate irradiation

positions is available.

2. The material for the fabrication of the required irradiation

devices, the so-called capsules, was already existing.

3. The expenditures for the irradiation are low as compared to other possible

test reactors.

4. A further advantage is the cióse contact to the other institutions

taking part in the experiment, for example, in the field of fuel

pin fabrication, capsule assembling, control of irradiation and

post-irradiation examination.

-48-

4»2 Irradiation Device

In the experiment it is intended to vary 5 parameters (see Table 2). These are

linear rod power (450 W/cm and 200 W/cm), ciad surface' temperature (520 C and

600 C ) , fabrication process (marked by H, und EL), plutonium enrichment (15 %

and 30 %) and burnup (low burnup A1 and médium burnup A,,) .

In the capsule which was selected for the experiment three of the short fuel

pins can be tested simultaneously. In order to satisfy under this condition the

required number of parameter variations a minimum number of 4 capsules must be

irradiated.

Due to the axial neutrón-flux distribution in the core of the FR2 two of the

three fuel pins in a capsule (No.2 and 3) can be tested under approximately

the same operation conditions. The third fuel pin (No.l) is located in some-

what lower flux. Its operation condition differs from those of the other two

according to the actual flux and trie plutonium enrichment chosen for the fuel.

The parameter variation was concentrated on the high linear rod power for which

the largest restructuring effect of the fuel can be expected. Three capsules

(No.i,2 and 3) are to opérate at this power.

4.3 Instrumentation

The capsules are provided with 8 thermocouples in each one which record the

temperatures at the surface of the fuel pins. The positions of the thermocouples

are shown in Fig. 3. From the records of the thermocouples No. 3 to 8 the linear

rod power is derived. The thermocouples No.9 and 10 are used for the control of

the level of the NaK filling of the capsule.

5. Irradiation Behaviour

5.1 Consequence of the Thermal Neutrón Flux

The irradiation of the test pins will be performed in thermal neutrón flux.

This means: Due to the self-shielding of the fuel against thermal neutrons the

power rating will not be homogeneousover a transverse section of the fuel as

it is in fast neutrón flux. Fig. 4 shows, for example, the flux depression for

the fuel with 15 % plutonium enrichment. The power rating is expected to be

about 30 % lower in the centre of the fuel pellet than at the periphery of the

pellet.

-49-

This difference leads, of course, to a different irradiation behaviour of the

fuel. The computer code system SATURN predicts at the beginning of the irra-

diation for a linear rod power of 450 W/cm under thermal neutrón flux a cen-

terline tempera ture which is in the fuel pellets with 15 % PuO_ about 100 C,

in those with 30 % PuO0 about 250 C lower than under fast flux (see Fig.5).

But the deviation of the temperatúre profile is tolerable. The temperature

gradient in the outer región of the fuel pellet is approximately the same.

Despite the initial differences of the radial temperature profile the restruc-

turing of the fuel pellets can be expected to be very similar to that under

fast flux. There is, therefore, no doubt that the experimental data which will

be obtained under thermal flux can be extrapolated to fast flux.

5.2 Influence of the Fabrication Tolerances

Further calculations have been performed again using the computer code system

SATURN to predict the behaviour of the fuel pins under irradiation in more detall.

One aim of these calculations'is to show that the desired operation conditions

will be attained in the reactor. The second aim is to assure that no critical

limits will be exceeded taking into account deviations of the actual pellet

parameters from the nominal valúes due to the fabrication tolerances.

A critical valué is, for example, the temperature in the centre of the fuel

pellet. It is the highest temperature valué which appears in the fuel. This

valué must be sufficiently far away from the melting point of the fuel to avoid

melting (for 15 % Pu0o: T = ~ 2760 °C, for 30 % Pu0o: T = ~ 2690 °C).2 m ¿ m

Fig. 6, for example, shows the influence of the oxygen content on the fuel

centerline temperature at the beginning of the irradiation (BOL). The heat

conductivity of the fuel decreases with decreasing 0/M ratio. As consequence,

the centerline temperature increases. Under the chosen operation conditions

the 0/M ratio must not be lower than 1.95 to make absolutely sure that melting

will not occur.

Besides the centerline temperature fuel surface temperature and hot radial gap

width are given in Fig. 6. Both decrease with decreasing 0/M ratio according

to the increasing thermal expansión of the fuel.

-50-

Another example for the influence of the fabrication parameters on the fuel

behaviour is pictured in Fig. 7. The heat transfer out of the fuel pin into

the coolant of the reactor depends very strongly on the heat conductivity of the

gap between fuel pellets and ciad. The gap conductivity is, primarily, determined

by the gas in the gap. This is usually helium which is filled in at fabrication

with high purity. Generally a valué of at least 95 % is specified.

Additional gas is brought into the pin by the fuel in its pores and matrix. It

comes from the atmosphere of the sinter process as well as of the storage faci-

lity. It is released when the fuel is heated up in. the reactor. The composition

of the fill gas changes, therefore, already at beginning of operation. The con-

sequences for the fuel behaviour are explained in Fig. 7. The heat conductivity

of the gap decreases with increasing pollution of the fill gas helium. Surface

and centerline temperature of the fuel increase accordingly.

The experiment IVO-FR2-Vg7 is intended to improve the knowledge on this and

similar phenomena by means of a careful characterization of the fuel before

and after irradiation. This is the way we want to deduce criteria for the

specification of an optimized fuel.

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IMF-O6-3-é-13Ob

F i g . 1 : The f ue ! p in f o r FR2-Vg7

- 5 2 -

Table 1 : u v _ j i y i i u u i u I U I

Fuel pellets:

Material

U-235 enrichmentPu conteníPellet density0/M ratioTheoretical densityPellet diameterPellet hight

Cladding:

MaterialOuter diameter

Inner diameter

Fuei p i n :

Smeared densityGap width, radialFill gas

i i\ ¿- ~ v y . /

U0?- PuO 7

15and 3 0 %9 4 % t h . d .1 9 7- 11 ,06g /cm 3

6,40mm8,00mm

ss 1.4970 cw7,60 mm6,60mm

8 8 % t h . d .100 jjm

He

Table 2: Parameter variations of the irradiation experiment FR2-Vg7

capsule

No.

1

2

3

4

fuel pin

No.

1

2

3

1

2

3

1

2

3

1

2

3

rod power

(w/cm)

< 450

450

450

< 450

450

4 50

< 450

450

45O

< 2OO

< 200

< 200

ciad surface temp.

t (° c)

< 52O

52O

520

< 520

52O

520

< 52O

520

520

- 600

= 600

- 600

fabr. process

Hl

H1

H2

Hl

Hl

H2

H2

Hl

H2

Hl

Hl

H2

plutonium enrichm.

( % )

15

30

30

15

30

30

15

15

15

3O

15

15

burnup

< A 1Al

Al

< A 2 .

A2

A2

<A2

A2

A2

A2

A2

A2

eni

A. low burnup, middle to high burnup

exx

650

—i 600

550

500

Oo

¿50

400

350

en-p»I

300100 120

z fcmj

F i g . 2: Ax ia l d i s t r i b u t i o n of l i n e a r rod power and c iad sur face

temperature fo r a Fast Breeder fue l p in (schematic diagram)

-55-

topnumber offuel pin

positions ofthermocouples

TC 10

TC 9

TC 8

TC 7, 6.5

average position of neutrón fluxmáximum in the core

TC 4,3

bottom

F i g . 3: P o s i t i o n o f f u e ! p i n s and thermocouDles i n t he capsu le

o f FR2-Vg7 ( s c h e m a t i c d i a g r a m ) .

-56 -

eno

O g

O)

0)

0)3

<u

X3

CO

s_

3

C

so

c0)

c

O3

•f- O .

o eno

00o

CD

o"tno

3X¡

S_

V)

T3

(Q

"O

ce

Oí

vr>

I

ex

- 5 7 -

oo

2700

2500

2300

2100

1900

1700

1500

1300

1100

900

7000

f . f l .

\ \\ \ ^\ XA

\¡

•

1J , 1 5 % P U

30%Pu

\

\

\

\

\

0.2 0.L 0.6

r/r0

0.8 1.0

. 5: Radial teraperature d i s t r i b u t i o n at BOL inthe fue! pe l l e t of FR2-Vg7

( f . f l . LMFBR condi t ions,t h . f l . FR2 condit ions )

-58-

2600

2500

Oo

h-

2¿00

2300

22001.9¿ 1.95 1.96 1.97 1.98

0/M

1,99 ZOO

Fig. 6: Centerline and surface temperature of the fue! peiletand radial gap width at BOL as function of the 0/Mratio of the fue! in FR2-Vg7.

-59-

2600

Oo

2500

2 ¿00

2300

—•—'—"T^

\

\\

\

1100 * T 0.80

oo

900 — 0,60

800 -^ 0,50

oo

0.70°^O

Q.Oen

100 95 90 85

He content [%]

80

Fig. 7: Centeriine and surface temperature of the fue!peiiet and gap conductance at BOL as functionof the He content 1n the gap in FR2-Vg7.

-60-

DESIGN 0F IV0-FR2-Vg7-«fERíMENT

by

J. López Jiménez

JUNTA DE ENERGÍA NUCLEARDivisión de Metalurgia

Contents:

1. Introduction2. Thermal Design of the Irradiation Capsules3. Neutronics of the Experiment

3.1. Reactor Type3.2. Flux depression

4. Thermal and Structural Behaviour of Fue!

4.1. Thermal behaviour4.2. Structure analysis4.3. Stoichiometric profiles4.4. Porosity profiles4.5. Plutonium segregation4.6. Formation of columnar and equiaxed grains

5. Final Remarks

- 6 1 -

1. INTRODUCTION

The Karlsruhe Nuclear Research Centre (Projekt Schneller Brüter,PSB and Instituí für Material - und Fest-orperforschung, IMF)and the Junta de Energía Nuclear (División de M e t a l u r g i a ) havedecided to carry-out jointly a programme of fue! rod irradiationusing advanced mixed oxide fuel within the framework of theco-operation agreement existing between the two research centres,

The scope of this programme has been defined during a seriesof working meetings of joint commitees from both centres.

The IVO-Irratiation Programme will be conducted in the FR2experimental reactor in Karlsruhe. In t o t a l , twelve mixed -oxide fuel rods will be irradiated in four capsules.

2. THERMAL DESIGN OF THE IRRADIATION CAPSULES

The irradiation capsule consists of a series of concentricallyarranged metallic or gaseous m a t e r i a l s , which determine thetemperature to be obtained at the outer surface of the rod atthe established linear power.

At the given operation conditions: 450 W/cm and 200 W/cm linearrod p o w e r , and 520°C and 6 0 0 ° C , two different types ofcapsules are required.

Fig. 1 shows the sequence of the materials of the Type I-Capsuleand the radial temperature profile across this capsule. A NaK(78% K) mixture is used owing to its 1ow melting point (-11°C)and its good thermal properties. The highest temperature rise isachieved in this versión in the intermedíate zircaloy tube.

-62-

The temperature profile has been computed using the TECAPprogramme.

Fig. 2 shows the conditions for the Type 2-Capsule. In orderto obtain the necessary temperature rise, it was required tointroduce a gas gap containing helium.

3. NEUTRONICS OF THE EXPERIMENT

3.1. Reactor-Type

As stated before, the irradiation will be performed in the FR2reactor in Karlsruhe, This is a 44 MW thermal reactor with heavywater as moderator and coolant. In the moderator, the máximum

14 2neutrón fluxes, thermal, epithermal and fast, are of 1.1x10 n/cm12 23x10 n/cm .s (per lethargy interval; E> 0,5 eV) and

7 x l O 1 2 n/cm 2.s (E > 0,1 M e V ) , respectively.

Due to the self-shielding effect of the capsule materials andthe fue! rod against thermal neutrons, a flux depression existsacross the capsule.

3.2. Flux depression

The radial profile or the thermal neutrón flux inside the irra-diation capsule and in the fuel rod has been computed usingthe WIMS programme. The basic cell used consists of four driverfuel rods with the irradiation capsule in the center. Theprogramme considers 69 energy groups.

The mean fluxes in the different materials normalised to theundisturbed flux are given in Figs. 3 and 4, respectively.

-63-

Because we do not know what kind of piutonium composition willbe used for pellet fabrication, we have considered two possibili-ties: clean plutonium and dirty plutonium, both for 15% and30% enri ched fue!.

The fuel causing the most pronounced flux depression is 30% Pu(clean), foTlowed by 30% Pu (dirty), 15% Pu (clean) and 15% Pu(dirty).

A stronger lower depression is observed in the capsule of Type-2than in Type-1. That can be explained by the f a c f t h a t its interme-diate tube is Inox instead of Zy-2.

The ratio of the mean flux in the moderator to the one in thefuel isapproximately 3 or 5 for Pu contents of 15% and 3 0 % ,respecti vely.

Figs. 5 and 6 present the thermal flux depression in the fuelpellet for the four types of plutonium being studied for thecapsule of Type 1 and 2, respectively. There are no major diffe-rences between clean and dirty plutonium for the same content.The flux depression for 15% Pu is about 2 5 % , for 30% Pu about 5 0 % .

The desired linear rod power of the various fuel pins is obtai-ned by choosing appropriate irradiation positions in the reac-tor, taking into account the expected flux depression.

4. THERMAL AND STRUCTURAL BEHAVIOUR 0F THE FUEL

4.1. Thermal behaviour

The criterion governing the design of fuel rods for fast reac-tors establishes that the temperature of the fuel must remainbelow the melting poin.t.

The fusión temperature of the mixed oxide decreases as the contentof plutonium increases, so, for 15% Pu, the melting temperatureis 2760 °C and for 30%, 2690 °C.

-64-

The flux depression reduces the temperature in the center ofthe fue! peí 1et.

Recent calculations with the Saturn programme for the case ofrods in Versión 1- Capsule show the foliowing results:

- The temperature obtained in the center of the 30% enrichedfue! is about 2370°C, much lower than the fusión temperatureof 2690 °C.

- The temperature obtained in the center of the 15% enrichedfue! is about 2500 °C, also much lower than the fusión tempe-rature .

4.2. Structure analysis

A preliminary analysis of the fue! pellet has been performedwith the aid of the ÁTICO programme. This programme covers thefollowing phenomena and related influences: pore migration,formation of a central channel, oxigen migration, increase ofthe mean oxigen content as a result of fue! burnup ands finally,plutonium segregation through the evaporation-condensation andthermodiffusion mechanism.

Figs. 7 and 8 show the temperature profiles corresponding tothe Versión 1 - Capsule for plutonium contents of 15% and 30%,respectively. We nave calculated the temperature evolution upto 10000 h of operation at a power of 450 W/cm.

Under the simplified condition considered the temperature inthe center of 30% plutonium fue!, for example, decreases from2370 °C at the start of the irradiation to about 2300°C after100 hours and 2150°C at the end.

The decrease of temperature with time is explained by the for-mation of the central channel and by the improvement of thethermal conductivity due of the increase of the 0/Me-ratio.

- 6 5 -

For c o m p a r i s o n , the temperature profile which would be expectedunder fast flux is given in Figs 7 and 8 for the beginning ofthe irradiation for 15% and 3 0 % plutonium c o n t e n í , respectively.In these c a s e s , the center temperatures are 4% and 10% higheras compared to thermal f l u x , respectively.

4.3. Stoichiometric profiles

The same Figs 7 and 8 show the radial profiles for O/Me ratioafter a few hours of operation. It is worth to be mentioned thatthe valué at the outside of the pellet is cióse to 2.00.

4.4. Porosit.y profiles

Fig s . 9 and 10 represent the expected radial porosity profilesin the pellets with plutonium contents of 15% and 3 0 % , respec-ti v e 1 y .

The fue! is deemed to retain permanenly a 4% residual porosityin the calculation. Three well-defined porosity áreas are identi-fiel: one densified central á r e a , one non-restructured peripheralárea and a intermediate área. These three á r e a s , calculated witha pore migration m o d e l , correspond to the columnar grain, equia-xed-grain and non-restructured z o n e s .

4.5. Plutonium segreqation

Plutonium which initially is distributed "homogenously" thoughoutthe fuel pellet migrates towards the interior área of the pellet(the hotter área) through an e v a p o r a t i o n - c o n d e n s a t i o n mechanismas well as thermodiffusion through the solid phase. The firstof the mechanisms mentioned acts in short term, while the secondbecomes important for long irratiation t i m e s .

- 6 6 -

The plutonium concentrating in the center of the pellet elevatessomewhat its central temperature simultanously decreasing thefusión temperature of the mixed-oxide fue!.

4.6. Formation of columnar and equiaxed grains

The formation of the zones with columnar and equiaxed grains hasbeen calculated, in a d d i t i o n , using the COLEQ computer programme.

The evolution of the radii and temoeratures at the columnar andequiaxed-grain zone b o u n d a r i e s , r and r , has been estimatedin terms of the irradiation time for the case of Capsule-Type 1with 15% Pu fue! (Fig. 1 1 ) .

The boundaries of the columnar and equiaxed grain zones move towardthe outside of the pellet with irradiation time until they reachvalúes of r = 2.35 mm and of r 2.72 mm after 10.000 h o u r s s

cg eg

while the temperature reach 1640 °C and 1330°C, respectively.These results agree with the calculation in 4.4 (Fig. 9 ) .

5. FINAL REMARKS

The IV0 experiment comprises a broad range of activities withinthe fast reactor technology, particulary in the field of fue!elements development. The tasks cover, bes i des the design andphilosophy of the irradiation exneriment as a whole the prepa-ration of capsules and fue! r o d s , quality c o n t r o l , assembling,i r r a d i a t i o n , post-irradiation examination and interpretationof the results .

The Junta de Energía Nuclear (JEN) will contribute to all thesestages with its staff and physical m e a n s , according with anestablished programme which extends in time from January 1978to the end of 1981. Particularly in the present stage, the JENlaboratories (Metallurgy D i v i s i ó n ) are engaged in the fabricationof two Versión 1 - c a p s u l e s , with all the quality control functionssuch as X-ray and ultrasound t e s t s , sea! t i g h t n e s s , etc.

-67-

1 Fuel pellet 6,4mm. $2 Cladding (S.S. 1.4970) 7,6mm. $ x 0,5mm.3 Nak- gap 1,25mm.4 Anticonvection pipe (S.S. 1.4571) 10.5mm.<f>xO,2mm5 Nak - gap 1,25 mm.6 Intermedióte pipe (Zy- 2) 19mm. 4 x 3mm.7 Nak-gap 2,5 mm.8 Capsule pipe (S.S. 1.4571) 27mm. <j> x 1,5 mm.9 Coolant D2 O (50°C)

700 t-

600+-

500 4-

4001-

oo

300 4-

2001-

100 4-

R (mm)

FIG. 1 '• Radial temperature p r o f i l e at the Capsule (Vers ión 1 ) )Z = 4 5 0 w / c m .

- 68 -

1 Fuel pellet 6,4 mm.2 Clodding ( S.S. 1.4970) 7,6mm. <j> x 0,5mm.3 Nak - gap 1,8 mm.4 Intermedióte pipe (S.S. t.4571)13,2mm. $xlmm.5 Heliíum-gap 200 pm (en caliente)6 Intermedíate pipe (S.S. 1.4571)19mm.$x2,7mm.7 Nak-gap 2,5 mm.8 Capsule pipe (S.S. 1. 4571) 27mm. • x 1,5 mm.9 Coolant D2 0 (50°C)

oo

700-

600"

500-

400-

300-

200-

IOO-

1 2 3 4 5 6

\¡

7 8 9

10 15

R ( m m )

FÍG. 2 : Radia l t e m p e r a t u r e p r o f i l e a t t h e Capsule ( V e r s i ó n 2 ) , Z = 2 0 0 w / c m

-69-1 Fuel pellet2 Cladding ( S.S.l .4970)3 NaK - gap4 Anticonvection pipe ( S.S . 1 . 4571)5 NaK-gap6 Intermedióte p ipe(Zy-2)7 NaK-gap8 Capsule (S .S .1 .4571)9 Coolant D2 O (50°C)

15% Pu Dtrty

15% Pu Clean

30% Pu Dirty

30% Pu Clean

o 2

5

1

-

•

-

-

•

2

—

3 4 5 6 7 8 9

(50°C)

10 15

R (m m)

18

Fig 3 : Thermal neutrón flux depression at the Capsule (Versión 1) in FR2

-70 -

1 Fue! pel le t

2 C l a d d i n g ( S.S. 1.4970 )

3 Na K - gap4 Intermedíate pipe (S.S. 1.4571)5 Hellium - gap

6 Intermedíate pipe (S.S. 1.4571)7 NaK-gap3 Capsule pipe (S.S. 1.4571 )•9 Coolant D2 O ( 50°C)

15% Pu Dirty

15% Pu Clean

-30% Pu Dírty

3 0 % Pu Clean

co

o oe ¿É 3

1

-

. _ . __.—_. _ . . _

-

2 3 4 5 6 7 8 9

( 5 0 ° C )

10 15 18

R (mm)-

FIG.4 l Thermal neutrón flux depression at the Capsule (Versión 2) ' in FR2

- 7 1 -

4 - -

o

>

II* '-e

CAPSULE VERSIÓN 1

1 15% Pu Dirty

2 15% Pu Clean

3 30% Pu Dirty

4 30% Pu Clean

R (mm)

FIG. 5 : Thermal f lux depression in fuel rod at Capsule (Versión 1)

-72 -

aX 3

£ "a

i!

CAPSULE VERSIÓN 2

1 15% Pu Dirty

2 15% Pu Ciean

3 30%Pu Dirty

4 30%Pu Ciean

R (rain)

F1G. 6 : Thermal flux depression in fuel rod at Capsule (Versión 2)

-73 -

2.500-

2.000--

. 5 0 0 "

.000--

CAPSULE VERSIÓN 1

3 0 % Pu

Oí =450w/cm

No Flux depression

Flux depression

-2.00

R (mm)

Fig 7 : Radial temperature and stoichiometric profile in the fuel rod

-74-

2.500.

2.ooa-

I.000--

CAPSULE VERSIÓN 1

15% Pu

y -- 450w/cm

No Flux depression

Flux depression

..2.00

R(mm>-

Fig.8: Radial temperature and stoichiometric profile in the fuel rod

- 7 5 -

3O.

0,25.

0,20 .

0,15 .

0,10

oo

<rHUJ

o

CAPSULE

\

PELLET

VERSIÓN

15 %

6 %

X s

P

)

Jí

rcg.

1

Pu

Porosity

450 W/cm.

íreg.

1

2?"al

1 2

R ( m m . )

FIG. 9 ; Radial piutonium and porosity profile m the fuel rod

•76-

o.4a.

0.35-

o.0.301

0.25..

*Pu

CAPSULE VERSIÓN 1

PELLET

30 % Pu

6 % Porosity

X = 450 W/cm.

. 7

-. 4

- • 3

R ( mm )

RG.10 : Radial piutonium and porosity profile in the fuel rod

CAPSULE VERSIÓN 1

3 -

Ee

2

15% Pu

X = 4 50 w/cm

0/MB = 1.97

2.000

oo

- 1.500

- 1.000

10 100

Irradiation time (h) - -••

1.000 10,000

FIG. 11 LocQÜon and temperature of the columnar and equiaxed grain región in fuel rodas a function of i rradiat ion time

J.E.N. 454

Junta de Energía Nuclear. División de Metalurgia. Madrid."Es tado ac tual del exper imen to de i r r a d i a c i ó n de b a -

rras combustibles para reactores rápidos IVO-FR2-Vg7l'OTERÜ DE LA GÁNDARA, J .L . ; KUMMERER, K.; BOJARSKY, K.; ELBEL, I I . ; LÓPEZ JIMÉNEZ, J .

(1979) 77 pp. 11 f igs .

Informe sobre el Seminario celebrado en Madrid entre KFK (Karlsruhe) y la JEN (Ma-

drid) relativo a un programa conjunto de irradiación de barras combustibles para reac-

tores rápidos. Se define el diseño de barras en general, y, en particular, de aquellas

a irradiar de densidad 9'$ DT y diámetro 7,0 mm hasta un quemado del T/> FIMA. Junto con

el diseño de las cápsulas de NaK y pared única empleadas en la irradiación, que tendrá

lugar en el reactor FR2 de Karlsruhe, se describen otras posibilidades de irradiación

del reactor.

CLASIFICACIÓN INIS Y DESCRIPTORES: B25; E23. Fuel rods. I r rad iaron. FR-2 reactor.

Fast reactors.

J.E.N. 454

Junta de Energía Nuclear. División de Metalurgia. Madrid."Es tado ac tual del exper imento de i r r ad i ac ión de b a -

rras combustibles para reactores rápidos IVO-FR2-Vg7'.'OTERO DE LA GÁNDARA, J .L . ; KUMMERER, K.; BOJARSKY, K.; ELBEL, H.; LÓPEZ JIMÉNEZ, J .

(1979) 77 pp. 11 f igs .

Informe sobre el Seminario celebrado en Madrid entre KFK (Karlsruhe) y la JEN (Ma-

drid) relat ivo a un programa conjunto de irradiación de barras combustibles para reac-

tores rápidos. Se define el diseño de barras en general, y, en particular, de aquellas

a irradiar de densidad 9'$ DT y diámetro 7,6 mm hasta un quemado del 1% FIMA. Junto con

el diseño de las cápsulas de NaK y pared única empleadas en la irradiación, que tendrá

lugar en el reactor FR2 de Karlsruhe, se describen otras posibilidades de irradiación

del reactor.

CLASIFICACIÓN INIS Y DESCRIPTORES: B25; E23. Fuel rods. Irradiat ion. FR-2 reactors.

Fast reactors.

J.E.N. 454

Junta de Energía Nuclear. División de Metalurgia. Madrid."Es tado actual del exper imento de i r r a d i a c i ó n de b a -

rras combustibles para reactores rápidos IVO-FRZOTERO DE LA GÁNDARA, J .L . ; KUMMERER, K.; BOJARSKY, K.; ELBEL, I I . ; LÓPEZ JIMÉNEZ, J .

(1979) 77 pp. 11 f igs .

Informe sobre el Seminario celebrado en Madrid entre KFK (Karlsruhe) y la JEN (Ma-

drid) relativo a un programa conjunto de irradiación de barras combustibles para reac-

tores rápidos. Se define el diseño de barras en general, y, en particular, de aquellas

a irradiar de densidad 94$ DT y diámetro 7,6 mm hasta un quemado del T% FIMA. Junto con¡

el diseño de las cápsulas de NaK y pared única empleadas en la irradiación que tendrá

lugar en el reactor FR2 de Karlsruhe, se describen otras posibilidades de irradiación

del reactor.

CLASIFICACIÓN INIS Y DESCRIPTORES: B25; E23. Fuel rods. Irradiat ion, FR-2 reactor.Fast reactors.

J.E.N. 454

Junta de Energía Nuclear. División de Metalurgia. Madrid."Es tado ac tua l del exper imen to de i r r a d i a c i ó n de b a -

rras combustibles para reactores rápidos IVO-FR2-Vg7'.'OTERO DE LA GÁNDARA, J .L . ; KUMMERER, K.; BOJARSKY, K.; ELBEL, H.; LÓPEZ JIMÉNEZ, J .

(1979) 77 pp. 11 f igs .

Informe sobre el Seminario celebrado en Madrid entre KFK (Karlsruhe) y la JEN (Ma-

drid) relat ivo a un programa conjunto de irradiación de barras combustibles para reac-

tores rápidos. Se define el diseño de barras en general, y, en particular, de aquellas

a irradiar de densidad §k% DT y diámetro 7,6 mm hasta un quemado del 7% FIMA. Junto con

el diseño de las cápsulas de NaK y pared única empleadas en l a irradiación que tendrá

lugar en el reactor FR2 de Karlsruhe, se describen otras posibilidades de irradiación

del reactor.

CLASIFICACIÓN INIS Y DESCRIPTORES: B25; E23, Fuel rods. Irradiat ion. FR-2 reactor.Fast reactors.

J.E.N. 454

Junta de Energía Nuclear. División de Metalurgia. Madrid."Status oí IVO-FR2-Vg7- exper iment for i r rad ia t ion

of fast r eac to r fuel r o d s " .OTERO DE LA GÁNDARA, J .L . ; KUMMERER, K.; BOJARSKY, K.; ELBEL, I I . ; LÓPEZ JIMÉNEZ, J .

(1979) 77 pp. 11 f i gs .

Report on the Seminar celebrated in Madrid between KfK (Karlsruhe) and JEN (Madrid)

conceming a Joint Irradiation Program of Fast Reactor Fuel Rods.

I t is defined the design of fuel rods in general, and, in particular of those with a

density 9'$ DT and diameter 7.6 mm up to a bum-up of 1% FIMA, to be irradiated in the

FR2 Reactor (Karlsruhe). Together with the design of NaK and single-wall capsules used

in this i r radiat ion, other possibi l i t ies of irradiat ion in the nactor wl l l also be

described.

INIS CLASSIFICATIÓN AND DESCRIPTORS: B25; E23. Fuel rods. Irradiat ion. FR-2 reactor.

Fast reactors.

J.E.N. 454

Junta de Energía Nuclear. División de Metalurgia. Madrid."Status of IVO-FR2-Vg7- exper iment for i r rad ia t ion

of fast r e ac to r fuel r o d s " .OTERO DE LA GÁNDARA, J .L . ; KUMMERER, K.; BOJARSKY, K.; ELBEL, I I . ; LÓPEZ JIMÉNEZ, J .

(1979) 77 pp. 11 f i gs .

Report on the Seminar celebrated in Madrid between KfK (Karlsruhe) and JEN (Madrid)

conceming a Joint Irradiation Program of Fast Reactor Fuel Rods.

I t is defined the design of fuel rods In general, and, in particular of those with ai

density 9'$ DT and diameter 7.6 mm up to a bum-up of 1% FIMA, to be irradiated in th

FR2 Reactor (Karlsruhe). Together with the design of NaK and single-wall capsules used

in this i r radiat ion, other possib i l i t ios of irradiat ion in the reactor wi l l also be

doscribed.

INIS CLASSIFI CATIÓN AND ESCRIPTORS: B25; E23. Fuel rods. Irradiat ion. FR-2 reactor.

Fast reactors.

J.E.N. 454

Junta de Energía Nuclear. División de Metalurgia. Madrid.

"Status of IVO-FR2-Vg7- experiment for irradiationof fast reactor fuel rods".OTERO DE LA GÁNDARA, J .L . ; KUMMERER, K.; BOJARSKY, K.; ELBEL, H.; LÓPEZ JIMÉNEZ, J .

(1979) 77 pp. 11 f i gs .

Report on the Seminar celebrated in Madrid between KfK (Karlsruhe) and JEN (Madrid)

conceming a Joint Irradiation Program of Fast Reactor Fuel Rods.

I t is defined the design of fuel rods in general, and, in particular of those with a!

density 9'$ DT and diameter 7.6 mm up to a burn-up of 1% FIMA, to be irradiated in the

FR2 Reactor (Karlsruhe). Together with the design of NaK and single-wall capsules used

in this i rradiat ion, other possibi l i t ies of irradiation in the reactor wi l l also be

described.

INIS CLASSIFICATION AND DESCRIPTORS: B25; E23. Fuel rods. I rradiat ion. FR-2 reactor.

Fast reactors.

J.E.N. 454

Junta de Energía Nuclear. División de Metalurgia. Madrid.

"Status of IVO-FR2-Vg7- experiment for irradiation

OTERO DE LA GÁNDARA, J .L . ; KUMMERER, K.'; BOJARSKY, K.; ELBEL, I I . ; LÓPEZ JIMÉNEZ, J .

(1979) 77 pp. 11 f i gs .

Report on the Seminar celebrated in Madrid between KfK (Karlsruhe) and JEN (Madrid)

concerníng a Joint Irradiation Program of Fast Reactor. Fuel Rods.

I t is defined the design of fuel rods in general, and, in part icular of those witha

density 9'$ DT and diameter 7.6 mm up to a bum-up of % FIMA, to be irradiated in the

FR2 Reactor (Karlsruhe). Together with the design of NaK and single-wall capsule used

in th is i rradiat ion, other possibi l i t ies of irradiat ion in the psactor wi l l also be

described.

INIS CLASSIFICATION AND DESCRIPTORS: B25; E23. Fuel rods. I rradiat ion. FR-2 reactor.

Fast reactors.