t NEA COMMITTEEON REACTOR PHYSICS !

REACTOR PHYSICS ACTIVITIES IN OECD COUNTRIES

October 1979-September 1980

OECD NUCLEAR ENERGY AGENCY 38 boulevard Suchet 75016 Paris

NEA COMMITTEE ON REACTOR P H Y S I C S

REACTOR P H Y S I C S A C T I V I T I E S I N

NEA MEMBER COUNTRIES

O c t o b e r 1979 - ~ e ~ t e m b e r 1980

OECD NUCLEAR ENERGY AGENCY 38 B o u l e v a r d Suchet, 75016 P A R I S

REACTOR PHYSICS ACTIVITIES IN NEA MEMBER COUNTRIES

This document is presented to the Reactor Physics. Idaho. from 22nd

a compilation of national activity reports Twenty-Third Meeting of the NEA Committee on held at Argonne National Laboratory.West. to 26th September 1980 .

Australia Austria ................................... 3 Belgium ................................... 11 Canada ................................... 22 Denmark ................................... 24 Finland ................................... 32 France ................................... 37 F.R. Germany ................................... 47 Italy ................................... 93 Japan ................................... 100 Netherlands ................................... 129 Norway ................................... 138 Spain ................................... 148 Sweden ................................... 160 Switzerland ................................... 166 United Kingdom ................................... 173 United States ................................... 186 JRC-Ispra ................................... 193

RERCMR PHYSICS ACTIVITIES I N AUSTRALIA

October 1979 - September 1980

D.B. MCCULLOCH

Australian Atomic Energy Commission Research Establishment Lucas Heights, New South Wales, Australia

1. REACTOR CODES DEVE,LOPMENT . Work has continued on POW-3D,the three-dimensional diffusion theory

'work-horse' module of the AUS scheme.

The e s sen t i a l mathematical rout ines to solve the large sparse system of l i nea r equations have been completed and extensively tested. Three methods capable of solving such systems a r e available:- Successive Line Overrelaxation (SLOR), Method of Incomplete Conjugate Gradients (ICCG) and Method of Implici t Nonstationary I t e r a t ion ( M I N I ) . Tests show t h a t ICCG is s l i g h t l y superior to M I N I i n the use of machine time f o r some problems, but addi t ional 1/0 a c t i v i t y is required, par t icu lar ly fo r the three-dimensional problem. Convergence of group equations is ass i s ted through the use of M I N I .

The code is serving a s a t e s t vehicle to study the e f f e c t of var ia t iona l methods a s a secondary means of accelerat ing convergence of large l i nea r systems. The dis junct ive par t i t ioning and dis junct ive weighting (DPDW) coarse mesh rebalancing lnethod used e a r l i e r in the two dimensional code POW

.was successfully extended t o three-dimensions, and has proved compatible w i t h a l l th ree i t e r a t i v e schemes.

~ w o more sophisticated systems a r e being tes ted. One involves a multiplicative pyramid correction fonn combined w i t h d is junct ive weighting (MPDw), and the other an addi t ive pyramid with dis junct ive weighting (APDW). MPDW involves considerable overhead but provides much superior performance t o APDW on ce r t a in t e s t problems. Compared with DPDW, however, the addi t ional overheads appear to date t o outweigh any improvement i n convergence, but this s i tua t ion may well be reversed when coarse mesh rebalancing i s applied f i l l y t o

b the eigen value problem

Although both new schemes seem t o be compatible with the three i t e r a t i v e methods on a number of t e s t problems, DPDW leads t o a smaller system of . equations t h a t is e f f i c i e n t l y solved with M I N I , while MPDW and APDW f a i l i n general to preserve the mathematical propert ies of t h e or ig ina l system, leaving d i r e c t methods a s the most e f f i c i e n t means of solution of the reduced System.

2. BURNUP METHODS

The burnup methods used within the ALjS scheme a r e current ly being upgraded. To date , burnup calculat ions have re l ied on CHAR wh.ich is a multi-region burnup module using an ana ly t ic method t o solve the nuclide depletion equations. CHAR has been applied i n the past mainly t o l a t t i c e burnup calculations. Its appl icabi l i ty t o global calculat ions was l imited to few-region calculat ions by the necessi ty i n most reactor types t o perform subsidiary l a t t i c e calculat ions a t each time s t ep for each region.

Although t h e methods development i s being done i n conjunction with an inves t iga t ion of burnup modelling i n l a r g e f a s t r e a c t o r s , requirements f o r thermal r e a c t o r burnup a r e a l s o being considered. The inves t iga t ion inc ludes t h e e f f e c t s of spec t ra used i n group condensation, mesh i n t e r v a l s , d i f fus ion theory versus SN method, he terogeni ty , number of regions with cons tan t burnup, time s t e p , v a r i a t i o n of i so tope c r o s s s e c t i o n s with i r r a d i a t i o n , and f i s s i o n product representa t ion .

The changes made t o t h e AUS scheme include:

(a) provis ion o f e d i t i n g f a c i l i t i e s i n CHAR,

(b ) allowing i so tope c r o s s sec t ions t o be i r r a d i a t i o n dependent which extends the use of CHAR i n g lobal c a l c u l a t i o n s ,

( c ) provision f o r energy condensation of i so topes over spec t ra from a g lobal ca lcu la t ion , and

(d ) add i t ion of a module which group-condenses t h e main c r o s s sec t ion l i b r a r y .

A new f i s s i o n product l i b r a r y is being generated from ENDFB using t h e OWL code XLACS. Some changes t o the XLACS resonance t rea tment were made t o reduce the required computer time.

A simple burnup module,BUWMAC,which adopts t h e usual assumption t h a t macroscopic l a t t i c e d a t a may be tabula ted aga ins t i r r a d i a t i o n has a l s o been w r i t t e n f o r AUS.

3. GROUP CROSS SECTION LIBRARY

Because a v a i l a b i l i t y of ENDFBV da ta is r e s t r i c t e d and t h e r e i s a l a r g e d i s c r e ancy i n 2 3 8 ~ resonance captures using ENDFB E, a modified ENDFB f i l e f o r 23iiU has been formed using t h e resonance da ta e z u a t e d by de Saussure e t a l t f o r ENDFBV. This modified da ta f i l e was used t o prepare new AUS c r o s s s e c t i o n s f z r 2 3 8 ~ which were t e s t e d i n AUE; c a l c u l a t i o n s o f t h e TRX-1 l a t t i c e experiment. Compared with E N D F B ~ data,keff increased by 0.2% and p2* decreased by 1%, s t i l l leaving a?% e r r o r i n p 2 @ compared with experiment. The matter was n o t pursued f u r t h e r because t h e change was r e l a t i v e l y small .

i de Saussure G . , Olsen D.K. , Perez R.B. and D i f i l i p p F.C. - ORNL/TM-6152

NEACRP-L-244 AUSTRIA

REACTOR PHYSICS ACTIVITIES IN AUSTRIA

September 1979 - September 1980

compiled by

B. Putz -

List of contributing organizations:

Atominstitut der Gsterreichischen Universitaten, Wien (AI)

Institut fiir Theoretische Physik der Technischen Universitlt Graz (ITE/TU Graz)

Institut fiir Theoretische Physik der Universitlt Innsbruck (ITP/U Innsbruck)

Gsterreichisches Forschungszentrum Seibersdorf Ges.m.b.H. (former Gsterreichische Studiengesellschaft fiir Atomenergie Ges.m.b.H.) (FZS)

1. REACTOR THEORY

1.1 Reactor Analysis

The investigations carried out at FZS on the consequences

a of the reduction of the fuel enrichment in the research reactor ASTRA have been completed. The lower enrichment

necessitates a higher total uranium inventory per fuel

element which may entail metallurgical problems.

. Criticality calculations have been performed at the same institute for the storage of BWR fuel elements in a high

density fuel rack made of boronated steel. The objective

of the studies had been the dependence of the results on

different calculational methods and on important design

parameters such as the concentration of boron, the thick-

ness of the boronated steel plates and the width of the

watergap /I/. ,

The neu t ron f l u x d i s t r i b u t i o n i n mu l t i sphe re c o n f i g u r a t i o n s

h a s been s t u d i e d a t ITP/TU Graz. The f l u x modulat ion by

t h e i n f l u e n c e o f ne ighbour ing sphe re s has been t aken i n t o

account by de t e rmin ing t h e f l u x o f a p o i n t s o u r c e o f neu-

t r o n s i n a n i n f i n i t e medium, which c o n t a i n s a s p h e r i c a l

p e . r t u r b a t i o n zone e c c e n t r i c t o t h e p o i n t s o u r c e . An i t e r a -

t i o n method a l l ows c o n t i n u a l l y improving approximat ions .

/ 2 / .

An a t t e m p t w a s made a t t h e above : i n s t i t u t e t o f i n d o u t

whether t h e i d e a l i z a t i o n o f a s p h e r i c a l u n i t ce l l i s

j u s t i f i e d o r n o t when c a l c u l a t i n g t h e : . n e u t r d n _ . s p e c t r ~ m .

o f pebble-bed co re s . To t h i s end t h e f i n e s t r u c t u r e

o f t h e f l u x d i s t r i b u t i o n i n a s p h e r i c a l f u e l e lement

and i n i t s ambient medium h a s b e e n determined u s i n g

i n t e g r a l t r a n s p o r t t h e o r y f3 / .

The c r i t i c a l masses f o r f u e l e lements o f r e s e a r c h r e a c t o r s

haye been c a l c u l a t e d a t ITPfTU Griiz f o r medium and h i g h

enr ichment and compared w i t h lowly e n r i c h e d uranium

dioxide-water-systems i n o r d e r t o i n v e s t i g a t e t h e i n f l u -

ence o f t h e h e t e r o g e n e i t y and o f t h e c l a d d i n g on t h e

c r i t i c a l i t y 1 4 J . a

S t u d i e s underway a t pZS on t h e method of Doppler we igh t ing ,

which e n a b l e s space dependent t empera ture e f f e c t s be ing . t aken i n t o account i n t h e p o i n t k i n e t i c s e q u a t i o n s , have

cont inued .

Resonance Absorption

The less significant and usually neglected temperature

dependence of resonances in the thermal region has

been investigated at FZS comparing the results obtained

with the y-J-formalism to those of a more exact

formula /5 / .

Neutron Thermalization

The development of a thermalization method combining

Selengut's method of overlapping neutron spectra with

the multicollision probability method has been completed

at FZS. The new method is suitable to a wider field of

application, especially to the homogenization of reactor

cells. Subdividing the cell into N regions leads to a

system of 2 N~ unknowns for the neutron currents. By

a recurrence formalism this system can be reduced to a

system with only N unknowns resulting in a considerable

saving of computer time / 6 / .

Synergetic Fusion-Fission Systems

Study of synergetic fusion-fission systems has continued

to be an important activity at ITP/TU Graz and at ITP/U

Innsbruck.

One of the topics dealt with at TU Graz is the impact of the integration of conventional fission reactors with

non-fission neutron sources (such as spallation neutrons

and fusion neutron sources) on the overal thermal-to-

electric conversion efficiency of the system /7/. Some

effort were directed toward the identification of a set

of efficiency merit parameters which more accurately

assesses conventional and synergetic nuclear energy systems

/ a / -

I n v e s t i g a t i o n s c a r r i e d o u t a t TU Graz on t h e mathe- '

m a t i c a l - p h y s i c a l s i m u l a r i t i e s and d i f f e r e n c e s between

f u s i o n and f i s s i o n m u l t i p l i c a t i o n p r o c e s s e s showed

t h a t advanced f u s i o n c y c l e s can s u s t a i n e x c u r s i o n

t e n d e n c i e s e s s e n t i a l l y analogous t o c o n v e n t i o n a l

f i s s i o n c y c l e s /9 / .

The energy break-even c o n d i t i o n s of a f u s i o n - f i s s i o n r e a c t o r system, i n which t h e f u s i o n d e v i c e i s f u e l e d

w i t h deu te r ium o n l y and d r i v e n by n e u t r a l beam i n j e c t i o n ,

were s t u d i e d a t U Innsbruck . The i n t e r r e l a t i o n s h i p be t -

ween t h e f u s i o n neu t ron p roduc t ion r a t e , t h e plasma f u s i o n

g a i n and t h e p roduc t n e T E , and p a r t i c u l a r i l y t h e . e f f e c t

o f t h e i n j e c t e d h i g h e n e r g e t i c d u e t e r o n s on t h e s e para-.

m e t e r s were examined. The r e s u l t s i n d i c a t e t h a t even

t h e D-D f u s i o n p r o c e s s may be viewed a s 3 neut ron sou rce

s u f f i c i e n t t o d r i v e a s u b c r i t i c a l E i s s i o n / c o n v e r s i o n

assembly / lo( .

A t ITP/U Innsbruck a n a l y t i c a l app rox ima t ions have been

developed f ~ r t h e c h a r a c t e r i s t i c p a r a m e t e r s o f a hybr id

b r e e d e r , which e n a b l e s a f u l l y a n a l y t i c a l d e s c r i p t i o n

o f t h e b l a n k e t performance va ry ing w i t h f u e l r e s i d e n c e

t ime . A t t e n t i o n h a s been p l a c e d On the i n f l u e n c e of f i s s i l e f u e l enr ichment and on t h e b u i l d up o f f i s s i o n

p rqduc t s [11[.

EXPERIMENTAL REACTOR PHYSICS

Water I n g r e s s i n t o Graph i t e Assemblies

Using t h e r e a c t o r code GAMTEREX t h e l a y o u t of exper iments

h a s been determined a t ITP/TUGraz aimed a t s t u d y i n g t h e

w a t e r i n g r e s s i n t o pebble beds o f AVR f u e l e lements . These

experiments are planned to be performed at the siemens-

Argonout-Reactor (SAR) in Graz and first measurements

of reaction rates in "dry" pebble beds are provided as

preliminary tests for the experiments at the "wet"

core /12/.

At the same institute some effort has been devoted to

investigitions on how insertion of water in the internal

graphite reflector of the SAR influences the reactivity

/ 13 / .

0 2.2 Neutron Flux Control

A code enabling the adjustment of a constant neutron

flux in a research reactor has been written at ITP/TU

Graz for the microprocessor MC 6800 /14/.

2.3 Neutron Spectrum

The fast neutron emission spectrum of 252~f has been

investigated at A1 by means of proton recoil spectro-

meters. With a large counter tube of 900 mrn length the

neutron distribution between 0.9 MeV and 10 MeV could

be determined. Monte Carlo calculated response functions

were applied to infold the measured proton recoil

distributions. The energy interval between 1 MeV and

3 MeV had been examined with a smaller tube (466 man)

in a search for neutron fine-structure groups. No such

groups could be established. . 2.4 Cross Section Measurement

As an application of photoneutron sources absolute

measurements of the absorption cross section of tan-

talum were performed at A1 using a transmission method

and a long counter as neutron detector. The neutron

e n e r g i e s were 20.9 keV, 121.8 keV, 215.3 keV and

837.8 keV. The r e s u l t s may, i n p r i n c i p l e ; be u s e d ' a s

r e f e r e n c e v a l u e s f o r r e l a t i v e measurements.

2 . 5 Delayed Neutron Measurement

A method u s i n g a s o l i d s t a t e n u c l e a r t r a c k d e t e c t o r

h a s been developed a t A 1 t o t e t e c t d e l a y e d n e u t r o n s

from f i s s i o n p r o d u c t s con ta ined i n t h e p r imary c o o l a n t

o f a n u c l e a r r e a c t o r . I n an in -core l o o p o f t h e TRIGA

r e a c t o r Vienna a sma l l sample o f 93% e n r i c h e d uranium

was i r r a d i a t e d and t h e f i s s i o n p r o d u c t s w e r e t r a n s -

p o r t e d by a purg ing system t o t h e t r a c k d e t e c t o r . A

c o r r e l a t i o n could be o b t a i n e d between t h e number o f

t r a c k s and t h e r e a c t o r power and t h r e e g r o u p s o f delayed

n e u t r o n s w e r e i d e n t i f i e d /15 / .

3. GENERAL

A d e t a i l e d su rvey h a s been worked o u t on t h e e x p e r i -

mental and t h e o r e t i c a l s t u d i e s t h a t have been performed

a t ITP/TU Graz between 1973 and 1979 conce rn ing i n v e s t i -

g a t i o n s on t h e n u c l e a r p h y s i c a l behav iou r o f wa te r

moderated pebble beds and t h e i r E e a s i b i l i t y f o r power

p l a n t s / I 6/.

Based on t h e e q u i v a l e n t f u e l concep t and t h e f u e l s t o c k p i l e

concept t h e f i s s i l e f u e l t r a j e c t o r y c o n c e p t were developed

and a p p l i e d t o b u r n e r , c o n v e r t e r and b r e e d e r r e a c t o r s / l 7 / . .

REFERENCES :

F. WOLOCH, G. SDOUZ, M. SUDA, Neutronenphysikalische Aspekte der NaRlagerung von SWR-BE-Bundeln. ATKE - 35, 166 (1980).

F. SCHuRRER, A Diffusion-Theoretical Method to Cal- culate the Neutron Flux Distribution in Multisphere Configurations. ATKE - 35, 179 (1980). F. SCHORRER, Successive Approximation of the Neutron Flux Distribution in Spherical Configurations. Acta Physics Austriaca (in the press).

H. MULLER, H. RABITSCH, F. SCHtiRRER, The Criticality of Water Reflected Homogeneous Arrays and of Heterogeneous Reactor Fuel Elements. Acta Physica Austriaca (in the press).

G. KAMELANDER, Reactor Physical Effects of Thermal Resonances. ATKE (in the press).

G. KAMELANDER, F. PUTZ, Application of the Multigroup Collision Probability Method to Selengut's Theory.of Overlapping Neutron Spectra. Nuc1.S~. Eng. - 74, 13 (1980). M. HENDLER, A.A. HARMS, The Efficiency Decrement of Self-sufficient Nuclear Energy Systems. Trans. Am. Nucl. Soc. - 33, 785 (1979). M. HEINDLER, A.A. HARMS, Efficiency Merit Assessment of emerging Synergetic Nuclear Energy Systems. ATKE - 36, 7 (1980).

A.A. HARMS, M. HEINDLER, The Existence and Characteri- zation of Self-sustaining Multiplicative Fuston and Fission Reaction Chains. Acta Physica Austriaca - 52 CDec. 1980) (in the press) . K.F. SCH~PF, Beam Driven D-Fusion Plasma within a Fusion-Fission Hybrid System. ATKE - 36, 26 (1980).

. - 1 K. SCHBPF, G. STRASSER, Analytical Description of the Fuel Dynamics in a Hybrid Fusion Breeder. ATKE (in the press).

/12/ F. SCHtiRRER, First Series of Measurements to the Project "Water Ingress into Pebble Beds of AVR Fuel Elements". Interal Report ITPR-79009, TU Graz 1979.

/13/ Hj. MuLLER, W. NINAUS, K.OSWALD, Changes in Reactivity by Insertion of Water in the Internal Graphite Reflector of the Argonaut. Acta Physica Austriaca (in the press).

/14/ W. NINAUS, G . KAHR, Neut ron F l u x C o n t r o l o f a ~ e s e a r c h R e a c t o r by a Microcomputer System. A c t a P h y s i c a A u s t r i a c a ( i n t h e p r e s s ) .

/15/ H. BBCK, D e t e c t i o n o f Delayed N e u t r o n s i n a N u c l e a r R e a c t o r Using t h e S o l i d S t a t e Track E t c h T e c h n i q u e . P a p e r p r e s e n t e d a t t h e 1 0 t h Int . .Conf. on S o l i d S t a t e N u c l e a r Track D e t e c t o r s , 2nd-7th J u l y 1 9 7 9 , Lyon, F r a n c e .

/ 1 6 / E. LEDINEGG, M. HEINDLER, Hj. MULLER, W. NINAUS, H . RBBITSCH, F . SCHuRRER, N u c l e a r P h y s i c a l Behav iour o f Water Moderated P e b b l e Beds. I n t e r n a l r e p o r t ITPR- 79010, T U Graz , 1979.

/17/ A.A.HARMS, M . HEINDLER, L i f e t i m e F u e l T r a j e c t o r i e s f o r F i s s i o n R e a c t o r s . T r a n s . An,. Nucl . SO;. - 3 3 , 125 ( 1 9 7 9 ) .

WACRP-L-24 4 BELGIUM

REACTOR PYYSICS ACTIVITIES IN BELGIUM

Progress report to the NEA Committee on Reactor Physics

Compiled by J. DEBRUE, SCK-CEN, Mol

Septeaber 1980

EiERMAL REACTORS

1 ,. Fuel Cycle'

a) Low power experiments in the VENUS criticality facility ....................................................... The variation of the reactivity of Pu02-U02 fuel ccnfigurations over

long periods of time (w 10 ears) was further investigated. This

variation is due to the 241~m build-u? resulting from the natural decay

of 241p~ (half-life : 14.4 years), The theoretical analysis has been

performed with the DLC 43B/CSRL cross-section library (218 groups, P3)

based on ENDF/B IV. The code packages AMPX-I1 A and MARS have been

used to produce 68 group cross-section sets which allowed to calculate '

weighting spectra in the different regions of the loadings by means

of ANISN; collapsing in 7 groups was finally made to perform 2 dinen- sional XY calculations with DOT 3,5. Comparing the calculated keff

with the experimental values, as obtained from 1969 to 1979 for near

critical configurations (adjustment is made by adding peripheral rods), 241 . indicates that the neutron capture in Am could be overestimated in

the calculation by about 10 to 20 %. This assuves that the effect of 241

the Pu decay is exactly calculated i,e, that the 241 Pu cross-sections

, in the thermal and resonance energy range are correct in ENDF/B IV.

However, according to the recent 241

Pu capture cross-section measurement

by Weston and Todd [I], the resonance capture for this isotope is significantly underestimated in ENDF/B IV. Taking into account the

Weston and Todd date, the calculations agree with thc criticzlity

measurements in VENUS within the experimental error margin.

b) Power reactor calculations .......................... - The application of the calculation methods currently in use at Electrobel for three-dimensional simulation of P'fl reactors was

presented at the NEACRP sponsored Specralists' Meeting in November

1979 [ 2 ] . In order to avoid true 3 D calculations, the core is

mdelled as a perturbation of the "base" reference situation calcu-

lated by the MERCATOR-XY nodal simulator [j]. MERCATOR-Z operates

on the macroscopic cross-sections, condensed in each Z plane, to

solve the diffusion equation in one dimension, in terms of XY ave-

rages of the flux at each Z level. Most of the experience with

the LWR-WIHS/blERCATOR XY/MERCATOR Z chain, has been gained on fol-

lowing the TIHANGE I reactor. Control bank insertion, boron con-

centration and axial profiles as calculated for different steady

and transient conditions have been satisfactorily compared with

experimental observation.

- The TRILUX code, initially developed by GUNF, has been improved and extended by BELGONUCLEAIRE. TRILUX calculates 3 D nodal power density

distributions using a modified one-group nodal coupling calculation 2

in which each fuel node is characterized by km and M . The LVR-WIMS' code is used as assembly constant generator. Two options have been

added to TRILUX : XENOLUX for the evaluation of xenon transients and a

MICROLUX for the determination of the pin power distribution over

selected nodes or over an "average" plane of the whole core. These

calculation tools have been compared with more sophisticated codes. . Operation data from SENA,and TIHANGE I were also recalculated for

checking the validity of the TRILUX - MICROLUX system. The interest of using the MICROLUX option has been demonstrated in the calculation

of power and burn-up maps when important gradients of the thermal flux

exist at the interface of fuel assemblies, e.g. in Pu recycle confi-

gurations. Taking into account these gradients increases the power

release in a Pu assembly by about 6 7; [ L I ] .

- The parane , t r i c s tudy of s teady st a t e cond i t i ons and acc iden t ' ana - l y s i s f o r a P m of 900 MWe loaded with 0 , 30 and 70 % plutonium

assembl ies has been completed a t BELGONUCLEAIRE wi th in t h e frame

of a c o n t r a c t wi th t h e CEC.

c ) Out-of-pile f u e l cyc l e ...................... - The m u l t i p l i c a t i o n f a c t o r of f r e s h f u e l s t o r a g e racks was c a l c u l a t e d ,

assuming t h a t t h e system is sub jec t ed t o water of a d e n s i t y vary ing

from 0 t o 100 %. S a f e t y c r i t e r i a (ANSI 18.2) impose t h a t ke f f must be lower than 0.98 f o r t h e opt imal d e n s i t y ; t h i s optimal d e n s i t y is

of t he ' o rde r of 1 0 % t o 20 % f o r u sua l s t o r a g e racks of P'dR assem-

b l i e s . A t nominal d e n s i t y , kef f may not reach 0.95.

A p r a c t i c a l arrangement is a squa re a r r a y of assembl ies , each of them

being loaded i n a s t a i n l e s s s t e e l can of 5 mm th ickness . Such a

system, with an assembly p i t c h of 35 cm, is reasonably compact f o r

f r e s h f u e l s t o r a g e (10 x 1 0 assembl ies f o r example), provides t h e

mechanical p r o t e c t i o n normally necessary b u t does not r e q u i r e s p e c i a l

absorbing m a t e r i a l l i k e boron s t e e l . Moreover, t h e presence of t h e

s t e e l cas ing reduces s i g n i f i c a n t l y t he keff peaking a t low dens i ty .

The c a l c u l a t i o n s were performed, f o r a 2 3 5 ~ enrichment of 3.5 %, w i t h

neutron t r a n s p o r t codes :

- DTF-IV wi th 40 neutron energy groups : 1 D c y l i n d r i c a l approxima- t i o n f o r a s i n g l e assembly i n an i n f i n i t e a r r a y

- DOT 3.5 wi th 6 neut ron energy groups f o r 2 D c a l c u l a t i o n s . The v a l i d a t i o n of c a l c u l a t i o n methods remains a problem a s long a s

s y s t e n a t i c a l in tercomparisons a r e not a v a i l a b l e .

- I n r e l a t i o n with t h e s a f e t y a s p e c t s of handl ing and s t o r a g e of mixed oxide f u e l assembl ies f o r PWR's, neu t ron and gamma dose r a t e s i n t h e

v i c i n i t y of Pu02-U02 f u e l assembl ies were c a l c u l a t e d by BELGONUCLEAIRE

f o r comparison with measured va lues on d i f f e r e n t types of assembl ies

manufactured i n t h e p a s t f o r B R 3 , SENA and DODEYAARD. The c a l c u l a t i o n

methods were app l i ed t o p lu ton i -~m assembl ies designed f o r a TIAANGE

type r e a c t o r . C r i t i c a l i t y c n l c u l n t i o n s were a l s o perforned i n t h e ' g!pJ-ld 16

light of the criteria ANSI 18.2, for fresh fuel assenblies in normal1.y

dry storage rlcks and for spent fuel assemblies in high density sto-

rage racks with boron containing cans around the assemblies. Mixed

oxide fuel as well as uranium oxide fuel were considered in this

evaluation [5].

2. Pressure vessel studies

The objective of these studies is to improve the neutronic aspects of LVR

pressure vessel surveillance methods and to validzte the neutron embrittle-

ment characteristics for the steel type used in the new belgian power plants.

The interlaboratory cooperation with US laboratories has been pursued in

the framework of the L1tfR Pressure Vessel Irradiation Surveillance Dosimetry

progranme supported by the NRC [ 6 ] .

Host of the efforts have been devoted to the analysis of the experimental

results obtained at the O R N L Pool Critical Assembly (PCA) and to the

characterization of the Pool Side Facility (PSF) at the O R R where the

irradiation of steel specimens has been started.

- Two configurations of the PCA Pressure Vessel mock-up were studied, cor- responding to two positions of the thermal shield and pressure vessel ,

simulators with respect to the PCA reactor core. A detailed map of the

fission density in the core itself was first performed to provide an

exact picture of the fast neutron source distribution on an absolute

basis.

Threshold reaction rates were measured with fission chambers ( 2 B U 1 237Np)

and activation detectors ('031?h, 'I5.In, 58Ni, 27~1) at different neutron 6

penetration depths in water and in the simulators. ~i(n,a) neutron

spectrometry measurenents were finally made at three locations within

,the pressure vessel steel.

The accuracy of the measurements is better than - + 5 % for the spectral indices and better than - + 7 % for the absolute equivalent fission fluxes per unit PCA core neutron strength. A transport theory analysis of both

configurations was carried out : one- and two-dimensional multigroup

S8 - P3 cnlculations were performed including coupled neutron-gamma

calculations to correct neutron dosimeters for gamma ray induced res-

ponses.

The calculations reproduce the integral measurement results to within

an uncertainty of 2 25 :A. However it remains to ascertain the signi-

ficance of errors associated to the treatment of vertical neutron

leakage effects in the 2 D calculations, resulting from the more

limited height of the PCA core as compared with power reactors. This

calla in particular for pursuing the leakage sensitivity study under-

taken at SCK/CEM and included in the report presented at the PCA 6

"Blind Test" meeting [7]. The consistency of the Li(n,a) neutron

spectra -[a] with the spectral indices also requires further investi- gation although a general agreement of + 10 % has been reached when considering only the reaction rates mostly sensitive in the neutron

energy range covered by the spectrometry technique.

The Blind Test exercise was organized in the frame of the validation

effort of transport theory computations needed to extrapolate, into

the pressure vessel of a power reactor, the dosimetry results obtained

at a surveillance position. The initial comparison of the solutions

proposed by the participants took place in May, 1980, at the NBS [g].

- The dosimetry radiometric measurements performed at PCA have Seen repeated in the actual PSF configuration at low power (10 - 20 \Id) and at high power (30 MW); the equivalent fission fluxes for 03~h,

lq51n, 58~i and 27~1, reported to a unit core power defined by fission

chanber neasurements at PCA and by thermal balance at ORR, have been

found identical within - + 3 7L or better.

As part of this PSF "start up" characterization programme, various

damage exposure parameters, such as jb > 1 MeV, fl > 0.1 MeV and dpa were derived from the radiometric results coupled with transport

theory calculations. It is concluded that in-vessel projections of

surveillance capsule enbrittlement data based on ,6 > 1 MeV as the damage correlation parameter may be non-conservative by up to 15 %

if dpa proves adequate and by up to 40 if @ > 0.1 MeV proves adequate,

a s s u g g e s t e d by some e x p e r i m e n t s . The s t e e l spec imen i r r a d i i t i o n s i n

PSF, a t t h e s u r v e i l l a n c e p o s i t i o n and i n t h e p r e s s u r e v e s s e l s i m u l a t o r ,

w i l l h e l p t o c l a r i f y t h i s i m p o r t a n t problem.

- Dosimetry measurements a r e under p r o g r e s s i n BR3 f o r comparison w i t h c a l c u l a t i o n s i n t h e r e a l c o n d i t i o n s of a p p l i c a t i o n of t h e t h e o r e t i c a l

methods. A f i r s t s e r i e s o f r e s u l t s were o b t a i n e d i n B R 3 a t v a r i o u s

l o c a t i o n s be tween t h e p e r i p h e r y o f t h e c o r e and t h e o u t e r s i d e o f

t h e p r e s s u r e v e s s e l . On t h e o t h e r h a n d , s e v e r a l s u r v e i l l a n c e c a p s u l e s

were un loaded from t h e TIHANGE and DOEL r e a c t o r s . The r a d i a l and

a z i m u t h a l g r a d i e n t s o f t h e f a s t n e u t r o n f l u x i n t h e c a p s u l e s have been

de te rmined a c c u r a t e l y by measur ing t h e 54Mn a c t i v i t y i n t h e remnants

a v a i l a b l e a f t e r t h e mechan ica l t e s t s . A s a l a r g e number o f niobium

d o s i m e t e r s had been l o a d e d i n one of t h e s e c a p s u l e s , i t h a s been pos-

s i b l e t o d e f i n e a n e f f e c t i v e a c t i v a t i o n c r o s s - s e c t i o n f o r t h e r e a c t i o n

9 3 ~ b ( n , n ' ) 9 3 m ~ b on t h e b a s i s of t h e measured f l u e n c e by means of t h e

more c o n v e n t i o n a l d o s i m e t e r s . Fo r l o n g i r r a d i a t i o n t i m e s , n iobium,

w i t h i t s h a l f - l i f e of 16.4 y e a r s , w i l l b e t h e most v a l u a b l e f l u e n c e

mon i to r .

3. R e a c t o r o p e r a t i o n

The BR2 r e a c t o r h a s been r e s t a r t e d a t t h e b e g i n n i n g o f J u l y , 1980 , a f t e r

r ep lacemen t o f t h e b e r y l l i u m m a t r i x . The r e a c t o r was shutdown s i n c e 1 8

months. The o r i g i n a l m a t r i x had been s u b m i t t e d t o a n e u t r o n f l u e n c e of

a b o u t 8 x 1 0 ~ ~ n / c m ~ (> 1 MeV) i n t h e h i g h e s t r a t e d p i e c e s . S y s t e m a t i c a l

measurements were c a r r i e d o u t b e f o r e d i s m a n t l i n g i n o r d e r t o c o r r e l a t e the

d i m e n s i o n a l c h a n e e s a t d i f f e r e n t p o s i t i o n s w i t h t h e l o c a l f l u e n c e v a l u e s ,

The d i a m e t e r of t h e c h a n n e l s , which i s n o m i n a l l y 8 4 m m , i n c r e a s e d w i t h t h e 22 2 f l u e n c e t o r e a c h 1.1 m m a t 8 x 1 0 n/cm . By r e a s o n of t h e t r e n d o f t h e

c u r v e , i t seems t h a t t h e c r i t i c a l t e m p e r a t u r e f o r b e r y l l i u m equa led t h e 2 2 2

p h y s i c a l t e r n p e r a t w e (" 50' C) a t a f l u e n c e o f a b o u t 6.5 x 10 n/cm . Above t h i s v a l u e , t h e s w e l l i n g r a t e i n c r e a s e d more r a p i d l y (he l ium bubb le

f o r n a t i o n on rain b o u n d a r i e s ) ; a n i n c r e a s e o f t h c t r i t i u m c o n t e n t i n w a t e r

was obse rved i n p a r a l l e l . Helium c o n c e n t r a t i o n measurements i n s m a l l

samples of the unloaded pieces will allow to confirm the fast neutron

dose distribution. Fresh beryllium samples were previously irradiated

together with neutron dosimeters for calibration purpose.

The new matrix was loaded in January 1980 [lo] and the reactor was made

critical again at low power on May 12, 1980.

FAST REACTORS - 1. Critical experiments

The analysis of the experiments carried out in ZEBRA at Winfrith in the

frame of th-e BIZET programme under AEEW/KfK agreement was pursued.

BELGONUCLEAIRE, as DeBeNe partner, took part to this work :

- the correction factors for the modified source multiplication method applied to the conventional 2-zone core loading BZA were recalculated

using SNR methods and data

- the calculations of the control rod experiments in BZA were reanalysed in order to compare the UK and DeBeNe results; nine symmetrical B4C

control rod configurations were considered. On the DeBeNe side, diffu-

sion theory is used with cross-sections condensed from the

26-group KFK/INR 001 set; the calculations were made in two and three

dimensions

- the gamma-ray energy deposition measurements in the heterogeneous loading BZC/I [II] were also calculated with the SNR design methods and photon

libraries. Differences in the fissile zone are attributed to the photon I( source library and in the fertile zones to the diffusion/transport

effects

- the relative reactivity of pins and plate cells was calculated using the FGL5/MURAL cross-sections and the pin sector substitution measurements in

BZD/3 were extrapolated to a fuel pin core

- the effect of insertion of a Na/steel channel in the central fertile island of BZD/3 was evaluated in function of the channel diameter.

On the experimental side, an intercomparison of gamma dose measurements

by means of thermoluminescent detectors was undertaken. The reference 60

gamma fields ( Co source) at Harwell and Mol, which were used for cali-

bration of the TLD's in the BIZET programme, were first intercompared by

means of the "MOL" ionization chamber exposed in both facilities; the

absolute value measured at Harwell agrees with the "HARWELL" ionization

chamber result within the 3 to 4 % systematical plus statistical uncer-

tainty. Moreover, TLD's from a batch calibrated at Mol were exposed at

Harlwell and a similar agreement was obtained. An improved proce6ure of

selection and calibration of TLD's is now 'being developed at Mol in order

to achieve a better control of the uncertainty on individual measurement

results in a critical facility.

2. Safety studies

The feasibility study of PAHR (Post Accide:nt Heat Removal) irradiation

experiments in BR2 is investigated in order to evaluate the heat removal

capabilities in the case of a core meltdown in a fast reactor, when the

core debris are collected in the core catcher. The irradiation device

would be located in the 20 cm diameter central channel of the reactor;

the fufl particle bed diameter simulating the core debris would amount

8 to 10 cm.

The neutronic calculations were carried out with the SCK/CEN version of

the neutron transport code DTF-IV and the BR2 fourty energy group library.

All calculations were made in R-geometry. The gamma calculations were

also performed with the DTF-IV code, together with the EURLIB ( Y , Y )

library with twenty energy groups. Preliminary two-dimensional (R,Z)

ganna heating calculations wzre performed with the aid of the DOT 3.5 code.

The possibility of increasing the diameter of the central channel up to

40 cm was also investigated with a view to accepting larger PAHR experi-

ments, with a fuel particle bed diameter of about 20 cm. Neutronic cal-

culations indicate that such n major modification is acceptable from the

reactor operation point of view and that irradiations of standard devices

as those used in the past cnn be carried on.

I n each c a s e , a r e l a t i v e l y f l a t r a d i a l f i s s i o n d e n s i t y d i s t r i b u t i o n could

be ob ta ined , t h e gamma h e a t i n g being of minor importance a s t h e r e a c t o r

would ope ra t e a t a reduced power l e v e l .

I n p a r a l l e l was s t a r t e d a d e t a i l e d modelling of t h e phenomena occur r ing i n

a d e b r i s bed made of oxide f u e l and s t a i n l e s s s t e e l p a r t i c l e s s a t u r a t e d

with l i q u i d sodium, wi th i n t e r n a l hea t genera t ion . One- and two-dimensional

approaches of t h e problem a r e be ing developed t o e v a l u a t e t h e u se fu lnes s

of PAHR experiments i n BR2 [12].

3. Neutron dosimetry f o r f u e l and m a t e r i a l i r r a d i a t i o n s

The c o l l a b o r a t i o n with t h e Max Planck I n s t i t u t e (MPI) i n S t u t t g a r t was

continued with a view t o t h e u t i l i z a t i o n of niobium a s monitor f o r long-

term i r r a d i a t i o n s . Very pure niobium m a t e r i a l , f a b r i c a t e d at t h i s

I n s t i t u t e , was i r r a d i a t e d t o g e t h e r with commercially a v a i l a b l e niobium

i n BR2. During t h e f i r s t weeks o r months a f t e r t h e i r r a d i a t i o n , t he

9 3 m ~ b a c t i v i t y ( h a l f - l i f e 16.4 ears) i s d i s t u r b e d by t h e 1 8 3 ~ a , 9 5 m ~ b ,

1 8 2 ~ a and 9 5 ~ b a c t i v i t i e s . A f t e r h a l f a y e a r , the. 93"~b a c t i v i t y is only

in f luenced by t h e IG2Ta a c t i v i t y ( h a l f - l i f e 115 days) . Th i s i n f luence

is p r a c t i c a l l y n e g l i e i b l e i n t h e MPI niobium whereas i t can c o n t r i b u t e . t o 20 ... 100 9: of t h e measured a c t i v i t y i n commercial niobium, depending on t h e f o i l t h i c k n e s s i n t h e range 1 0 t o 100 pm.

22 -2 Niobium f o i l s i r r a d i a r e d i n EBR-I1 a t a t o t a l f l uence of 1 0 n.cm and

2 0 -2 i n BR2 a t a f i s s i o n equ iva l en t f luence of 5 x 10 n.cm were measured by

s i x l a b o r a t o r i e s . The measured 93m~b a c t i v i t i e s ag ree wi th in a few

pe rcen t s when t h e same d a t a a r e used f o r t h e h a l f - l i f e and t h e K(X)-ray 7

emission p r o b a b i l i t y .

The niobium measurements f o r EBR-I1 were p a r t o f m u l t i f o i l dosimetry

measurements ( T i , Fe, Co, N i , Cu, Nb, 235iJ 9 2381J, 2 3 7 ~ p ) a t d i f f e r e n t

a x i a l l e v e l s i n a c e n t r a l core channel. These measurements have been

completed; t h e s p e c t r a l unfo ld ing is performed a t HEDL.

REFERENCES

[ I ] L.W. WESTON and J . H . TODD (ORNL)

Neutron Cap tu re and F i s s i o n Cross S e c t i o n s of ~ l u t o n i u m - 2 4 1

Nucl. Sc. and Eng., 65, 4.54-463, 1978

[2] M. MELICE ( E l e c t r o b e l )

HERCATOR-2, a P e r t u r b a t i o n Approach t o PWR Core S i m u l a t i o n

P r o c e e d i n g s of a S p e c i a l i s t s ' Meeting on " C a l c u l a t i o n of 3-Dimensional

R a t i n g D i s t r i b u t i o n s i n O p e r a t i n g R e a c t o r s " ,

P a r i s , November 26-28, 1979

[3] M. HELICE ( E l e c t r o b e l )

A Nodal-Modal Coarse-Mesh Method f o r S o l v i n g t h e Two-Group D i f f u s i o n

E q u a t i o n

Repor t NEACRP-L-228, November 1978

[4] A. CIIARLIER, A. MOCKEI, and J.P. TESCH ( B e l g o n u c 1 6 a i r e )

TRILUX Nodal Sys tem f o r In-Core F u e l Manngement S t u d i e s

P r o c e e d i n g s of a S p e c i a l i s t s ' Meeting on " C a l c u l a t i o n of 3-Dimensional

R a t i n g D i s t r i b u t i o n s i n O p e r a t i n g R e a c t o r s 1 ' ,

P a r i s , November 26-28, 1979

[5] C. VANDENBERG ( B e l g o n u c l b a i r e )

Gamma and Neu t ron Dose R a t e s i n t h e Handl ing and t h e S t o r a g e of

P lu tonium F u e l A s s e m b l i e s

X i s $ , Oc tobe r 2 5 , 1979

[6] W.N. Mc ELROY (HEDL) e t a l .

LWR P r e s s u r e V e s s e l S u r v e i l l a n c e Dosimetry Improvement Program.

1979 Annual R e p o r t

NUREG/CR-1291, HEDL-SA 1949

[ 7 ] G. MINSART (SCK/CEN) ...

N e u t r o n i c Computa t ions o f t h e Poo l C r i t i c a l Assemt,ly P r e s s u r e V e s s e l

F a c i l i t y (PCA-PVF)

"PCA B1ip.d T e s t 1 ' Mee t ing , Washington, Nay 22-23 , 1 9 8 0

?2 , , , .,

[8] G. DE LEEUW-GIERTS, D. LANGELA (SCK/CEN) 6 Li Spectrometry Results in PCA

ItPCA Blind Test" Meeting, Washington, May 22-23, 1980

- [9] C.Z. SERPAN (NRC), M.N. Mc ELROY (HEDL), F.B.K. KAM (ORNL) and

A. FABRY (SCK/CEN)

Minutes of the May 22-23, 1980, Computational Blind Test on the

Pool Critical Assembly Pressure Vessel Mock-up Facility

'[lo] F. LEONARD, A. FALLA (SCK/CEN) Remplacement de la Matrice de Beryllium du R6acteur BR2

Irradiation Devices Working Group (EURATOM), 26th Meeting,

Geesthacht, October 8-10, 1980

I ] A.D. KNIPE (AEEkl) , R. de WOUTERS (Belgonucl6aire) Gamma-Ray Energy Deposition Measurements in a Heterogeneous Core

and their Analysis

International Symposium on Fast Reactor Physics (IAEA-SM 244).

Aix-en-Provence, September 24-28, 1979

1:12] C. BENOCCI, J.M. BUCHLIN (von KARMAN Institute, ~hode-Ste-Gensse,

Belgium), C. JOLY, A. SIEBERTZ (SCKICEN)

Pre-Boiling State in - Post - Accident - Heat gemoval Situation : 1 D and 2 D Theoretical Approaches Including

Natural Convection Effects

International Seminar on Nuclear Reactor Safety Heat Transfer,

. Dubrovnik, September 1980. ,

NEACRP-L-244 CANADA

REACTOR PHYSICS ACTIVITIES I N CANADA -

M.F. D u r e t

POWER REACTOR PROGRAM

The Bruce A and P i c k e r i n g c o n t i n u e t o o p e r a t e a t near maximum power. A g r e a t dea l o f i n t e r e s t has r e c e n t l y been expressed by O n t a r i o i n d u s t r i e s i n u s i n g steam produced i n t h e Bruce g e n e r a t i n g s t a t i o n . T h i s o b j e c t i v e i s b e i n g g i v e n h i g h p r i o r i t y by t h e government.

The r e a c t o r c o n s t r u c t i o n program i s on schedule w i t h abou t 15 Gw( e ) expected t o be i n s e r v i c e by 1990.

EXPERIMENTAL REACTOR PHYSICS

Work i n t h e ZED-I1 r e a c t o r d u r i n g t h e p a s t y e a r has g e n e r a l l y been s e r v i c e o r commercia l work t o s a t i s f y requ i rements o f o t h e r groups. T h i s i n c l u d e s work t o measure f l u x d i s t r i b u t i o n s and r e a c t i v i t y o f 36-e lement bund les o f PuD2/U0 f u e l w i t h a i r and D20 c o o l a n t s p r i o r to i r r a d i a t i o n t e s t i n g i n an NRU ? oop and f l u x measurements w i ' i h i n new des igns o f s e l f - powered f l u x d e t e c t o r s . The n e x t 1 a rge measurement program w i 11 i n v o l v e PuO /Th02.which i s p r e s e n t l y b e i n g f a b r i c a t e d a t CRNL and i s expected 6 t o e a v a i l a b l e e a r l y i n 1981.

D u r i n g t h e p a s t y e a r CRNL p a r t i c i p a t e d i n i in IAEA benchmark a c t i v i t y to i n t e r c o m p a r e measurements on s t a n d a r d gama-sources u s i n g Ge-Li d e t e c t o r s .

ANALYTICAL REACTOR PHYSICS

An approx imate method f o r s o l v i n g t h e B o l t m a n n e q u a t i o n i n t h e v i c i n i t y o f a p l a n e boundary between m a t e r i a l s w i t h d i f f e r e n t p r o p e r t i e s has been developed. A paper on t h i s t o p i c has been s u b m i t t e d t o t h e IAEAIENS s p e c i a l i s t s m e e t i n g i n A p r i l 1981.

The computer program most o f t e n used i n r e a c t o r p h y s i c s c a l c u l a t i o n s a t CRNL has been t h e c e l l code LATREP. T h i s program has been s u b s t a n t i a l l y r e v i s e d r e c e n t l y t o i n c l u d e a n u c l e a r da ta l i b r a r y , i n t h e W e s t c o t t fo rmal ism, based on ENDFIB. I n t h e process o f t e s t i n g t h i rogram, d e f i c i e n c i e s i n t h e ENDFIB c r o s s s e c t i o n s f o r L L I ~ ~ ~ and InT1! have been d i scovered .

ADVANCED FUEL CYCLES AND ASSESSMENT

The p o s s i b i l i t y o f u s i n g a once- through t h o r i u m c y c l e has been i n v e s t i g a t e d . T h i s concep t i n v o l v e s u s i n g s l i g h t l y e n r i c h e d uran ium " d r i v e r " f u e l i n c o n j u n c t i o n w i t h pu re t h o r i u m f u e l bund les . From t h e p o i n t o f v iew o f n e u t r o n ba lance and economics t h e c y c l e appears f e a s i b l e w i t h no c r e d i t t aken f o r t h e U-233 i n t h e spent f u e l . However, p r a c t i c a l i m p l i m e n t a t i o n r e q u i r e s f u r t h e r work on f u e l mariagement s t r a t e g i e s t o de te rm ine whether h e a t t r a n s f e r r e q u i r e m e n t s car1 be s a t i s f i e d .

Several concepts for advanced fuel cycles i n CANDU reactors proposed by A. Radkowski have been investigated and abandoned.

A study of the advantages of using s l i gh t ly enriched uranium in CANDU reactors has been completed. The optimum enrichment appears to be about 0.93%. A t t h i s enrichment, b u r n u p i s increased to about 15 MW.d/kg which reduces fuel consumption by about 25% and requires no change i n the reactor

* design. A t higher enrichments, fur ther improvements i n burnup and fuel consumption are possible b u t flux d i s tor t ions and hot spots begin to appear in the core and t h i s would probably require changes i n both the core and fuel design.

The concept of "spal la t ion breeding'' is being pursued a t CRNL and reactor physics calculations have been made for several accelerator-target- blanket combinations.

a A study of the s u i t a b i l i t y of several advanced fuel cycles for the Canadian s i tuat ion has been made. A moderate to ta l energy growth of 2.7%/ year was assumed together w i t h two growth ra tes of ins ta l led nuclear capacity i n the f i r s t 50 years of the next century. Thus annual growth ra tes are 3% and 4% between 2000 and 2050. I t i s expected tha t reactors operating on advanced cycles will not be introduced i n s ign i f ican t numbers before the year 2000. Over the period considered the P u / t h cycle conserves the most uranium, largely because i t uses the plutonium i n spent natural uranium fuel.

SMALL REACTOR DEVELOPMENT

A 2 MW reactor for low temperature heating ( 100•‹C) i s in the conceptual design stage. The core cons is t s of 200 CANDU-type fuel elements containing 5% enriched uranium oxide. Reactivity i s controlled by a motor- driven beryllium annulus surrounding the core. There are no other mechanical control devices.

Economic f e a s i b i l i t y a t such a low power level depends on achieving a high degree of inherent safety a t low cost , and eliminating the need for fu l l time sk i l led operators. These a t t r i b u t e s have already been demonstrated by the 20 kW research reactors over the past ten years. The 2 MW concept has similar inherent safety charac te r i s t ics , based on limited reac t iv i ty additions and a 1 arge negative void coeff ic ient . Theno- hydraulic experiments to study and demonstrate the inherent safety pr inciples are now i n progress. .

Ris@ Nat iona l Laboratory

Department of Reactor Technology

NEACRP-L-244 DENMARK

September 1980

Recent Reactor Phys ics A c t i v i t i e s i n Denmark

by

Hans Nel t rup

1. Fue l Box C a l c u l a t i o n wi th Response Ma t r i ce s

A program REPRO c a l c u l a t i n g t h e response ma t r ix f o r he te ro-

geneous square p i n c e l l s has been developed. The v a r i a b l e s

of t h e problem a r e t h e expansion components of t h e angula r

f l u x on t h e c e l l s u r f a c e .

The expansion, complete and independent f o r each h a l f space,

i s one of s e v e r a l p o s s i b l e f o r angula r f l u x e s which a r e sym-

me t r i c w i th r e s p e c t t o an xy-plane perpendicu la r t o t h e p lane

d e f i n i n g t h e two h a l f spaces .

The elements of t h e response m a t r i x r e p r e s e n t t h e coupl ing i n

s e v e r a l energy groups between i n - and outgoing components i n

a set o f p o i n t s on t h e c e l l s i d e s e.g. p o i n t s f o r g a u s s i a n

i n t e g r a t i o n a long each s i d e .

. C a l c u l a t i o n o f : t h e response m a t r i x i s performed w i t h c o l l i s i o n p r o b a b i l i t i e s i n s i d e t h e ce l l which is rep re sen ted by a g r e a t

number o f subregions . The f l u x response i n t h e s e i s r e g i s t e r e d

a s e lements i n a r eg ion f l u x response mat r ix .

A four - fo ld symmetry reduces cons ide rab ly t h e necessary com-

a p u t a t i o n s and t h e number o f m a t r i x e lements t o be s t o r e d . A second program FLUSO s o l v e s t h e e igen v a l u e equa t ions o f t h e

in - and outgoing components o f a r e c t a n g u l a r a r r a y of p i n c e l l s

w i t h s u i t a b l e (b lack or t r u l y r e f l e c t i n g ) boundar ies . . -- .- ..

The i n t e r n a l r eg ion f l u x e s i n a l l cells are found by o p e r a t i n g

t h e r eg ion f l u x response m a t r i c e s on t h e component e igen

v e c t o r .

C a l c u l a t i o n s w i t h two energy groups , two angu la r componenks

and up t o 10 g a u s s i a n p o i n t s p e r ce l l s i d e and 25 i n t e r n a l

r eg ions - 24 i n four - fo ld symmetry and one c e n t r a l c i r c u l a r r eg ion - p e r c a l l have been performed w i t h r a p i d convergency i n t h e FLUS0 r o u t i n e and y i e l d i n g reasonable f l u x d i s t r i b u , t i o n s . Unfor tuna te ly it is d i f f i c u l t t o f i n d comparable

measurements o r c a l c u l a t i o n s . However, t h e f l a t f l u x and t h e

- e x a c t c a l c u l a b l e keff and f a s t t o t h e thermal f l u x r a t i o i n a t o t a l l y r e f l e c t e d homogenous s q u a r e c e l l is e x e l l e n t l y re-

produced, a l though t h i s ce l l i s n o t s p e c i a l l y s u i t e d f o r t h i s

c a l c u l a t i o n method. The f a s t f l u x is f l a t w i t h i n 0.5% from t h e

mean f l u x and t h e thermal w i t h i n 0.1%. The f a s t t o thermal -. .. i..:

f l u x r a t i o is c o r r e c t w i t h i n 0.08% and t h e ensu r ing keff w i t h i n 7.2 CS3

0.1%. (3 c-4

~~ . i::.:r

2. Core Performance Eva lua t ion (core-s imula tor ) e ..f e\

Opera t iona l r e s t r i c t i o n s a r e imposed on l i g h t wate r r e a c t o r s

i n o r d e r t o avoid f u e l f a i l u r e s , o r a t l e a s t t o d imin i sh t h e

number of f a i l u r e s . A s such r e s t r i c t i o n s n e c e s s a r i l y r e s u l t i n

reduced power p roduc t ion from t h e r e a c t o r s they a r e u n d e s i r a b l e

from an economical p o i n t of view. Knowledge of t h e l o c a l power

ramps and t h e i r consequelices f o r t h e f u e l i s r e q u i r e d i n o r d e r

t o reduce t h e l e v e l of r e s t r i c t i o n s c o n s i s t e n t w i th s a f e t y

requi rement . : . . ., - . . .~

A comprehensive system f o r t h e c a l c u l a t i o n of t h e f a i l u r e proba-

b i l i t y f o r t h e i n d i v i d u a l f u e l r o d s throughout t h e r e a c t o r c o r e

i n l i g h t water r e a c t o r s i s be ing developed. The c a l c u l a t i o n a l

system i s s e t up a s a modular system. The modules t o be - inc luded

a r e : 3D-nodal n e u t r o n i c / h y d r a u l i c module f o r t h e c a l c u l a t i o n of

t h e 3D power d i s t r i b u t i o n (ANTI o r NOTAM), f u e l box module f o r

t h e c a l c u l a t i o n of homogenized c r o s s - s e o t i o n s f o r t h e i n d i v i d u a l

f u e l boxes (CDB), and f u e l r e l i a b i l i t y module f o r t h e c a l c u l a -

t i o n of t h e f a i l u r e p r o b a b i l i t y f o r t h e i n d i v i d u a l f u e l r o d s

(FRP). For t h e f u e l f a i l u r e c a l c u l a t i o n s , t h e power h i s t o r y f o r

each i n d i v i d u a l f u e l p i n i s r e q u i r e d ; a module f o r t h e c a l c u l a - t i o n s of t h e s e h i s t o r i e s a r e l i k e w i s e t o be inc luded .

A s p a r t of t h e Ph.D. d i s s e r t a t i o n a number of methods of ca l cu -

l a t i n g t h e l o c a l p i n power i n a BWR has been i n v e s t i g a t e d .

1. A s a s imple approximat ion , t h e loca l . f l u x d i s t r i b u t i o n

found i n t h e f i r s t s t e p is renormali .zed; i n t h i s way t h e

assembly average power a g r e e s wi th the one ob ta ined from

t h e g l o b a l coarse-mesh s o l u t i o n .

2 . A more s o p h i s t i c a t e d method i s based on t h e modulat ion

model where t h e heterogeneous s o l u t i o n from t h e f i r s t s t e p

is m u l t i p l i e d w i t h a smooth f l u x - d i s t r i b u t i o n making use

of t h e boundary c o n d i t i o n s ob ta ined from t h e coarse-mesh

s o l u t i o n .

3 . A s u p e r p o s i t i o n of t h e two s o l u t i o n s , a s supposed t o t h e

modulat ion model, makes i s p o s s i b l e t o p r e s e r v e both t h e

average power and t h e e igenva lue from t h e g l o b a l coa r se -

mesh s o l u t i o n . T h i s method seems t o g i v e b e t t e r r e s u l t s

than t h a t u s ing t h e modulat ion model when t h e d i f f e r e n c e

between t h e two s o l u t i o n s can be regarded a s a sma l l p e r t u r -

b a t i o n of one of them.

Both t h e modulation and t he superposi t ion model a r e very sen-

s i t i v e with regard t o s t rong heterogeni t ies . To avoid some.of

t h e d i f f i c u l t i e s when deal ing with very heterogeneous regions,

a procedure based on t h e response matrix method has been

examined.

. 4 . When ca l cu l a t i ng t h e smooth f lux-d i s t f ibu t ion i n s ide t h e

f u e l boxes of a BWR t h e average fluxed and cu r r en t s a t t h e

boundaries a r e t r ans fe r red through t h e watergaps and im-

pressed d i r e c t l y on t h e homogenized f u e l region. In t h i s

way t h e inaccuracy involved when homogenizing very

a heterogeneous regions i s reduced. A comparison i s made of t h e above mentioned four s t r a t e g i e s

f o r combining a heterogeneous box-solution and t he r e s u l t s from

the ove ra l l coarse-mesh solu t ion .

- . --. - - -- - - - - -

The inves t iga t ion shows t h a t t he b e s t way t o combine t h e t w o so lu t ions seems t o occur when t h e heterogeneous so lu t ion from

the box ca l cu l a t i on (with r e f l e c t i n g boundaries) is superposed

with a smooth f lux-d i s t r ibu t ion i n t h e homogenized f u e l region

of t h e f u e l box.

The t h r ee ca lc i i la t ion s t e p s then follow:

0 - A t f i r s t two sets of average c ro s s sec t ions are ca lcu la ted ,

one v a l i d f o r t h e whole f u e l box and t he o the r f o r t h e f u e l

region alone.

- I n a second s t e p t h e o v e r a l l f lux-d i s t r ibu t ion is found from a coarse-mesh ca lcu la t ion .

- I n a t h i r d s t e p t he boundary condi t ions found i n t h e second s t e p a r e t r ans fe r red through watergaps and con t ro l rods and

impressed d i r e c t l y on t h e f u e l region. A smooth f l ux -d i s t r i -

but ion i n t h e f u e l region of t h e box i s then ca lcu la ted and

t he so lu t ion obtained is superposed with t h e heterogeneous

so lu t ion from t h e f i r s t s t ep .

, BY use of t h i s method it i s assumed t h a t t h e l o c a l power ramps

t o be used l a t e r i n f u e l performance s t u d i e s can 5 e e s t ima ted

t o an accep tab le accuracy.

!

3 . Core s u r v e i l l a n c e

A programme f o r on-l ink s imu la t ion of t h e t h r e e dimensional

power distribution i n l i g h t wate r r e a c t o r s i s under c o n s t r u c t i o n .

The programme u s e s d e t e c t o r s i g n a l s an6 nuc lea r and o p e r a t i n g

d a t a a s i n p u t . I t i s assumed t h a t t h e d e t e c t o r s i g n a l is pro-

p o r t i o n a l t o t h e average power d e n s i t y i n t h e f o u r f u e l e l e -

ments surrounding t h e d e t e c t o r . Pseudo-s ignals a t each d e t e c t o r

l e v e l a r e c a l c u l a t e d by so lv ing a two-dimensional nodal equa t ion

i n which t h e f i s s i o n source o f each ins t rumented c e l l is norma-

l i z e d s o t h a t c a l c u l a t e d and measured power d e n s i t y agree .

Havlng determined d e t e c t o r s i g n a l s , pseudo o r r e a l , f o r each

c e l l i n t h e r e a c t o r , t h e a x i a l power d i s t r i b u t i o n of a l l c e l l s

i s c a l c u l a t e d by a d j u s t i n g r a d i a l i n t e r a c t i o n . F i n a l l y i n d i v i d u a l

segment powers a r e determined us ing a v o i d , exposure and c o n t r o l

rod dependent mismatch f a c t o r . I n doing s o a s imple thermo-

hydrau l i c model is a p p l i e d .

4 . ANTI

For t h e c a l c u l a t i o n of t r a n s i e n t s i n a PWR c o r e , t h e t h ree -

dimensional computer program ANTI wi th coupled n e u t r o n i c s and

thermal -hydrau l ics is under development, The program combines

t h e n e u t r o n i c s p a r t of t h e BWR program ANDYCAP wi th t h e sub- - channel h y d r a u l i c s program TINA. It is in tended f o r t r a n s i e n t s - where t h e s p a t i a l d i s t r i b u t i o n of power and c o o l a n t f low i n

t h e c o r e i s important , p a r t i c u l a r l y c a s e s where a l o c a l power

i n c r e a s e occurs . The s t eady s t a t e p a r t of t h e program is used .... .- . ,..~ i n connect ion wi th t h e Core-Sinulator work f o r c a l c u l a t i o n of . ,. . . ~, , _.I

t h e o v e r a l l power d i s t r i b u t i o n . ...- * .-I . : ,.: .. c .. , . .,

The program i s now i n t h e running- in phase where t e s t i n g i s . ,

going on i n p a r a l l e l wi th mod i f i ca t ions and improvements. A 6:";

testcase, simulating a control rod ejection from a small reac-

tor core, has been calculated and is reported in Ris@-M-2209.

The report also contains a brief program description.

While such initial test calculations have demonstrated that

the ANTI program is able to carry out transient calculations

they are not very useful with regard to the verification of

the results. Therefore, calculations are needed either for

more realistic cases with comparison to measured data or, at

least, cases which can be compared to the results of other

computer programs.

A test case (also control rod ejection) which has previously

been calculated by the ANDYCAP program has been repeated by

ANTI. Rather large differences were found between the ANDYCAP

and the ANTI results, and the main reason seems to be the

different fuel rod models. The results indicate that it is

important to describe the heat conduction in the fuel rod

cladding, which is done in ANTI. In the ANDYCAP fuel rod model

the cladding is described as a simple resistance to the heat

transfer from fuel to coolant.

A more realistic study of the Westinghouse 3000 M W t reactor has

been initiated. So far only static calculations have been per-

formed using data from safety analysis reports. For verification

of ANTI, power shapes have been calculated with the finite

difference program TWODIM and successfully compared to Westing-

house results. ANTI is a nodal programme involving internodal

coupling parameters with a significant influence on the results.

However, for a given nodal configuration it is an easy task to find a set of parameters which results in an acceptable solution

for very different power shapes. For each nodal configuration

a new set of parameters should be found, otherwise serious errors

may be introduced. The study has been carried out with two

different nodal configurations of either one or four almost

cubic nodes per horizontal layer of a fuel assembly.

5. A Model for a Westinghouse PWR-P0we.r Plant

The old model PWR/PLASIM from 1975 has been revised and im-

proved for calculation of more severe transients. The model is

now developed in a similar way as BWR/PLASIM for the Barseback

plant and based upon data from the Westinghouse RESAR 31.

The main features are as follows:

The reactor is described in one dimension with 14 core nodes

for neutron kinetic and hydraulic calculation. The diffusion

equation is used with one energy group and prompt-jump approxi-

mation. Neutron cross sections are taken as functions of cool-

ant density, fuel temperature, control rod density and boron

concentration. Three groups of delayed neutrons are used and

six source groups for delayed heat release. The calculation of

delayed heat is done in a global manner disregarding the local

variation. The two cooling channels are used: a mean power

channel with calculation of coolant temperature and a hot

channel with calculation of both coolant temperature and void.

A fixed hot channel factor is used. The void in the mean power

channel is found from the hot channel void using a fixed

weighing factor.

The primary circuit has only one loop with steam generator,

pump and pressuriser. The propagation of temperature variation

is simulated with pure time delays for the tubes and pure

mixing in reactor and steam generator volumes and in the pump.

The heat transfer section in the steam generator is divided

into 3 nodes for the secondary side and 6 for the primary side.

The steam load circuit is not included in RESAR 31 so the model

for the turbine and feedwater heaters is only provisional with

one HP and one LP section for both turbine and feedwater heaters.

The description of the control circuit:s in RESAR 31 does not

give sufficient information for simulation, so only a very simple

control algorithm. with estimate? parameters have been used to

close the main control loops. -, ! .: ..; :... : -.~, .. , .* : , ; . . ' , ,., .,; . '.ye ,. .

- I Before t h e model can be used f o r c a l c u l a t i o n o f t r a n s i e n t s

w i th a reasonable p r e c i s i o n a l o t o f d a t a must be provided,

n o t on ly f o r t h e steam and c o n t r o l c i r c u i t s , b u t a l s o f o r t h e

primary c i r c u i t and t h e r e a c t o r .

NKACRP-L-244. FINLAND

STATUS REPORT FOR TNE NEACRP 1 9 8 0

REACTOR PHYSICS ACTIVITIES I N FINLAND

R e s e a r c h a n d d e v e l o p m e n t i n a p p l i e d r e a c t o r p h y s i c s are

c o n c e n t r a t e d a t t h e T e c h n i c a l R e s e a r c h C e n t r e o f F i n l a n d

w h e r e most o f t h e s t u d i e s c a r r i e d o u t h a v e b e e n c o n n e c t e d

b o t h w i t h core f o l l o w a n d f u e l management c a l c u l a t i o n s and

w i t h t r a n s i e n t a n a l y s e s f o r L o v i i s a (PWR) a n d O l k i l u o t o

( B ! i R ) r e a c t o r s .

C e l l c a l c u l a t i o n s

T h e c e l l c a l c u l a t i o n p r o g r a m s have b e e n u s e d i n t h e

e v a l u a t i o n o f t h e n e u t r o n d o s e s a b o v e 0 . 4 MeV i n v a r i o u s

l o c a t i o n s i n s i d e t h e r e a c t o r p r e s s u r e v e s s e l w i t h

d i f f e r e n t core c o n f i g u r a t i o n s .

Work h a s s t a r t e d o n c e r t a i n c a l c u l a t i o n s c o n c e r n i n g t h e

s a f e t y o f s p e n t f u e l t r z n s p o r t , s p e c i a l l y o n c a l c u l a t i o n

c f t h e n e u t r o n d o s e r a t e o u t s i d e a d r y s p e n t f u e l

c o n t a i n e r a n d o n t h e c r i t i c a l i t y b e n c h m a r k c a l c u l a t i o n s

f o r s p e n t f u e l t r a n s p o r t c a s e s a g r e e d on a t a CSNI

w o r k s h o p i n May.

Core c a l c u l a t i o n s

I n t h e BWR f i e l d t h e c o d e d e v e l o p m e n t work h a s m a i n l y b e e n

r e s t r i c t e d t o m o d i f i c a t i o n s r e q u i r e d by t h e c h a n g e o f

c o m p u t e r s y s t e m a n d t o a u x i l i a r y c o d e s n e e d e d t o p r e p a r e

i n p u t d a t a f o r t h e c o r e s i m u l a t o r BOREAS. U t i l i z i n g two-

g r o u p d a t a c o m p u t e d b y t h e CASMO c o d e , a number o f s u r v e y

s t u d i e s h a v e b e e n made c o n c e r n i n g t h e f i r s t f o u r c y c l e s o f

t h e TVO I a n d t h e f i r s t two o f t h e TVO I1 r e a c t o r

i n c l u d i n g i n v e s t i g a t i o n s r e g a r d i n g d i f f e r e n t c y c l e l e n g t h s

and burnable absorber contents in the fuel. A separate

simulation study on the first year of operation of the

TVO I reactor has also been performed which has made

possible some comparisons between the computed results of

BOREAS and the real state of the reactor core represented

by power distributions obtained by combining the results

of measurements and calculations made by the power station

computer. These comparisons have, on the whole, turned out

quite satisfactory with differences of 1-2 %, typically,

in the horizontal power distributions. Vertically, the

differences vary more; sometimes the code gives very

accurate predictions, sometimes it is less successful.

However, these differences are, on an average, usually

smaller than 5 %.

The testing of the three-dimensional PWR-simulator

HEXBU-3D has continued with operating data of the second

and third cycles of the Loviisa 1 reactor. The

comparisons between calculations and measurements for the

first cycle have been reported in the NEACRP Meeting in

Paris, November 1979. Results for the two subsequent

cycles are also good and mostly similar to earlier

a comparisons. Due to a slight overestimation of the lenght of the first cycle a correction factor for reactivity,

which adjusts the energy release per fission, was

introduced into the program. Reducing the energy release

by 1 % the simulation of operating history of the Loviisa

1 reactor gives the lenghts of the first three cycles with

an accuracy of about 2 %.

The treatment of thermal flux spatial transients between

neighbouring fuel assemblies has been modified in HEXBU-3D

to include an input coefficient for multiplying of the

transients. It was observed that, especially in the first

a n d s e c o n d c y c l e s o f t h e reactor, when f u e l e n r i c h m e n t

r a n g e d f r o m 1 . 6 % t o 3 . 6 8 , t h e power was s y s t e m a t i c a l l y

o v e r e s t i m a t e d i n a s s e ~ n b l i e s o f h i g h e r e n r i c h m e n t . The

r e d u c t i o n o.f t r a n s i e n t s d o e s i m p r o v e t h e s i t u a t i o n , b u t

e v e n a r e d u c t i o n o f 50 % w i l l n o t e n t i r e l y r emove t h e

d e v i a t i o n s f r o m m e a s u r e d a s s e m b l y p o w e r s . I t seems t h a t

a t l e a s t p a r t o f t h e d i s c r e p a n c y is c a u s e d by t h e

h o m o g e n i z a t i o n O F f u e l a s s e m b l i e s w h i c h i n a t w o - g r o u p

c a l c u l a t i o n c r e a t e s t o o b i g t r a n s i e n t s o f f l u x b e t w e e n

n e i y h b o t i r i n g a s s e m b l i e s . T h i s p r o b l e m is known f r o m

B N R - r e a c t o r s w h e r e , as i n W E R - 4 4 0 r e a c t o r s , w a t e r g a p s

a n d s h r o u d s s u r r o u n d i n g f u e l a s s e m b . L i e s make t h e s e

n e u t r o n i c a l l y i s o l a t e d f r o m e a c h o t h e r . The p r o b l e i n w i l l

b e s t u d i e d f u r t h e r .

Dynamic c a l c u l a t i o n s

T h e r e a c t o r d y n a m i c s p r o g r a m s TRAWA, TAPP a n d TRAB h a v e

b e e n u s e d i n t h e t r a n s i e n t a n a l y s e s o f d i f f e r e n t s i t u t i o n s

o n t h e L o v i i s a a n d O l k i l u o t o r e a c t o r s .

F o r t h e d y n a m i c c a l c u l a t i o n s a new p r o g r a m ODD h a s b e e n

made t o c r ea t e a x i a l l y o n e - d i m e n s i o n a l t w o - g r o u p

d i f f u s i o n p a r a m e t e r s a n d t h e r m o h y d r a u l i c f e e d b a c k

c o e f f i c i e n t s o n t h e b a s i s o f t h r e e - d i m e n s i o n a l s t a t i c

c a l c u l a t i o n s . N o w e v e n d e t a i l e d core m o d i f i c a t i o n s c a n b e

a c c o u n t e d i n t h e t r a n s i e n t a n a l y s e s . T h e r a d i a l

d i s t r i b u t i o n s o f t h e f a s t a n d t h e r m a l n e u t r o n f l u x e s , t h e

power d e n s i t y a n d t h e t h e r m o h y d r a u l i c v a r i a b l e s c a l c u l a t e d

by t h e c o a r s e mesh P W R - s i m u l a t o r HEXBU-3D a r e u t i l i z e d i n

t h e p r o g r a m ODD. T h e 1 - D g r o u p . c o n s t a n t s g i v e same

a v e r a g e a x i a l d i s t r i b u t i o n s as t h e t h r e e -

d i m e n s i o n a l c a l c u l a t i o n a n d a l so t h e c o r r e c t d y n a m i c a l

b e h a v i o r , i f t h e r a d i a l s h a p e s r e m a i n e s s e n t i a l l y

u n c h a n g e d .

In figure 1 the relative axial power distribution calculated by HEXBU-3D with 10 nodes axially and 349 nodes

radially in the end of fuel cycle is compared with the

stationary distribution calculated by TRAMA with 41 mesh points in 10 axial fuel regions. In spite of even the

dissimilar thermohydraulic models, the agreement is good,

the differences in the relative axial distributions are

usually smaller than 1 % and the maximum is below 2 %.

The development work, the goal of which is to eliminate

a the most of the existing thermohydraulic restrictions in the presant dynamic programs, e.g. flow reversals, has

been continued.

The number of available thermohydraulic correlations

describing slip between phases and evaporation or

condensation has been increased in the dynamic programs.

The correlations are comprehensively compared and with

them it is possible to cover a wide variety of situations

in different reactor types.

D 0

0 I 1 I 1 I I I I 1 I I I-- OaO0 0.10 0.20 0.30 0.40 0.50 0.60 0.70 0 -80 0.90 1 a00

H E I G H T F K R C T I O N F R O M B O T T O M O F C O R E

DRRW VTT/YOI 290280

- 37 - CCI4MISSARIAT A L'ENERGIE ATOMIQUE

Reactor Physics Act iv i t i es i n FRANCE

October 1979 - September 1980 23D NEACRP Meeting

September 22-26, 1980 - IDAHO I. BOUCHARD - Ph. HAMMER

1 - GENERAL. - The f r ench nuclear programme

meeting i n October 1979 f i v e new PWR's g r i d : TRICASTIN 1 and 2, GRAVELINES 1

NEACRP-L - 244 FRANCE

September 1980

i s going on s a t i s f a c t o r i l y . Since t h e l a s t NEACRP have been s t a r t e d up and covpled t o t h e EDF and 2 and DAMPIERRE I . They a r e 920 tW(e) a s -

t h e two FESSENHEIM and fbu r BUGEY u n i t s a l r e a d y i n ope ra t ion and i s twenty more u n i t s under cons t ruc t ion . The f i r s t 1300 MW(e) u n i t i s expected t o s t a r t up i n 1984. The load f a c t o r s of ope ra t ing p l a n t s a r e r a t h e r b e t t e r than expected and t h e time between t h e f i r s t s t a r t up and t h e f u l l power o p e r a t i o n bas been cons iderably reduced.

The c o n s t r u c t i o n of SUPER PHENIX 1 i s going on according t o t h e expected time schedule. The v e s s e l i s a r r i v e d a t CREYS-MALVILLE by J u l y 1980 and i s now i n t h e r e a c t o r bui ld ing .

PHENIX is operated s a t i s f a c t o r i l y and has a l r e a d y de l ive red more than 7 b i l - l i o n s K W ~ .

To prepare a foreseen o r d e r of two 1500 MWe f a s t breeders (SUPER PHENIX 2) important s t u d i e s a r e i n progress f o r decreas ing t h e c o s t of such u n i t s .

. The EURODIF p l a n t a t TRICASTIN has reached an o p e r a t i o n l e v e l corresponding t o 6 Mi l l ions of UTS pe r year . More than 100 T of LWR f u e l have been reprocessed a t LA HAGUE dur ing t h e l a s t s i x months campaign.

2 - FAST REACTOR 1'IIYSICS

During the period elapsed between October 1979 and September 1980, the major points of the French Fast Reactor Physics programme have concerned :

. the KACINE programme which is devoted to the study of neutronic charnctcristics of commercial breeders and in particular to problems related to heterogeneous core concept ;

. experiments performed on PHENIX in order to determine the feed back coefficients (Doppler, power, temperature) ;

fuel cycle studies ;

neutron shielding studies ;

. development of experimental technics for critical experiments and for Power plant operation.

It must be underlined that most of the experimental programmes concerning the fast reactor physics and shielding are now prepared and performed in the framework of the CEA-CNEN-DEBENE cooperation.

The present status of the French studies in fast reactor physics and shielding areas are described in detail in t:he paper presented at the ANS Sun Valley meeting (1). Therefore onlya summary of thisstatus will be given here.

An increasi~ig effort is going on to use PHENTX operation results for testing and improving the multigroup data sets and design calculational methods for various neutronic parameters :

- The discrepancy between the reactivity loss per day calculated with the CARNAVAL pseudo fission-product and the measured one has been completely analyzed : this discrepancy (9 5 ) was due to the fact that the axial fuel dilatation was not fully taken into account in the reactivity loss calculations. Presently che predicted value of this reactivity loss

and the experimental value are consistent by less that 5 %. This confirms the validity of Lhe pseudo fission product capture multigroup cross section adjusteil wi~li the integral experiments (including ERMINE and the PROFIL irradiation in PIENIX).

- Measurements of the PHENIX residual power after a reactor shut down have been performed in order to check and improve the data and calculatio- nal methods used forpredicting the residual power . For times going from 0 t o e 7 0 hours after the reactor shut-down, the residual power is presently .

overestimated by 3210 - + 5 % by the design calculational method. - The discrepancy between the calculated and measured compositions of

the Pu included in the unloaded blanket subassemblies is presently investigated.

- Systematic measurements of the feedback coefficients (temperature, powcr, Doppler) have been undertaken to check and improve the safety analysis of fast breeders. A companion paper at this meeting gives the preliminary results obtained (2).

2.3 - CRITICAL FACILITIES

After the PRE-RACINE programme, mainly devoted to the physics study of the heterogeneous core concept and performed within the framework of the CEA-CNEN cooperation (3), the RACIKE programme has started on September 1979.It is performed in the framework of the CEA-CNEN-DEEENE cooperation on fast breeders and involves the use of fuel provided by the three partners (4).

Presently one investigates the reference configuration which reached criticality on the 24th of March 1980. This configuration includes a central fertile zone (15 cm radius) and one fertile ring (10 cm thick).

DEBENE

280 (platlets)

450 (platlets)

Pu (kg)

U235 (kg)

.

CEA

220 (rodlets)

750 (rodlets)

CNEN

370 (rodlets)

/

Soiiic preliminary results obtained up to now are presented at the Sun Valley ANS ~~icctin]: (1) : the coinplete results will be compared to the previous rcsul ts obtained tliirinj; the P - C I N E programme(concerningthe clcan core and the configuraLion with one central fertile zone) in order to check and improve the calculational methods for heterogeneous cores.

The JASON programme devoted to new shielding concepts for fast breeders ( 1 ) has started at the beginning of 1900. This programme aims at improving the PROPANE formulaire, devoted to fast breeder shielding design calculations, for :

. new materials such as BqC special steels including high contents of Idi, materials including hydrogen (such as ZrIi2) ;

. new shielding concepts (e.g. localized shields). 0 Preliminary results of this programme will be presented at the

f~recomi%l~E~CRP specialists'meeting on shielding (Paris, October 27-29, 1960).

In order to improve the stuctural material nuclear data fo the version V or the CARNAVAL cross section Set two specific experimental programnes have been undertaken on the RB2 (Ci?314 - BOLOGNA) and ERXINE (CEA - CADAlUCtIE) fast thermal coupled facilities.

Both experiments are of the k o o = 1 type and the investigated media have been selected in order to fit to commercial breeder spectra.

For ERMINE the six month programme a:Llows to study three media (5) :

OUlO : The reference medium basic cell includes one enrichided U02 (27 %) MASURCA rodlet and three natural uranium oxyde rodlets.

OAlO and ON10 : The two basic cells include one enriched U02 rodlet, one natural uranium oxyde rodlet and respectively two steel or two nickel rodlets.

The measurements performd concern :

- reaction rate ratios (fission chambers and detectors), - reactivities using the oscillation technique.

2. 4 - IRRADIATED 1'UI:LS The irradiated fuel analysis in progress concerns :

- fertile pins irradiated in the core 1 of PHENIX, - U02-pu02 pins including a high content of higher plutonium isotopes

.. (TRAPU experiment), - PROFIL I1 experiment (mainly actinide sample irradiation).

2. 5 - THEORETICAL WORK - The major topics presently under study are :

- blanket calculational method development (6) - this development uses the results of a specific experimental programme (NEFERTITI) which started on TAPIR0 within the framework of a CEA-CNEN cooperation ;

- sensitivity code development for time dependant problems such as actinide and F.P build up (7) ;

- anisotropic.diffusion : the method developed for the treatment of the interface problemhas been extended to 2D problems (8) ;

- finite element method : this method is now applied to hexagonal 3D calculations and the corresponding code is being tested on a SUPERPHENIX type core.

2. 6 - DEVELOPMENT OF EXPERIMENTAL TECHNIQUES FOR CRITICAL FACILITIES AND POWER REACTORS

0 - The effort concerning the reactivity absolute measurement using the rod-drop technique is going on. - Within the RACINE programme, systematic comparisons of the fission or U238 capture rate measurements performed with different technics are made. The same inter comparison is made for Y heating measu- rement using either different technics or different TLD detectors.

Thc 'development of t h e NEPTUNE system of codes and i t s a s s e s s m e n t a r e g o i n g on and d u r i n g t h e l a s t y e a r t h e main e f f o r c s were devo ted t o improve t h e sys tcm f o r p r a c t i c a l a p p l i c a t i o n s and Lo check i t on power r e a c t o r f o l l o w i n g ( 9 , 1 0 , 1 1 ) Exper imenta l s t u d i e s were concern ing t e m p e r a t u r e c o e f f i c i e n t s , s t o r a g e c r i t i c a l i t y and s p e n t f u e l a n a l y s e s .

3 .1 - T h e o r e t i c a l s t u d i e s 3 .1 .1 - -- Improvements of NEPTUNE

CRONOS, f o r t r a n s i e n t 3 D c a l c u l a t i o n s A new module named CROFOS h a s been added t o t h e sys tem i n o r d e r t o

p r o v i d e a n e u t r o n k i n e t i c s c a l c u l a t i o n with. t h e t h e r m o h y d r a u l i c f eedback t a k e n i n t o accoun t d u r i n g t h e t r a n s i e n t s . In CRONOS t h e same f i n i t e e l ement t r e a t m e n t a s i n ELECTRE, t h e p a r t of NEPTUNE d e s i g n e d f o r t h e 3 D power d i s - t r i b u t i o n c a l c u l a t i o n s , i s used f o r t h e s p a t i a l r e p r e s e n t a t i o n of t h e f l u x . The expans ion c o e f f i c i e n t s of t h e f l u x depend on t i m e , t h e y a r e r e p r e s e n t e d by a one-s tep s c h m e .

CRONOS a l l o w s t h e c a l c u l a t i o n s of any 3 D t r a n s i e n t problem, a c c o u n t i n g f o r c o u p l i n g between t h e sub-channels of t h e thermol iydraul ic r e p r e s e n t a t i o n . I t