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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
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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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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:
- 20-
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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Etiali lartrnlll darchlakhmmgi Nr
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.
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