REACTOR PHYSICS ACT IVlTlES IN OECD COUNTRIES · REACTOR PHYSICS ACT IVlTlES IN OECD COUNTRIES ......

NUCLEAR ENERGY

AGENCY

COMMITTEE ON

REACTOR PHYSICS

REACTOR PHYSICS ACT IVlTlES I N OECD COUNTRIES

JUNE 1975 - MAY 1976

v- 0 0 c3 C4) b \O C O .

OECD NUCLEAR ENERGYAGENCY 38, boulevard Suchet, 75016 PARIS

NUCLEAR ENERGY AGENCY COMMITTEE

ON REACTOR PHYSICS

REACTOR PHYSICS ACTIVITIES IN OECD COUNTRIES

JUNE 1975 - MAY 1976

OECD NUCLEAR ENERGY AGENCY

38, boulevard Suchet, 75016 PARIS

REACTOR PHYSICS ACTIVITIES IN OECD COUNTRIES

This document is a compilation of the reports on reactor physics a c t i v i t i e s durina the oeriod June 1975 . Mav 1976. resented a t ~~ ~ ~ ~~

the Nineteenth ~ ; ? e t i n ~ ' o f the Committee (chalk ~ i & ' r Nuclear Laboratories. CANADA. June 1976) .

Australia .... Belgium .... Canada .... Denmark .... Euratom .... France .... Spain ....

.... Netherlands

I t a l y .... Japan .... U K ....

.... Switzerland

Norway .... Sweden .... United S ta tes .. Germany ....

(No progress reports were received f r h A u s t r i a . Iceland. Ireland and Turkey) .

NEACRP - L - 155 a (Australia)

EACTOR PHYSICS ACTIVITIES IN AUSTRALIA

JUNE 1975 - MAY 1976

D.B. MCCULMCH

Australian Atomic Energy Commission Research Establishment

Lucas Heights, NSW, ~ustralia

1. CRITICAL FACILITY EXPERIMENTS

The experimental program on the FC1 core has been completed. The

program included measurements of reaction-rate distributions using fission-

chambers and manganese foils, central reactivityworthsof a variety of

samples, neutron noise correlations, and some proton-recoil spectrum data.

The limitations to accuracy of reactivity perturbation measurements

arising from the reproducibility of 'table-closed' position,control-rod

settings, etc. were extensively investigated and it was established that

-6 a precision of - f 10 Ak can be achieved.

Detailed calculations using the AAEC code system AUS together with

predominantly ENDF/B-IV data are now underway for comparison with all the

experimental results. At this early stage it would appear that the

excellent agreement between the experimentally determined critical mass

and that predicted by the preliminary design calculations (using old

AAEC-GYMEA data and nominal values for such parameters as graphite density,

etc.) may have been somewhat fortuitous. The epithermal nature of the

core spectrum together with high leakage results in significant spectrum

variations throughout the whole assembly, thus making calculated parameters

ra ther sens i t ive t o the model adopted ( i . e . c e l l representat ion, energy

condensation procedure, e t c . ) . These e f fec t s a re current ly under

ingest igat ion.

Detailed s tudies have a l s o been made t o es tabl i sh the s e n s i t i v i t y

of calculated quant i t ies t o known uncertaint ies i n such parameters as

graphite density and absorption cross-section, fue l composition, e t c .

No new experiment has been mounted a t present , s ince the ant icipated

increase i n fue l inventory has not eventuated; t h i s necessi tates a f u l l

reassessment of the scope and d i rec t ion of a feas ib le fu ture program.

2. RBACTIVITY TRANSIENT STUDIES

2.1 ZAPP Code

The ZAPP point kinetics/plane geometry heat t r ans fe r code developed

f o r analysis of the SPERT 1 t rans ien t t e s t s , was extended t o handle

cyl indr ica l geometry, and a very simple model t o represent coolant flow

e f f e c t s was incorporated. The flow model permits zones t o be defined,

and a t each time s t ep , t r ans fe r s material from a speci f ied region of one zone

t o a specif ied region of another, a t a r a t e corresponding t o the flow

veloci ty external ly imposed by the code user. The material moved has the

average temperature of the zone from which it comes, and is mixed uniformly

with the material of the dest inat ion zone. The e f f e c t is a p a r t i a l

smearing i n a r a the r a rb i t r a ry manner, of the temperature p r o f i l e b u i l t up

i n each region. This imprecision necessari ly leads t o flow correct ions

which are a t bes t only semi-quantitative, and consideration i s now being

given t o implementation of more r e a l i s t i c flow models.

2.2 ZAPP Applications

ZAPP was used t o analyse the SPERT I1 B18/68 D 0 moderated core 2

t r ans ien t experiments, which included some forced flow data. Agreement with

experiment comparable t o t h a t obtained f o r SPERT I H20 moderated cores t

' Clancy, B.E. , Connolly, J.W. and Harrington, B. AAEC Reports AAEC/E345 (1975)and AAEC/E383 (1976) .

was again obtained.

The ~ ~ ~ c ' s 1 0 0 kW UTR-type reac tor MOATA uses f u e l p l a t e s generally

s imi lar t o those of the SPERT I cores, but having subs tant ia l ly higher

f u e l loading per p l a t e and wider coolant spaces between p la t e s . The ZAPP

model was used with r eac t iv i ty feedback coeff ic ients calculated using the

AUS scheme,to ca lcula te the reac tor response t o s t e p r eac t iv i ty inputs

f o r sa fe ty assessment purposes. The r e s u l t s mdica ted t h a t for f a i l u r e of

a l l i n s t a l l e d shutdown devices,a s t e p of 0.011 Ak/k could be to lera ted

without any p a r t of the fue l clad exceeding the melting temperature of

aluminium. However, the calculated core-only r eac t iv i ty feedback coeff ic ient

was -13.7 x Ak/k OC-I compared with -7.4 x Ak/k OC-I deduced from

measured r eac t iv i ty changes under pseudo-static conditions involving very

slow heating of the whole reactor . I t was therefore considered desirable

t o t e s t the ZAPP r e s u l t s by experimental measurements of some self-terminating

- 1 power t r ans ien t s i n MOATA, f o r i n i t i a l a down t o about 0.1 s f o r which

0

it was predicted t h a t the normal high power t r i p l eve l of 120 kW nominal

would not be exceeded.

A s shown i n Figure 1, experimental peak powers, P were i n very good max'

agreement with ZAPP (zero flow model), bu t the observed energy re leases t o

time of peak power, E were about 60% grea ter than those calculated. This t m suggests a s ign i f i can t hea t lo s s mechanism close t o peak power, r e su l t ing i n

broadening of the top of the power b u r s t , and l i k e l y t o be consistent w i t h

onset of na tura l convective coolant flow.

A crude estimate of convective flow e f f e c t s was obtained f o r the 16.5 sec

period t r ans ien t by mounting 2 thermocouples v e r t i c a l l y separated by 5 cm ,

i n the coolant above a cen t ra l f u e l element. A temperature f ron t was

observed a t a power of some 50 kW, w i t h a delay of -7 sec between the

thermocouples, suggesting an i n i t i a l flow veloci ty of -0.7 cm s-' , corresponding

t o a flow r a t e of some 0.7 9. s-l.

Although cons tan t flow r a t e i s a g r o s s l y overs impl i f i ed r e p r e s e n t a t i o n

of ensuing convective p rocesses , t h e ZAPP flow model desc r ibed above

could be t a i l o r e d t o g ive very good agreement wi th t h e observed power b u r s t

shapes by s u i t a b l e choice o f superimposed flow r a t e and power a t which it

commenced, i n a range commensurate w i t h t h e observed flow d a t a .

Overa l l , t h e r e s u l t s g ive confidence t h a t ZAPP p r e d i c t i o n s f o r t h e

f a s t e r t r a n s i e n t s which cannot be d i r e c t l y v e r i f i e d exper imental ly and f o r

which h e a t l o s s e f f e c t s have less s i g n i f i c a n c e , a r e v a l i d f o r s a f e t y

assessment purposes.

2.3 HIFAR Dynamics

Lack o f d e t a i l e d knowledge o f t h e heavy water flow p a t t e r n i n t h e HIFAR

r e a c t o r v e s s e l p rec ludes r e l i a b l e e s t i m a t i o n o f t h e time response of

temperature feedback under t r a n s i e n t cond i t ions . An experiment was designed

and c a r r i e d o u t t o measure t h e dynamic response of HIFAR, p a r t i c u l a r l y i n

t h e i n i t i a l few seconds o f a t r a n s i e n t .

A temperature t r a n s i e n t was induced by s h u t t i n g o f f t h e secondary

coo lan t flow t o t h e h e a t exchangers. The r e a c t o r power and t h e c o o l a n t

i n l e t temperature t o the f u e l elements were recorded a t 0.5 s i n t e r v a l s

dur ing the r e s u l t i n g t r a n s i e n t , which l a s t e d about 400 s, and inc luded

both t h e c l o s i n g and reopening o f t h e secondary coo lan t valves . R e a c t i v i t i e s

were c a l c u l a t e d from t h e power d a t a us ing an i n v e r s e k i n e t i c s code, and

cor rec ted f o r Xenon poison bu i ld up. Analysis of t h e d a t a us ing dynamic

models of t h e r e a c t o r showed t h a t no s i g n i f i c a n t temperature feedback

components were delayed by more than 2 seconds.

3 . PULSED INTEGRAL EXPERIMENTS I N HEAVY METAL ASSEMBLIES

3 .1 Experiments

The unusual behaviour of t h e ' ins tantaneous decay cons tan t s ' of both

235u and 2 3 9 P ~ r e a c t i o n r a t e s de r ived from Thorium s t a c k experiments i n

which L i (p ,n ) sources were used, has been, and s t i l l i s being, i n v e s t i g a t e d .

The ' ins tan taneous decay c o n s t a n t s ' vary q u i t e markedly w i t h t i m e a f t e r t h e

pulse and are qui te d i f f e ren t from the re su l t s of calculations. Even though

the experimental r e su l t s were repeatable a t the time of measurement,

subsequent measurements made a t a s ingle s p a t i a l location i n the thorium

assembly have not been i n agreement with those measurements made a t the

same locat ion i n the o r ig ina l experiment. Present indications are t h a t

the or ig inal experiments were subject t o some spurious e f fec t . Effor ts

a r e presently being made t o resolve this problem.

A quantity of depleted uranium (0.2% 2 3 5 ~ ) has recently been acquired.

Experiments with an assembly of this material a re being planned.

3.2 Sens i t iv i ty Studies

A program of calculat ional work i s proceeding w i t h the AAEC time-

dependent diffusion theory code TENDS. Calculations are being performed

t o indica te

(i) the s e n s i t i v i t y of pulsed experiments t o uncertaint ies i n current

nuclear data f i l e s .

(ii) the s e n s i t i v i t y of experiments t o various experimental parameters.

Various detector react ion r a t e s are being calculated fo r 232Th and

2 3 8 ~ assemblies w i t h Be(d,n) and a var ie ty of L i (p,n) sources.

The calculat ions show t h a t the time dependent decay r a t e , the parameter

measured experimentally, is primarily sens i t ive t o the value of absorption

cross sect ion i n the energy range below about 100 keV when a low energy

source is used, and t o the value of i n e l a s t i c cross sect ion a t energies

about 0.5 t o 5 MeV when a high energy source i s used. This type of

experiment i s therefore a useful source of in teg ra l data for f a s t reactor

applicat ions, t o augment o r c l a r i f y those available from more complex f a s t

c r i t i c a l assemblies.

The calculat ions showing the s e n s i t i v i t y of the experiments t o various

experimental parameters, a re used both t o c l a r i f y which parameters need t o

be closely controlled and t o evaluate whichare the most useful experiment

t o perform. Of current i n t e r e s t is the information t h a t might be obtained

using the organic s c i n t i l l a t o r NE213 as a detector with a well defined

neutron energy threshold, ra ther than as an energy spectrometer with good

timing resolut ion.

3.3 Enerqy Spectrum Measurements

Time of f l i g h t neutron spectrum measurements have been made f o r a

thick t a rge t Be (d,n) source a t deuteron energies of 1.4, 1.8, 2.3 and 2.8 MeV

and angles of 0, 30, 45, 60, 90, 120 and 150 degrees.

I n addition t o t h e i r use i n analysing pulse decay experiments, the

data have provided a check of calculated and measured ef f ic iencies of the

NE213 detector used f o r time dependent spectrum measurements i n the heavy

metal assemblies.

3.4 Gas Proportional Proton Recoil Spectrometry

Spherical detectors f i l l e d t o pressures of 1 atm. H2, 3 atm. H2, and

3 atm. CH have been used t o measure the angle dependent neutron energy 4

spectrum from a thick lithium ta rge t bombarded with protons i n the energy

range 2 t o 3 MeV. Good agreement was obtained i n the overlap regions of the

three counters. The spectra a l so agrees well w i t h calculat ions of the

thick t a rge t neutron energy spectrum based on measurement of the d i f f e r e n t i a l

7 7 * 7 cross sect ion fo r the L i ( p , n ) ~ e 7 and L i (p,n)Be reactions. This work

was undertaken pa r t ly t o prove the r e l i a b i l i t y of proton r e c o i l spectrometry

and pa r t ly t o provide experimental confirmation of the neutron energy

spectrum from the thick t a r g e t Li (p ,n) react ion which i s being used as a

source i n pulsed in t eg ra l experiments.

The thick t a rge t Be(d,n) react ion i s a l s o being used a s a source i n

such eqer iments . A s indicated above, the angle dependent spectrum of the

source has been measured by a time of f l i g h t method. Present indicat ions a re

t h a t gas proportional proton r e c o i l spectrometry can be used t o complement

these measurements i n the 0.04 t o 0.2 MeV region where the TOE has low

accuracy.

4. REACTOR CODES DEVELOPMENT

4.1 AUS Modular Scheme

Checkout of the AUS s u i t e of modules used f o r reactor calculat ions

continued. In pa r t i cu la r , several benchmark-type 1 and 2 dimensional k ine t i c s

calculat ions were undertaken w i t h the d i f fus ion module POW.

The work of inclusion of ENDF/B ( 3 and 4) data i n t o the AUS cross

sec t ion and k ine t i c s data pools has continued and i s almost complete.

4.2 AUS Module EDIT

A new EDIT module i s being wr i t ten which w i l l be able t o 'unscramble'

a succession of smearing and group collapsing operations t o give, f o r

example, t he f i s s ions due t o 235U i n a pa r t i cu la r p a r t of a reactor under-

going burnup. The user supplies very l i t t l e addit ional data a s EDIT uses

a l l the data of nuclide concentrations, smearing and group collapsing

f ac to r s , e t c . accumulated during the running of the job and held i n the

STATUS da ta pool.

4.3 AUS Module MIRANDA

The a v a i l a b i l i t y of ENDF/B cross sec t ion data has allowed improvement

of ce r t a in aspects of the MIRANDA resonance treatment. For nuclides with

subs tan t i a l resonance sca t t e r ing cross sect ions, the group t r ans fe r sca t t e r ing

cross sec t ions and group transport cross sec t ion have a resonance se l f -

shielding component included along s imi lar l i n e s t o the treatment of f u e l

nuclides.

An extensive comparison of MIRANDA w i t h the 128,000 group (20 KeV

t o thermal) numerical slowing down code PEARLS f o r 2 3 8 ~ captures i n simple

2-region cyl indr ica l geometry has been made. Results of t h i s comparison

are given i n Table 4.1 ( the t ab le was prepared p r i o r t o the ava i l ab i l i t y

of ENDF/B d a t a ) . It can be seen t h a t the agreement i s excel lent with

e r ro r s exceeding 1 per cent only f o r the l a rge r 8 cm diameter metal rods.

A c l u s t e r geometry co l l i s ion probabil i ty routine which involves repeated

cyl indr ica l calculat ions by the Bonalumi method has been included i n MIRANDA.

This routine has a l so been included i n the ICPP (general purpose co l l i s ion

probabil i ty) module, together with a number of more accurate and slower

co l l i s ion probabil i ty routines. The addit ion of c lus t e r geometry capabi l i ty

t o the AUS scheme is now almost completed.

4.4 AUS.ENDF/B Cross Section Library

The preparation of an AUS cross sec t ion l ib ra ry from ENDF/B data f i l e s

has progressed t o the stage where data f o r most important nuclides a re

avai lable and the addition of fur ther nuclides as required is f a i r l y automatic.

A number of new programs have been developed t o prepare data f o r fue l nuclides

i n the resolved and unresolved resonance regions. The methods used i n the

preparation of the l ib ra ry a re swnmarised i n the remainder of t h i s sect ion.

The ORNL code SUPERTOG i s used t o prepare group cross sect ions fo r a l l

data apar t from resonance cross sect ions and thermal sca t te r ing . SUPERTOG

'pr in ted ' output i s processed by the STOGAUS program t o produce an AUS cross

sec t ion f i l e . Scat ter ing matrices up t o P3 order have been prepared a s

standard.

Thermal sca t te r ing data f o r moderators (H 0 , D 0, CH , C ) have been 2 2 2

prepared from ENDF/B tabulat ions of S(a,B) smoothly joined t o the slowing down

data from SUPERTOG using the MERGER program. Thermal sca t t e r ing data f o r

other nuclides have been prepared by the AUSGAS program which uses a gas model.

A l l thermal sca t t e r ing matrices have been l imited t o PO and P1 data.

The bas ic approach t o resonance data i s t o prepare group resonance

in t eg ra l s as a function of po ten t i a l s ca t t e r ing and temperature, and t o f i t

these resonance in t eg ra l s w i t h subgoup parameters fo r use by the MIFX-JDA

module. In the resolved resonance region, poin t cross sect ions are generated

on an extremely f ine mesh in t e rva l using the POINTXSl program t o process the

s ingle-level o r multi-level Breit-Wigner parameters on the ENDF/B f i l e s . These

cross sect ions are generated i n the form of PEARLS tapes and the PEARLS

program was i n i t i a l l y used t o solve the slowing down equations numerically

fo r homogeneous mixtures of the resonance nuclide and hydrogen t o give group

resonance in tegra ls . The use of PEARLS proved ra ther time consuming on a

routine bas is and it has been replaced by the MINI-PEARLS program which solves

the same equations, but i s res t r i c t ed t o exactly the required problem and is

a fac tor of 4 f a s t e r . Tabulated sca t ter ing matrices are produced a s well as

group resonance in tegra ls . This process i s applied t o resonance sca t t e re r s

as well a s fue l nuclides.

Some d i f f i c u l t i e s have been experienced with the f i t t i n g of group

resonance in tegra ls with subgroup parameters and t h e i r subsequent use i n MIRANDA.

The i n i t i a l work i n t h i s area was performed fo r 238U data prepared by the

Br i t i sh GENEX code, which did not include interference sca t ter ing . The X

method has required some modification when interference sca t ter ing is

included and the accuracy of MIRANDA thus obtained has been established.

Two methods are available i n the unresolved resonance region, whose

use depends on the accuracy required and the data available. For most

2 nuclides, the approach used i n SUPERTOG and MC has been adopted. That i s ,

the resonance in tegra ls are computed from a sum of J functions with a

numerical integrat ion over the neutron and f i s s ion width d is t r ibut ions being

performed. This procedure has been coded i n the POINTXSl program. The

major approximations are i n the narrow resonance assumption and i n the

treatment of the overlap of neighbouring resonances, but the method i s

adequate f o r most applications. The a l te rnat ive approach of generating a

sequence of resonances by sampling from the width and spacing d is t r ibut ions

and hence preparing point cross sec t ions as i n the resolved region, has

been applied t o 238U only.

The prepared l ib ra ry , AUS.ENDF/B, is now the standard AUS cross sect ion

l ibrary . I t is of 128 groups which are iden t i ca l t o the GYMEA groups, apa r t

from an ex t ra group above 10 MeV (lethargy boundaries of -0.25 (0.25),

13.75, 13.9 (0.1) 20.5 (0.5) 23.0). The contents of the l ib ra ry are given

i n Table 4.2. A l l d a t a which are n o t ENDF/B3 o r ENDF/B4 have been e x t r a c t e d

from t h e GYMEA c r o s s s e c t i o n l i b r a r i e s . The l i b r a r y a l s o i n c l u d e s 41

i n d i v i d u a l f i s s i o n products and 1 pseudo f i s s i o n product from t h e AAEC

f i s s i o n product l i b r a r y .

4.5 AUS Module POW Kine t ics Checkout

A s p a r t o f t h e checkout o f t h e AUS 'workhorse' d i f f u s i o n module POW,

benchmark k i n e t i c s c a l c u l a t i o n s f o r 1 and 2D were undertaken. Resu l t s of

a comparison of POW with o t h e r codes f o r a 2D f a s t r e a c t o r benchmark wi th

t temperature feedback a r e given i n F igures 2-4. The comparison sugges t s

t h a t POW g i v e s acceptable r e s u l t s .

Flux T i l t i n Symmetric Systems Asymmetrically Disturbed

I n o r d e r t o understand t h e e x t e n t o f some o f t h e asymmetries a r i s i n g i n

t h e s p a t i a l f l u x d i s t r i b u t i o n from asymmetric absorber i n s e r t i o n i n some o f

t h e one and two dimensional k i n e t i c s c a l c u l a t i o n s undertaken a s p a r t of a

checkout o f POW, t h e fol lowing r e l a t i o n s h i p w a s der ived.

T = -pd/{l-d(1-p) } , p > - ( l -d ) / (2d) ,

where T = f l u x tilt i n d i c a t o r

- v x f i s s i o n s i n undis turbed h a l f - v x f i s s i o n s i n d i s tu rbed h a l f v x f i s s i o n s i n undis turbed h a l f + u x f i s s i o n s i n d i s t u r b e d h a l f

p = r e a c t i v i t y

- - dis tu rbed multiplication - undisturbed multiplication discurbed m u l t i p l i c a t i o n

and d = dominance r a t i o

- - second h i g h e s ~ m u l t i p l i c a t i o n eigenvalue undisturbed h ighes t m u l t i p l i c a t i o n eigenvalue

tt Without t h e (1 -p ) f a c t o r , t h e r e s u l t is t h a t obta ined by Wade and Rydin .

0

~ ~ ~ ~ - ~ p ~- ~ ~ -~

+ Buffoni, A. , F l e t c h e r , J . K . , G a l a t i , A . , McDonnell, F.N., Musco, A., and Vath, L. (1975) - Review o f Kine t i cs Benchmark Ca lcu la t ions . I n 'Proceedings of t h e J o i n t NEACRP/SCNI s p e c i a l i s t s ' meeting on new developments i n t h r e e dimensional neutron k i n e t i c s and review of k i n e t i c s benchmark c a l c u l a t i o n s . Technische Univers i ty a t Munchen, MRR 145, pp 505-551.

tt Wade, D.C. and Rydin, R.A. (1972) i n He t r i ck , D.K. (Ed.) - Dynamics of Nuclear Systems, Arizona Press .

Although ( l - p ) is usually close t o 1, the neglect of the fac tor f o r

extreme s i tua t ions (such as the very loosely coupled reactor reported i n the

Benchmark Problem ~ o o k ~ w i t h d = 0.99844) can reduce T t o as l i t t l e a s half

of t h a t calculated above (which agrees with t h a t calculated d i r ec t ly t o

within a few per cen t ) .

4.6 Comparison of Diffusion, S and Monte Carlo Codes N

Performance of the AUS diffusion theory module POW was compared with

t h a t of a Monte Carlo code KGNO f o r the calculat ion of the C r i t i c a l Fac i l i t y

configuration FC1, a small c - ~ ~ ~ u pseudo-cylindrical core with a graphite

r e f l ec to r . The calculated mult ipl icat ion fac tors agreed, but differences

of the order of 3 per cent were noted i n region averaged group fluxes. A

one dimensional spherical model of the system showed s imi lar differences

between region averaged fluxes from POW and SN transport (ANAUSN) calculat ions,

but the l a t t e r were i n good agreement with a Monte Carlo calculat ion. The

observed f lux differences were a t t r ibu tab le t o the POW code and were

consistent w i t h t he inherent d i f fus ion theory approximations.

A study of the c r i t i c a l mass and l i fe t imes of uranium assemblies

provided an opportunity t o compare the two one dimensional S Codes, N

ANISN and ANAUSN, and showed up some deficiencies i n both. A suggestion t o

improve the ANISN method of performing a calculat ions w i l l be made t o the

code authors.

t Argonne National Laboratory (1972) - ANL-7416, Supplement 1

- 16 - TABLE 4.1

COMPARISON OF RESONANCE CAPTURES IN 2 3 8 ~

Radius in cm

Fuel I Moderato! Metal Fugl

0.5 1

0.5 2

0.5 4

0.5 8

0.5 16

4 32

Oxide Fuel

0.5 0.7071

MIRANDA Error %

I D20 Moderator

Hz0 Moderator

0.24660 -0.5

MIRANDA PEARLS Error %

Graphite Moderator

UCLIDE

2 3 2 ~ h

23 3pa

2 3 2 ~

233u

23hU

235u

236u

2 3 7 ~ p

2 3 8 ~

2 3 9 ~ ~

~ " O P U

241m

2 4 2 ~ ~

CH2

H z 0

D20

C

0

Be0

N a

A 1

Cr

F e

N i

ZIRC2

6 ~ i

3 ~ e

3~

~ O B

N

ATA SOURCE t

ENDF/B4

PHF68

NDXD40

ENDF/B4

ENDF/B4

ENDF/B4

ENDF/B4

ENDF/B4

ENDF/B4

ENDF/B4

ENDF/B4

ENDF/B4

ENDF/B4

ENDF/B3

ENDF/B3

ENDF/B3

ENDF/B4

ENDF/B3

NDXD40

ENDF/B3

ENDF/B4

ENDF/B4

ENDF/B4

ENDF/B4

ENDF/B4

AEEW69

AEEW69

NDXD40

ENDF/B4

AEEW69

- 17 - TABLE 4.2

AUS.ENDF/B CONTENTS

SUBGROUP 'ARAMETERS

Yes

Yes

NO

NO

Y e s

Y e s

Y e s

NO

Yes

Yes

Y e s

Y e s

Yes

NO

NO

NO

NO

NO

NO

Y e s

NO

Y e s

Y e s

Yes

NO

NO

NO

NO

NO

NO

TEMPERATURES K

300

900,300

-

300

300

900,300

300

300

900,300

900,300

900,300

9OO.3OO

900,300

296

600,296

400,296

296

1200,900,600,296

1200,900,600,300

300

300

300

300

300

300

- - -

300

-

CATTERING ORDER

Cont inued

- 18 - TABLE 4 . 2 (Cont'd.)

+ NOTE: ENDF/B3: ENDF/B vers ion 3 d a t a f i l e - ENDF/B4: ENDF/B vers ion 4 d a t a f i l e

AEEW69 : w i n f r i t h d a t a f i l e 1969

BNL66 : BNL325, 1966

Others : Various o l d d a t a sources

SCATPERING ORDER

lo5 10

9 -

0 MEWRED P,,, - CALCULATED P- 8 - . MEASUF(ED E t m - . - CALCULATED E +

7 - - 7

6 - - 6

5- -5

4 - - 4 - r 'z r i 3- n.

W

2 - - 2

4 10 I I I I I I I I

0-02 0.03 0.04 005 O G 0.07 W.3 OCR 01

INlTlPL INVERSE PERIOD ( ?'I

F l C l CALCULATED h MEWRED VALES OF P,, h Em RXI MWTA

UK-RIS

0, OI 0 N F I ~ . ~ 2D fast reoctor benchmark I P I 0 0 0 2

\O

10

~ i g . 3 . 2D fast reoctor benchmark 1 power l l )

- UK-RIS

CRNL

--- GFK

- - - - - - - AAEC

F I ~ 4 2D fas t reactor benchrnork 1 ternp( t I

I I

NEACRP-L-155 b

(BELGIUM)

REACTOR PHYSICS ACTIVITIES I N BELGIUM

A report to the NEACRP. June 1976.

Compiled by J. Debrue, S.C.K./C.E.N., Mol

THERMAL REACTORS

BR3

The second par t of the i r r ad i a t ion of the t h i rd core (CR3/3B) was

completed a t the end of June 1975. This core was composed of 73 fue l

assemblies, 22 of which contain plutonium enriched fue l . The average

burn up i n the most heavily i r rad ia ted assembly reached 40,000 MWd/T.

The fourth core w i l l be operated during three i r r ad i a t ion cycles up t o

about 1980. The design of the loading BR3/4A, carr ied out by BELGONUCLEAIRE,

has been f inal ized. One th i rd of the pins contain plutonium and a r e

dis t r ibuted over three out of f i ve of the assemblies, giving to t he core

the general charac te r i s t ics of a recycle plutonium core. Among the 42

f r e sh assemblies, 40 a r e designed t o be dismountable i n the f u e l storage

well. Most of them a r e of the plutonium-island type; they contain

UO -Gd 0 pins. The r eac t iv i ty var ia t ion is thus controlled by burnable 2 2 3

poisons and, to a smaller extent, by boric acid i n the water moderator.

Detailed calculat ions of c r i t i c a l i t y power d i s t r ibu t ions and burn up a re

performed i n two dimensions (XY and RZ) by means of the CONDOR 3 code.

The nuclear data a r e generated by means of t he PANTHER code. The loading

of the reactor i s now completed. Neutronic t e s t s are being carried out

t o determine cold and hot c r i t i c a l conditions, r eac t iv i ty coef f ic ien ts

and control rod worths [I].

A mock up configuration representing plutonium-island assemblies has been

b u i l t i n the VENUS c r i t i c a l f a c i l i t y . Power d i s t r ibu t ions and 2 3 9 ~ u ~ 2 3 5 ~

f i s s ion r a t i o s were measured a t the most typ ica l locations, including the

U02-Gd 0 pins. 2 3

Power r e a c t o r programme

The r e a c t o r BR3 (PWR - 11 M e ) , s t a r t e d i n 1962, was t h e f i r s t r e a c t o r

designed i n t h e U.S. t o be exported from t h i s count ry . It has del- ivered

up t o now 420 GWh t o t h e g r i d , a l though t h e main i n t e r e s t l i e s i n i ts

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

SENA (PWR - 266 W e ) , s t a r t e d i n 1967, i s opera ted j o i n t l y by t h e f r ench

and be lg i an U t i l i t i e s ; t h i s was t h e f i r s t exper ience i n t h e u t i l i z a t i o n of

a commercial nuc lea r power p l a n t ; t h e e l e c t r i c i t y product ion is d iv ided

e q u a l l y between t h e two coun t r i e s .

A d e c i s i v e s t e p was r e a l i z e d i n 1975, w i t h t h e s t a r t up of DOEL I and I1

(PWR - 790 W e ) and TIHANGE (PWR - 870 MWe; 50% f o r France, 50% f o r

Belgium); from now on, t h e t o t a l e lec t ro-nuclear c a p a c i t y s a t i s f i e s

p r a c t i c a l l y 20% of t h e e l e c t r i c i t y demand i n Belgium. The cons t ruc t ion of

DOEL I11 and TIHANGE I1 is under way; ope ra t ion a t nominal power i s expected

i n 1980.

MATERIALS TESTING REACTOR BR2

The r e a c t o r i s r e g u l a r l y opera ted a t a nominal power of 80 MW according t o

a f o u r week o p e r a t i o d o n e week shu t down time schedule. The main i r r a d i a -

t i o n experiments t o be mentioned i n connect ion w i t h f a s t r e a c t o r developments

a r e t h e fo l lowing

- t h r e e ca rb ide p i n s i n a 250 kW sodium loop, completed i n September 1975

- 19 SNR type oxide p i n s i n a 500 kW sodium loop , under way s i n c e

June 1975

- 12 oxide p i n s i n a helium loop, t o be loaded around September 1976

i n t h e r e a c t o r

- 30 SNR type oxide p i n s i r r a d i a t e d dur ing a s h o r t per iod, followed

by blockage of t h e sodium dur ing i r r a d i a t i o n (MOL 7C); t h r e e t e s t s

of t h i s kind a r e planned, t h e f i r s t one i n November 1976.

The BR2 core loading w i l l be modified i n order t o s a t i s fy the specif icat ions

imposed by these l a s t two experiments. Extensive s tudies have therefore

been required.

The evolution of cracks i n the beryllium matrix continues t o be observed

regularly, since the f i r s t appearance of these defects i n March 1974.

The helium concentration, responsible for the damaging e f fec t , i s nearly

2% (atom f rac t ion i n Be) maximum a t present but no operation trouble has

resulted from t h i s phenomenon.

A new matrix i s ordered; although technically feas ib le from the end of

t h i s year, the replacement w i l l not take place before 1978.

FAST REACTORS

The par t ic ipat ion of C.E.N.1S.C.K. and of the belgian industry t o the

DeBeNeLux f a s t breeder project has been pursued i n d i f fe ren t areas.

I n collaboration with GFK and Interatom, an important e f f o r t is devoted

t o t he assessment of the performances of SNR core components, on the

bas i s of i r rad ia t ions i n Rapsodie, KNK-I1 and BR2.

A second neutronic analysis of the SNR-300 Mk I a core has been carr ied out,

according t o a modified configuration of t he di luent and control rod

system [2]. After s e t t i ng up the nominal lay-out, the objectives a r e t o

improve the performances of the reactor during normal operation, t o

calculate the plutonium e n r i c h e n t and t o provide data fo r thermohydraulic

and safety studies.

I n collaboration with U.L.B. (Universitg Libre de Bruxelles) i n t he

neutronic f i e l d and with a support from I.V.K. ( I n s t i t u t von Karman de

dynamique des f lu ides , Rhode-St. Gen'ese) i n the thermohydraulic f i e l d ,

C.E.N.1S.C.K. is elaborating a predisassembly phase code CASSANDRE :

t h i s code w i l l be integrated in to a common european core accident code

system of which the development i s sponsored by the Commission of the

European Comunit i e s .

C.E.N.1S.C.K. i s pursuing the construction of t he sodium loop designed

fo r the safety experiment MOL 7C i n BR2. This experiment i s aimed a t

the investigation of in-pile loca l sodium flow blockage, fue l f a i l u r e

propagation and molten f u e l sodium in te rac t ion i n a bundle of 30 pins.

The development and the standardization of dosimetry techniques i n

support t o the fue l s and materials i r rad ia t ion programme is, since the

end of 1975, the subject of a cooperation exchange agreement between

C.E.N.1S.C.K. and Hanford Engineering Development Laboratory. Dosimetry

applications [3] r e l y upon the work performed i n dosimetry benchmarks

(e.g. Big Ten, CFRMF, MOL-ZE); the present s t a tu s of the consistency

between in t eg ra l and d i f f e r e n t i a l data achieved i n these neutron f i e l d s

has recently been reviewed [ 4 ] .

The neutronic study of f a s t reactor s t ruc tura l materials is being pursued

i n the BR1 f a s t spectrum f a c i l i t y . ' ~ i and proton r eco i l spectrometers

a r e used t o measure neutron spectrum modifications i n one-material

assemblies of var iable thickness, i n one dimensional geometry. Different

cross section l i b r a r i e s a r e being considered i n the comparison theory-

experiment.

References

[I] Mesure de l ' e f f e t d ' inser t ion de barres de contrale

H. Bonet (Belgonucl6aure), P. Gubel (S.C.K.1C.E.N.)

Special is ts ' meetings on Experimental Techniques applied t o

Control Rod Measurements (Cadarache, Spring 1975).

[2] Use of compensation assemblies i n the f i r s t core of SNR-300

M. Billaux, R. de Wouters, S. P i la te , C. Vandenberg (BN)

H. Spenke, A. Stojadinovic, U. Wehmann (IA)

Trans. Am. Nucl. Soc. Vol. 20, 1975, p. 384

[3] Dosimetry work i n connection with i r r ad i a t ions i n the high f lux

materials t es t ing reactor BR2

J. Debrue, G. De Leeuw-Gierts, S. De Leeuw, Ch. De Raedt, A. Fabry,

L. Leenders, N. Maene, R. Menil (S.C.K.1C.E.N.)

ASTM-EURATOM 1st Internat ional Symposium on reactor dosimetry :

Developmentsand standardization

Petten, September 22-26 (1975) S.C.K. 1C.E.N. NN-493.

[ 4 ] Reactor Dosimetry in t eg ra l react ion r a t e data i n LMFBR benchmark

and standard neutron f i e l d s : s ta tus , accuracy and implications

A. Fabry, H. Ceulemans, P. Vandeplas (S.C.K./C.E.N., Mol, Belgium)

W.N. McElroy, E.P. Lippincott (HEDL, U.S.A.)

ASTM-EURATOM 1st Internat ional Symposium on reactor dosimetry :

Developments and standardization

Petten, September 22-26 (1975) S.C.K.1C.E.N. NN-495.

REACTOR PHYSICS ACTIVITIES IN CANADA

M.F. Duret and W.H. Walker

1. INTRODUCTION

The recen t o i l c r i s i s has focussed a t t e n t i o n on the energy

resources o f the w o r l d and concern has been expressed i n some quar te rs

t h a t p resent known and i n f e r r e d uranium reserves a r e n o t g r e a t enough

t o assure adequate growth r a t e s i n nuc lea r energy p roduc t i on f o r t he

nex t 20-30 years us ing o n l y the uranium cyc le . For a long term

s o l u t i o n many coun t r i es a re developing the f a s t breeder r e a c t o r . On

the o the r hand, t he re a re t h e op t im is t s , who b e l i e v e t h a t t he re i s

p l e n t y o f uranium - and thor ium - t o be found a t acceptable costs .

No one knows what t h e t r u e s i t u a t i o n i s b u t i t seems o n l y p rudent t o

be prepared f o r e i t h e r e v e n t u a l i t y .

Canada w i l l p robably n o t have t o en te r the f a s t r e a c t o r f i e l d ,

b u t w i l l be a b l e t o develop the CANDU r e a c t o r t o s t r e t c h o u t ou r uranium

resources by the use o f recyc led Pu and the Th-U f u e l cyc le .

Work in tended t o back-up t h i s development i s now underway,

i n c l u d i n g the b u i l d i n g o f a p i l o t p lu ton ium f u e l f a b r i c a t i o n l i n e a t

Chalk R i ve r and severa l computer s tud ies . These i n c l u d e the construc-

t i o n and c o s t o f spent f u e l processing p lan ts , waste storage, and the

use o f n a t U-Pu, Th-U and Th-U-Pu f u e l s i n advanced CANDU reac to rs .

2. REACTOR ASSESSMENT STUDIES

2.1 Pu Burners

The work on a 1200 MW pressur ized l i g h t water r e a c t o r w i t h Pu

enrichment, CANDU-BLW(PB), was fo l l owed by a s tudy o f a p ressur ized

heavy water r e a c t o r w i t h s i m i l a r f u e l . It was found t h a t zone c o n t r o l ,

a d j u s t e r and s h u t - o f f rods c o n t r o l 1.12% l e s s r e a c t i v i t y tham s i m i l a r

rods i n the s tandard CANDU.

The cos t ana lys is f o r t h e PHW(PB) i n d i c a t e s t h a t i t would

n o t be compet i t i ve w i t h the standard CANDU, i n c o s t per u n i t energy,

u n t i l ye l lowcake p r i ces reach an average o f about $30/lb. U3O8 (1975 8 ) .

2.2 Th-U f u e l

A s e n s i t i v i t y ana lys i s o f a CANDU r e a c t o r us ing Th-U f u e l i n

the s e l f - s u s t a i n i n g cyc le (%2.4%U w i t h 62%233, 23%234, 5%235, 10%236)

i n d i c a t e s t h a t the major u n c e r t a i n t i e s are f i s s i o n product absorpt ion

and 2 3 3 ~ thermal f i s s i o n cross sec t ion .

Ca lcu la ted f i s s i o n product absorp t ion f o r i r r a d i a t e d 2 3 3 U a re

12% g rea te r than measured f o r thermal neutrons. The two p i l e

o s c i l l a t o r experiments a r e i n agreement. D e t a i l s o f the comparison

are a v a i l a b l e i n d e s c r i p t i o n o f ENDF/B(IV) benchmarks (ENDF-230,

vo l . 1, pp V I 38 t o 42).

Unce r ta in t y i n the 2 3 3 ~ f i s s i o n cross sec t i on i s i n d i c a t e d by

a 1 .O% spread i n r e c e n t l y evaluated values - BNL-325(1973), ENDF/B(IV),

the recommendation f o r t h e ENDF/B(IV) no rma l i za t i on by the Normal-

i z a t i o n and Standards subcommittee o f the ENDF/B working group (CSEWG)

(Trans. A.N.S. - 18,351 (June/74)) and the p r e l i m i n a r y IAEA r e v i s i o n

(Lemmel N. B. S. Special Pub1 i c a t i o n 425 (Oct/75)).

3. REACTORS

3.1 Power Reactors

The major s ta tus change s ince the l a s t r e p o r t (NEARCP-L-120)

was t h e d iscovery o f leaks i n the Zr-Nb pressure tubes i n the no. 4

r e a c t o r a t P i cke r ing s i m i l a r t o those i n no. 3. A t o t a l o f 52 tubes

were rep laced du r ing a 10 month shutdown, and the r e a c t o r re tu rned

t o f u l l power March 25/76. Twenty-four o f the f o r t y enr iched f u e l

bundles mentioned i n NEACRP-L-120 are a l ready i n the Douglas Po in t

r e a c t o r and eleven of these have been s h i f t e d t o p o s i t i o n s o f h igh

power. No defec ts have occurred.

Commissioning experiments i n BRUCE-2 a r e now scheduled t o

beg in i n July/76.

3.2 SLOWPOKE Reactors

The f i r s t 2 o f 4 SLOWPOKE-2 reac to rs t o be i n s t a l l e d i n Canada

have been commissioned. These are a t the U n i v e r s i t y o f Toronto,

rep lac ing SLOWPOKE-1, and a t Ecole Polytechnique i n Montreal. The

o thers w i l l be loca ted a t Dalhousie Un ive rs i t y , Ha l i f ax , and the

U n i v e r s i t y o f A lbe r ta a t Edmonton.

Negot ia t ions cont inue f o r s a l e o f a 5th SLOWPOKE-2 f o r i n s t a l -

l a t i o n a t the U n i v e r s i t y o f Cologne.

A h igh power run was inc luded i n recen t t e s t s o f a SLOWPOKE-2

a t Commercial Products (Ottawa). The r e a c t o r was h e l d a t 100 kW

(F lux ~ 5 ~ 1 0 ~ ~ n / c m ~ s ) f o r 49 minutes. No i n s t a b i l i t i e s o r o the r

unusual behaviour were observed.

3.3

3.3.

been

AECL Research Reactors

1 ZED-I1

The 7-channel h o t p ressur ized loop and 2-phase absorber have

i n s t a l l e d and a r e now being commissioned. I n s t a l l a t i o n o f the

i o n exchange column f o r the poison i n j e c t i o n experiment i s n e a r l y

complete b u t the i n j e c t i o n system i s n o t ready. The t r a n s i e n t

experiments have been he ld up u n t i l the s p l i t core can be i n s t a l l e d

f o l l o w i n g complet ion o f o the r ZED-I1 experiments.

A 1•‹B de tec to r system i s now i n se rv i ce a t CRNL us ing ZED-I1

as t h e neutron source. I t i s capable o f de tec t i ng ' "B equ iva len t t o

0-5 mk r e a c t i v i t y t o b e t t e r than the requ i red accuracy o f + 1 mk.

Tests w i t h Am-Be and 2 5 2 ~ m neutron sources showed t h a t a po r tab le

ve rs ion f o r t e s t s o f CANDU r e a c t o r coo lan t us ing one o f these sources

would a l s o meet the requ i red accuracy (CRNL-1406).

3 . 3 . 2 - NRX

There have been no s t a t u s changes i n the pas t year.

3.3.3 - NRU

S i x loop p o s i t i o n s a r e now ava i l ab le . Four o f these are i n

use compared t o two p r i o r t o 1974.

The p o s s i b i l i t y o f rep lac ing n a t u r a l U f u e l w i t h Pu-enriched

Th i s be ing studied.

4. REACTOR DYNAMICS AND KINETICS

4.1 Hybr id Computers

Work cont inues on the s i m u l a t i o n o f steam d i v e r s i o n f rom the

G-1 r e a c t o r t o the LaPrade heavy water p l a n t . Cur ren t development

concerns pressure c o n t r o l o f G-1 coolant , and associated w i t h t h i s ,

a CDC-6600 s tudy o f the r e a c t o r core-steam drum-turbine-condenser

sys tern.

An 8-node r e a c t o r model i s being developed t o s tudy s p a t i a l

e f f e c t s i n loss-o f -coo lan t acc idents when t h i s a f f e c t s o n l y one

reg ion o f the r e a c t o r core.

Other work inc ludes s t a b i l i t y s tud ies f o r PHW(PB), c o n t r o l

o f WR-1 w i t h thor ium f u e l and a turbo-generator model f o r use w i t h

the steam cyc le o f any CANDU reac tor .

4.2 NEACRP/ANS Benchmarks

Work has cont inued on t h e 2D/3D BUR k i n e t i c problem s e t up

by Wagner a t Munich. It i s c u r r e n t l y h e l d up by a number o f

d i f f i c u l t i e s t h a t r e q u i r e c l a r i f i c a t i o n by the author. I t i s

expected t h a t these w i l l be reso lved by d iscussions a t the Toronto

ANS meeting.

4.3 CERBERUS

This program i s now i n an advanced s t a t e o f debugging, and

has a l ready been used t o so lve some 1- and 2-dimensional problems.

- 33 - Research Es tab l i shmen t . R i s @

Department of Reac to r Technology May 2 6 , 1975

Recent Reac to r P h y s i c s A c t i v i t i e s i n Denmark

by

fians N e l t r u p

1. I n t r o d u c t i o n

A q u i t e s u b s t a n - t i a l parxt o f t h e a c t i v i t y go ing on w i t h i n

t h e S e c t i o c f o r Reac to r p h y s i c s h a s been d e d i c a t e d t o s a f e t y

r e l a t e d work p a r t o f which u n f o r t u n a t e l y h a s l i t t l e b e a r i n g

on r e a c t o r p h y s i c s .

2 . F i s s i o n Product I n v e n t o r 1

The growing emphasis on problems of r e s i d u a l h e a t genera-

t i o n and r a d i a t i o n r e l e a s e h a s l e d t o t h e development of a new

f i s s i o n product b u i l d up r o u t i n e t o be i n c l u d e d i n burn-up

programmes. H i t h e r t o o n l y h a l f l i v e s l a r g e r t h a n 10 hours have

been c o n s i d e r e d i n a sys t em, where 1.64 f i s s i o n p roduc t s a r e

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

t i o n .

When s h o r t e r h a l f l i v e s , a t p r e s e n t down t o 10 minu tes ,

a r e i n t r o d u c e d a g r e a t e r f l e x i b i l i t y t h a t makes it more e a s y

t o i n t r o d u c e new d a t a and new n u c l i d e s becomes n e c e s s a r y . To

o b t a i n t h i s f l e x i b i l i t y numer i ca l i n t e g r a t i o n i s in t roduced

l e a d i n g t o e x p r e s s i o n s of t h e f o l l o w i n g form.

X(Z,N, t ) = S(X,N)(l-exp(-X9(X,N)(t-t0)f/X1(X,N)

+ X(Z,N, to)exp(-Xt(X,N))

where X(Z,N,t) i s t h e t i m e dependent c o n c e n t r a t i o n of a f i s s i o n

p roduc t c h a r a c t e r i z e d by a tomic number, 7 and a tomic mass N .

S(X,N) i s a te rm i n c o r p o r a t i n g t h e f o l l o w i n g mechanisms.

Direct yild from fission

isomeric transitions

neutron absorption

X' is d removal time constant including radioactive decay

L1~~d r~eutron absorbtion.

Introduction of new data or new nuclide into this system

will have little or no influence on the main body of allready

existing data.

In order to perform the numerical integration it is neces-

sary that S should be considered constant in the time interval.

The broad spectrumof time constants appearing in S is effectively

coped with by dividing the normal burn-up step At, in which

tlre flux is assumed constant, into subintervals starting with

a very small interval say ~t.2-lo and letting the interval

length increase by a factor 2 as a geometr-ic progression. In

this way the interval length is permitted to increase at a

rate corresponding to the time constants of concentrations that

have already become saturated.

3. Methods for Solution of Two-and Three-dimensioned Neutron

diffusion equation on power distribution

An investigation has been made of -the influence of the

S i u x representation on the rate of convergency .:f the fiiii-te

difference solution to the neutron diffusion equation. The re-

presentations considered are the mesh centre and the mesh cor-

nt:r repyesentation, but the investigation has been backed up

by use of the finite element representation described in ref.(].).

The results reported in ref.(2) show that both methods

,is predicted theoretically should converge assymptoticalLy for

decreasing mesh size in the same way, but that the convergency

of the cerrtri? mesh method shows some irregularities in cases

of Large mesh size,which can be explained at least qualitatively

from theory and which may give the center mesh method a slight

advantage in just these cases.

1 1 . Absorber Management -

Light -water r e a c t o r o p e r a t i o n i s based o n p e r i o d i c a l r e -

f u e l l i n g . During t h e c y c l e f u e l burn-up c a u s e s a r e d u c t i o n i n

r e a c t i v i t y , which can he c o u n t e r a c t e d by a g r a d u a l d e p l e t i o n

of a b u r n a b l e p o i s o n , e .g . gadol in ium.

T h i s d e p l e t i o n a f f e c t s t h e power shape and d i s c h a r g e burn-

up; and t h e r e i s a c o n f l i c t between r e q u i r e m e n t s f o r a minimum

power-pedking f a c % o r and a maximum d i s c h a r g e burn-up.

The problem i s t h e n t o f i n d t h e o p t i m a l ,,:?.y of d i s t r i b u t i n g

t h e b u r n a b l e p o i s o n , i . e . t h e d i s t r i b u t i o n g i v i n g munimum power

p r o d u c t i o n c o s t s , a l l c o s t s , i n c l u d i n g fuel-cyc1.e and c a p i t a l

c o s t s , a r e c o n s i d e r e d .

A s t u d y was c a r r i e d ou t on a two-dimensional model f o r a

PWR power r e a c t o r . The r e a c t o r c o r e was d i v i d e d i n t o two po i -

son c o n t r o l r e g i o n s and t h e r e l a t i o n s h i p between t h e power

peaking and burn-up was i n v e s t i g a t e d .

The r e s u l t s a r e i l l u s t r a t e d i n f i g . 1 where sub-index 1

r e f e r s t o t h e i n n e r r e g i o n and 2 t o t h e o u t e r r e g i o n . The d i f -

f e r e n t c u r v e s r e p r e s e n t burn-up h i s t o r i e s , f o r which a g i v e n

maximal power peaking f a c t o r , f , a p p e a r s d u r i n g burn-up. The

c u r v e f o r f = 1 .35 co r re sponds t o t h e d i s t r i b u t i o n o f po i son

a c c o r d i n g t o t h e Hal ing p r i n c i p l e , f o r which f i s l o w e s t p o s s i b l e

and i n c i d e n t a l l y c o n s t a n t d u r i n g burn-up. For l a r g e r f v a l u e s ,

two d i f f e r e n t s o l u t i o n s a r e p o s s i b l e a s i n d i c a t e d f o r f = 1 .57 .

The maximum g a i n i n d i s c h a r g e burn-up compared t o a homo-

geneous ly c o n t r o l l e d r e a c t o r i s 6.15%, whereas t h e power peaking

can be lowered by 3 0 % , i n d i c a t i n g a g r e a t e r p o t e n t i a l f o r c o s t

r e d u c t i o n by a r e d u c t i o n i n t h e power peaking f a c t o r .

5. Ifomogenisation of Fuel Elements with External Absorbers

In fuel element calcula-tion control basorbers are tr,edted

as external to the fuel element. The boundary is then repre-

scrit?d by a so called y-matrix, that establishes a linear- re-

l d lionsl~ip between the boundary f lux considered as a vector

with components after energy groups and the analogue vector

for the net current across the boundary.

For the calculation of keff and flux distribution inside

the fuel elements good resul-ts are obtained by this method.

It is also possible through this procedure to calculate the

total absorbtion and energy group transfere of the external

region. When cross sections for three-dimensional calculations I ( .,

are homogenized by flux weighting difficwlties arise because

the fluxes are not known in the nrternal absorbing areas. As

long as theese areas are small or almost black, then for the

normalisation over the entire fuel element theese fluxes may

be assumed zero.

In many cases, however, it may be an advantage to include

in the domain represented by the y- nlatrix other materials

than the absorber eg. water-gap and shroud. In this situation

the flux normalization has been solved with good results by

a simple extrapolation of the flux on the boundary into the

external domain.

As long as the diffusion constant is obtained by first

homogenizing the transport cross section no aditional problems

arise. In cases where a more direct homogenization of the dif-

fusion constant is needed it will be necessary to return to

the transport calculation (in slab geometri) which is used to

produce the y-matrix.

6. A a m i c Model of a BWR Nuclear Power Plant: - The digital program of a dynamic model of a BWK power

plant developed during the last three years has now been com-

pleted. The model includes a boiling water reactor, high-and

low-pressure turbines, moisture separator, reheater, condenser,

feedwater heaters and feedwater pump. All parts are treated

one-dimensional except for the nuclear part of the reactor

whicli i!; based on t h e p o i n t k i n e t i c s e q u a t i o n .

The model has demonst ra ted i t s a p p l i . c a b i l i t y f o r s t u d y i n g

both t r a n s i e n t s o c c u r r i n g d u r i n g normal o p e r a t i o n and t r a n s i e n t s

c a u s e d , f o r exampl.e, by t u r b i n e t r i p , l o s s of condenser vacuum

e t c . , t l ie s o - c a l l e d "abnormal" t r a n s i e n t i n c i d e n t s .

1)iffer.en-t c o n t r o l sys tems c l ~ a r a c t e r ~ i s t i c of a BWR n u c l e a r

power p l a n t have a l s o been s t u d i e d w i t h t h e model and p a r t i . c u l a r

emphasis has been l a i d on t h e r e a c t o r p r e s s u r e c o n t r o l sys tem

and t h e r e c i r c u l a t i o n f low c o n t r o l system.

References

1. I b M i s f e l d t , s o l u t i o n of t h e Mul t igroup Neutron

D i f f u s i o n Equat ions by t h e F i n i t e Element Method.

RIS6-M-1809.

2 . G . K . K r i s t i a n s e n , I n v e s t i g d t i o n of t h e Accuracy

of C e n t e r p o i n t - , C o r n e r p o i n t - , and Fin i te -e le rnent -

methods f o r S o l u t i o n of t h e Neutron D i f f u s i o n

Equat ion . NEACRP-L-149.

Reactor Physics Activities at jRL-ZV.?kTGr!l~Isprd

June 1975 - June 1976 R. NICK?

I. NUCLEAE DATA EVALUATION FOR SHIELDING APPLICATION

I. 1. To si.nplify cross section evaluation by giving maxi:xm output information for minimum input the grogran fJiGDZSTY was developed. It calculates all energetically possible reaction cross-sections and particle spectra within a nuclear decay chain initiated by a nuclear reaction.

The present version is based on the statistical nuclear model for nuclear reaction's and employs the optical nodel for the calculation of the partial widths for particle decay and the Blatt-Veisskopf single particle model for the decay.

All necessary nuclear data are automatically searched from an external library of fundanental data.

The programme is now running or1 an 1st-i 370/155. The reliability of the code is being tested by conparing it with the code STAPRE which is extensively used at Ispra for cross-section evaluation.

.. 1.2. Neutron-induced cross-section data and &' -spectra for the

isotopes: 34-133 Ba were extensively investigated in collaboration with the "Institut fuer 2adiamforschung und Kern2hysik" of the University of Vienna.

The work done up to now consists of three parts:

a) Search for the best values for the parameters of the nuclear models used and calculation of the cross-sections (n,~), (n,np), (n,pn), (n,2n) and (n,3n) for 1383a. The calculations of the (n, ,y) crass-section for 1 3 8 ~ 3 could reproduce many availablg experj roenral values . which differed by a .Factor twu to three from the data of the fission-product file on ENDFB/IV.

b ) Development of t he code CAFTRE espec i a l l y f o r the

calculation of t h e capture-cross-sectlons and 8 - spec t ra .

Th l s code w a s used t o c a l c u l a t e - spec t ra f o r 21 3 8 ~ a and

t o study t h e m f l u e n c e of t h e model used f o r t h e El - s t rength

function.

It w a s found t h a t t h e Brink-Axel model g i v e s b e t t e r agree-

ment w ~ t h t h e expenmenta l da ta .

c ) A s no expar lxen ta l d a t a f o r t h e neumon cap ture c ross -

s e c t i o n s f o r t h e 134-137i3a i so topes could be found a

systemat ic i n v e s t i g a t i o n on t h e dependence on va r ious model

pa rane t e r s w a s performed. Fur ther , a neutron o p t i c a l po-

t e n t i a l was worked out, which w i l l be used f o r t h e d e f i n i t e

eva lua t ion of a l l neutron induced c ross -sec t ions on 34-I 33Ba

11. SENSITIVITY SIUDI23 IN 3D-GEOUETBY

As already reported pr8viously, a program for sensitivity calculations in three dimensions is under development.

Zspacially in shielding calculations where ducts and other heterogeneities are of primary importance, two and three

dimensional sensitivity calculations are increasingly re- quested. Since shielding calculations for these problems are almost exciusively iione by Idonte Carlo cechiliq~es, %hz associated sensitivity studies also require the same methods.

One of the main problems, in this context, is the study of

space-dependent sensitivity factors in widely varying geometries and material compositions. For this reason, estimators allowing the calculation of fluxes at points will have to be introduced.

In the present approach one solves the "forward equation" and

samples the contributions to the different detector points from each collision point simaltaneously using the once-more

collided point estimator procedure.

'The sensitivity factors themselves are calculated by correlated sampling where the same tracks are used in the unperturbed and the perturbed case. Since in this technique the perturbed case

has to use the collision density functions of the unperturbed case, additional weight factors must be introduced to obtain

an unbiased result. This concept has been combined with a.

region dependent "expected leakage estimator" to improve sampllng in deep penetration problems.

The mathematical formulation of this problem. vms described by

Pd. Rief at the "specialists' meeting on sensitivity studies and shielding benchmark problems* at Paris in October 1375.

The method has been incorporated i n t o t h e TINOC-program.

So far mainly one and nu l t ig roup c a l c u l a t i o n s i n s p h e r i c a l

geometyies were performed and compared with equiva len t

ANLSN-SWANLAKE runs .

The comparison of t h e two methods gave s a t i s f a c t o r y r e s u l t s

f o r both , t h e s e n s i t i v i t i e s of t h e i n t e g r a l dose r a t e s as wel l

a s f o r t h e s e n s i t i v i t y p r o f i l e of t h e spec t r a .

111. HEACTOR SHIELDING

During t h e period i n quest ion, t h e ESIS (European Shie ld ing

Information Serv ice) a c t i v i t i e s progressed i n the f r a n e n o r ~

o f t h e follovring chap te r s : code assessment, nuc lear d a t a f o r

s h i e l d i n g ( i n co l l abo ra t ion with I N D A C ) t e chn ica l consu l ta t ion ,

sh i e ld ing d a t a bank and information dissemination.

I n t h e frame of t h e a c t i v i t y of code assessment developed

by ESIS, t h e performances of t h e Xonte Carlo code TlZI1OLI

have been t e s t e d by t h e study of a deep a t t e n u a t i o n problem.

An i s o t r o p i c monoenergetic plzne source emi t s p a r t i c l s s

i n t o a mediuri which i s 20 mean f r e e pa ths t h i ck . Scatker-

i n g i s i so . t rop ic ; t h e r e i s no slositing down ; absorp t ion

c ross -sec t ion equals h a l f of t he t o t a l cross-sect ion.

The same problem w a s a l ready solved with t h e d i s c r e t e

o rd ina t e code ANISN and with another Eonte Car lo code :

i t s s o l u t i o n i s now % e l l known with an accuracy of about

1 $, and w i l l be used as the "reference" so lu t ion . We

know t h a t t h e f l u x behaves approximately as ex (-KO.)

wi th KO = 0.037 c m ' ; so we have appl ied a s p a t i a l ixpor t -

ance func t ion exp (-x), ~ i t h va lues of I; ranging from

.06 t o 0.102.

Tables' 1 and 2 g ive a sumaary of t h e r e s u l t s . One of t h e

q u a n t i t i e s computed by th:: code i s t h e mew f l u x i n each

of t h e f o u r zones i n to which the medium was divided. "e"

i s t h e r e l a t i v e e r r o r ( 1 6 ) computed by t h e code with a

10 batches es t imate . "d" i s t h e percent d i f f e r e n c e con-

pared t o t h e r e f e rence resu l - t . Table 1 shows a comp~z-?ism

e vs. d i n t h e fou r zones f o r f i v e c a l c u l a t i o n s correspon-

ding t o dif'feren-i; K-values.

Table 2 shows t h e e f f e c t of K on t h e c o s t of ;the c a l c u l a t i o n .

For given va lues of t h e parameters t h e e r r o r of a c a l c u l a t i o n

i s propor t iona l t o I / ~ ( N = number of h i s t o r i e s ) and hence

t o l / f i (T = computing t ime) . The computing time needed t o

achieve a 10 $ e r r o r , i nd i ca t ed a s T (10 $) i n Table 2, f o r

a given resul t ; , i s roughly equal t o t h e product ll.e2.100.

The behaviour of T (10 $) vs. K i s shown i n Table 2. The

r e s u l t s of t h e t e s t case have provided u s e f u l i n d i c z t i o n s on

t h e performance of TRIPOLI ; t h e y show i n p a r t i c u l a r t h a t

t h e e r r o r va lues computed by TRIPOLI a r e r e l i a b l e and con-

s i s t e n t wi th t h e p r o p e r t i e s of t h e s tandard deviatjlon of a

normal d i s t r i b u t i o n .

More d e t a i l e d r e s u l t s and conclusions w i l l be presented at

t h e seminar-workshop i n t he TEIPOLI-code.

TABLX I : Z r r o r s ve r sus 'K of t h e meaQ f l u x by zone

K =. I0

e

2 0

e = r e l a t i v e e r r o r computed by TXIPOLI ($)

K =.030 K =.095 i e d : e d

I

1 0 I 1 -1

/ lone 1 K =.06 , ! K = . 0 8

2 1 5 -7 I 1 4 -1

-d / 5 -5 4 r 2 31 29 10 -11

d - - TRIPOLI-"Reference" . 100 Reference

N.3. = Rurcber of H i s t o r i e s

CPU = Computing time ( s e c ) .

I

I N.H. 3000

i C J J 1 76

i e d e d

4 1

5 0

7 - 4 1500

215

4 0

5 2

8 -3 3000

200

3000

123 b

1 / I 1 I 1 -1

2 -5 2 0

4 2

3000

273 I

TASLE 2 : Cost versus I<

T (10 $) = Estimated compidting time ( s e e ) requi red

t o achieve a 1.0 7; e r r o r on t h e t ransmiss ion.

Sensitivity s t u d i e s by S!;rANLA4KZ c o n s t i t u t e one of t h e keys

f o r providing a r a t i o n a l guidance i n s a t i s f y i n g cross-

s e c t i o n requirements. Sk?.UiLA!YE i s a ID s e n s i t i v i t y code

which c a l c u l a t e s t h e r e l a t i v e v a r i a t i o n of a response

func t ion f o r a one percent c ross -sec t ion v a r i a t i o n .

The programme has been implemented at I s p r a and appl ied

t o i r o n cross-sect ions . The r e s u l t s of t h e s e c a l c u l a t i o n s

have been presented at t h e Third Benchmark Meeting a t P a r i s ,

7-10 October 1975.

I n t e g r a l checks of i r o n group c ross -sec t ions have been made

us ing r e l a t i v e l y high energy neutron sources l i k e 2 5 2 ~ f

f i s s i o n neutrons ; experimental r e s u l t s have becorne ava i l -

ab l e from INR-Karlsruhe, where neutron leakage s p e c t r s were

measured from va r ious i r o n spheres surrounding a 25ZCf

. source . Ca lcu l a t i ons have been done ( i n c o l l a b o r a t i o n wi th

t h e sh i e ld ing Group of t h e Univers i ty of S t u t t g a r t ) t o check

t h e standard EURLIB group d a t a s e t of i r o n , which had been

generated from ET?DP/dIII. I r o n group cross-sect ions from

t h e newest ENDFP~IV d a t a were a l s o processed, so t h a t a l l

e f f e c t s could be s tud ied i n d e t a i l r e s u l t i n g from t h e neviest

po in t d a t a evaluated. . .

Using the nuc lear d a t a from ENDPj3, POPOP, and t h e G G - I 1

f i l e s a coupled neutron gamma l i b r a r y f o r 30 ele i2en. t~ and

nuc l ides was s e t up, toge ther wi th some a u x i l i a r y codes

f o r r e t r i e v i n g , co l l aps ing and p l o t t i n g . The l i b r a r y and

the subrout ines a r e a v a i l a b l e under t h e name EL4.

Bu.t t h e t r a d i t i o n a l 100 - groap - neutron s t r u c t u r a l

(GAX I11 + 1 thermal group) i s not wel l f i t t e d f o r deep-

pene t r a t i on sh i e ld ing problens. Therefore i n c o l l a b o r a t i o n

wi th IiK3 S t u t t g x t a n improved s t r u c t u r e w a s de-?eloped which

t akes i n t o account e x p l i c i t l y some prominent a n t i r e s o n a m e s

( f .i. those of Pe, C , 0 , K). These d a t a ( a g a i n i n 100 groups, and with soae standard co l l aps ing s t r u c t u r i : ~ pm- posed) a r e d i s t r i b u t e d as XJXIU.

R coupled neu-tron-plus-gama vers ion of iXRL$ri i s under

work ; i t con ta ins 100 neutron groups, too, p l u s 20 gamma

groups ( s i m i l a r t o t h e garma p a r t o f CASK). Standard cocles

a r e used f o r t he computations (.UiPX, POPOP, GAiZLEG).

111.2. Experimental a c t i v i t i e s

Becaase of t h e shut down of the r e a c t o r I s p r a I, i t w a s . . zecide,j 6 0 -i;ransZzr ;hi, skliel;li ,p i r ra j i ;3 t ion f a c i l i t y n,

EURACOS t o t h e TliIGB r e a c t o r of t he Univers i ty of Pavia.

I n a f i r s t phase t he EURACOS I1 w i l l be operate& a t a power

of 30 Watt. During 1974 the design of t h e new f a c i l i t y

w a s acconplished, whereas i n 1975 t h e main components

( i r r ac i i a t i on tunnel , mock-up) were constructed and trans- f e r r e d t o Pavia. The s t a r t up o f t h e f a c i l i t y i s fore-

seen i n t he f i r s t h a l f of 1976. I n a second phase, t h e

conver te r power w i l l be r a i s e d t o 300 ';!at-t. The f i r s t

experiment t o be performed concerns neu.tron propagation 3 i n an i r o n block ( 1 . 5 x 1.5 x 1.5 m ), i n the framework

o f t he Common B e n ~ b m r ~ Experimental Prograiixe executed

by t he Shie ld ing Groups of Viinfrith, Cadarache, Karlsruhe,

Mol, Casaccia and t h e Vnivers i ty of Tokyo.

111.3. Technical c o n w ~ l t a t i o n

Besides t h e cur ren t consu l t a t i on given f o r t h e use of

p a r t i c u l a r codes and l i b r a r i e s (DOT, !ZORSE, .SABIK3-3,

MEWUB3, EL-4), a nuraber of s t u d i e s and c a l c u l a t i o n s

were c a r r i e d out i n suppor-t t o t he s p e c i f i c needs of

Zuropean organisa t ions . I n p a t i c u l a r these s t u d i e s

concern neutron s t r e a ~ i n g c a l c u l a t i o n s a long t h e annular

gap of t h e PEC-reactor by DOT-DO;JII<O-XOilSI.:, and -the ca l -

c u l a t i o n of angular photon s p e c t r a a i d dose r a t e s i n i r o n

s h i e l d s due t o t h i n and th i ck sources (s tudy marle f o r t h z

German doard of Radiat ion P ro t ec t ion ) .

111.4. Information d i f f u s i o n

Through t h e e s t ab l i shnen t of a sh i e ld ing d a t a ban!.; ESIS intended t o speed up access t o and r e t r i e v a l of s h i e l d i n g

information. The bank handles b ib l iog raph ic i t e n s de-

l i v e r e d by t h e weekly scanning of p e r i o d i c a l s , r e p o r t s

and books. A t t h e present t i n e some 1500 bib l iographic

r e f e rences have been introduced.

The pub l i ca t ion of t h e ESIS Aewslecter proceeded accord-

i n g t o schedule : 4 Newslet ters were produced wi th con-

t r i b u t i o n s from var ious nucle&- es tab l i shments and in-

d u s t r i e s .

I V . CONZEPTUAL DESIGR OF A FUSION XZASTOR

IV.1. iieufronic probleas i n Fusion 243ctors :

Introduction- - - - - - - The development of t h e C o n c e ~ t u a l Design of a Tokonak

Exper inental Zeactor (FIiWOR) i n c o l l a b o r s t i o n with C3EB

( F r a s c a t i ) and the Univers i ty of X.?l_;js i s the c ? ? t ~ ? l

t a sk of t h e JFE-Ispra a c t i v i t y i n t h e f i e l d of Fmion

Reactorsf Nucleonics.

The ob jec t ive of the FIN'OX Z'roject i s the i d e n t i f i c a t i o n

of t he main physics and e n g i ~ e e r i n g l i m i t s t o t he ninimum

s i z e of an Zxperimental Power Xeactor, whose cons t ruc t ion

would precede t h a t o f a f u l l s ca l e prototype r eac to r .

The niain c h a r a c z e r i s t i c s of t h e PIiZ'GIi a r e :

- a se l f - sus ta ined plasma ( t h a t !neans a non dr iven p l a s m

hetited by fus ion p a r t i c l e s ) x i t h c i r c . ~ l a r c ross -sec t ion

and a s i n g l e n u l l d iverzor ; - t he continuous e f f o r t i n avoiding ad-~anced 'echnoiogical

s o l u t i o n s w h i c h r e q u i r e t o do y - o ~ i s i o n s f o r an aggressive

and unce r t a in development i n many f i e l d s , and i n t he near

term when an experimental r e s c t o r i s inten5ed t o be b u i l t ; - t o provide easy a c c e s s i b i l i t y f o r d a t a acqu i s i t i on , remote

maintenance and renewal.

One-dimensional Sh ie ld ing S tud ie s - - - - - - - - - - - - - - - I n an experimentel r e a c t o r of minimum s i z e , t h e major

problem from t h e nev:tronicsl po in t of view i s t h e design

of a minimum th ickness s h i e l d which f u l f i l l s t h e requi re -

ment of reducing t h e damage and t h e hea t depos i t ion i n t he

superconducting c o i l s t o acceptaule values .

The requirementr nf t he design a r e :

- a peak energy depo:",lon i n t he magnet l e s s than

1 x ~ / c m 3 ;

- radiat ion-induced r e s i s t i v i t y i n the copper s t a b i l i z e r

of t h e t o r o i d a l f i e l d c o i l s , and i n t h e d i v e r t o r c o i l s

l e s s than 2 x lo-'.

The l i m i t i n g geometr ical parameters a r e imposed b y plasma

balance condi t ions and by t h e maximum magnetic f i e l d i n

t he D-shaped t o r o i d a l c o i l ; t h i s corresponds t o t h e need

f o r a D-shaped blanket/shield conf igura t ion where t h e

i nne r s h i e l d occupies t h e reg ion of high magnetic f i e l d

and i t s dimensions a r e determined by t h e superconducting

magnetic p ro t ec t ion c r i t e r i a . The ou te r b lanket /shie ld

i s designed i n order t o achieve furthermore a t r i t i u m

breeding higher than one,. and t o convert t h e kine-t ic-

energy of f u s i o n neutrons i n t o recoverable hea t .

A l i m i t e d e f f o r t ha s been inade by t h e FIITTOR group i n the

set-up of c a l c u l a t i o n models, t h e modular code system used

i s t h e r e s u l t of a work of a c q u i s i t i o n t e s t and modif icat ion

of standard o r a l ready e x i s t i n g codes, and i n t h e i r con-

nec t ion i n t o a s i n g l e modular code system which a l lows a one run c a l c u l a t i o n of a l l t he responses requi red f o r t h e

design.

The one-D c a l c u l a t i o n s f o r t h e sh i e ld ing of she d i v e r t o r c o i l s which a r e t he components more exposed t o r a d i a t i o n

g ive a t h i ckness of 90 cm B4C t o minimize t h e displacement 2 i n copper d o m t o 4.6 1 0 d - y , g i v i n g :i irzimwn

r a d i a t i o n indu-ced r e s i s - t i v i t y (fr) af - te r one year of f u l l - 9 time ope ra t ion p r = 5.10 &-cx, t o be compared t o a,

8 maxinun nnagneto r e s i s t i v i t y of 3.6 10- ,% -cm.

The maximum nuc lear hea t ing i n t he c o i l i s i n t h i s case

2.2 x W/cm3 which i s very s m a l l c o q a r e d t o t he poss ib le ,Joule hea t ing t h a t such a ia.?gnez would be designed

f o r .

In the s t a i n l e s s s t e e l f irst wal l t h e number of dpa i s l e s s

than 3, a f t e r t h r e e y e a r s of Tu l l time operat ion. For

these f i g u r e s no problem e i t h e r f r o a s3!ielling o r from em- *&T.&ls <, a r e 2 ; c : ~ o ~ t e j i n the :;t-el.

L2 - Monte Gzrlo c a l c u l a t i o n s

' I n t h e s h i e l d i n g design of a Tokazlak exparimental r e a c t o r ,

s treaming of nuc lear r a d i a t i o n through t h e many necessary

pene t r a t i ons (e .g . d i v e r t o r channels, n e u t r a l beam i n j e c t -

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

I n order t o improve t h e design of t he neutron and gamma

sh i e ld ing inc lud ing t h e r e a l c o n f i g x a t i o n of t h e blanket

and t h e presence of d i v e r t o r s l o t s ne:xtral i n j e c t o r s ,

a Konte Carlo a n a l y s i s of t h e n a t r o n streaming i s being

performed by meas of t h e BOBSZ-Z code.

From t h e first r e s u l t s we can say t h e neutron streaming

i n t h e duc t s enhances t h e f l u x e s ou t s ide t'ne honogeneous

s h i e l d by f a c t o r s of 20 - 103.

IV.2. Radioac t iv i ty and a f t e r h e a t of FINTOR and r e l a t e d problems

of maintenance and waste d i sgosa l .

The r ad ioac - t i v i t y and decay hea t a f t e r shut down of t h e

316 SS s t r u c t u r a l components of FINTOR ( a minimum s i z e

experimental power r e a c t o r ) have been computed.

D i f f e r en t ope ra t ion t imes have been considered and the

evolu t ion of t h e r a d i o a c t i v i t y and decay hea t ve r sus t he

cool ing time have been compared with t h e corresponding

va lues of a comparable f i s s i o n power r e a c t o r .

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

at shut down of FINTOR (Ci/12: i th) i s more than one decade

l e s s than t h e f u e l r a d i o a c t i v i t y of t h e f i s s i o n p l an t .

The r ad ioac t ive decay i s r a t h e r slow ; t h e r a d i o a c t i v i t y

r e q u i r e s almost 3 y e a r s t o be reduced by a f a c t o r t en .

The inheren t r ad ioac t ive danger of FIXTO2 an3 of t h e

f i s s i o n r e a c t o r has been compared too.

The i j i o log ica l Hazard P o t e n t i a l of t h e 316 SS a c t i v a t i o n

i s loiv. I n f a c t , even t ak ing i n t o account t he t r i t i u m

which i s s t o r e d i n t h e FIXTOR p lan t , t h e environmectal

impact i s much l e s s than f o r a f i s s i o n p l a n t .

The s p e c i f i c decay hea t a t shut dorm (Ei1/'Kilth) of FIXTOR

i s one dscade l e s s than t h e corresponding value f o r a

f i s s i o n r e a c t o r .

A s t he decay h e a t of FIFTO?. i s spread over a g r e a t

quan t i t y of s t r u c t u r a l ma te r i a l s , no a u x i l i a r y cool ing

dev ices must be envisaged.

With t he va lues of t h e r a d i o a c t i v i t y a f t e r slut down,

t h e dose r a t e s i n t h e d i f f e r e n t zones of t h e FIIVTO2 p lan t

have been evaluated.

It can be seen t h a t behind the neutron and g a m a s h i e l d

t h e huxm i n t e r v e n t i o n i s poss ib l e without any sup;sler?ent-

a ry prc t ec t ion .

The a c t i v a t i o n of t he i nne r zones r e q u i r e s s 2 e c i a l

arrangements.

A s t h e plasma r e a c t i o n s r e q u i r e a vacuun chmber , a l l

t h e FIKIOR r e a c t o r i s enclosed i n t o a b i g vacuun ves se l .

I n f a c t , i t was recognized t h a t an easy disnountable

vacuuim t i g h t w a l l , as requi red by an experimental reactor ,

w a s no t p r a c t i c a l l y f e a s i b l e around the plasma c s v i t y .

Th i s extended vacuum zone c o n s t i t u t e s a good p ro t ec t ion

of ti22 envLr~n?ldnt aga ins t raCioac t ivs r e l a s e dwiilg :he

opera t ion . Koreover, even t h e minor disassembling

ope ra t ions a r e envisaged i n vacuum condi t ion.

The f a c i l i t y f o r nain-tenance and renewal of t h e ac - t i v s t e3

components c o n s i s t s of a dismounting shie lded t ruck which

runs on a r a i l system around t h e t o rus . A s to rage zone

f o r t h e daiiaged ac-t ivated components i s foreseen, wi th

t h e p o s s i b i l i t y of maintenance o r compactation i n view of

a f i n a l s to rage of t he wastes.

EXl'EBIMEN'TAL DETEPJJINATION OF THE INTEGRAL CAPTURE CROSS-

SSCTION OF STRUCTURAL TvIV4TERIALS I N THE 1 KeV-100 KeV ENERGY

3SGION, I N THE RB2/TV REACTOR, BY THX "FULL REACTIVITY METHOD" ("1

The compos i t ions o f t h e f o u r m i x t u r e s o f m i c r o s p h e r e s

measured u n t i l now are shown i n Tab le 1.

The measured v a l u e s f o r N 310/2p235 ( a t o m i c r a t i o ) of t h e

" n u l l r e a c t i v i t y " samples a r e :

mix tu re 1 2 3 4

N B10/#235 .383 - + 1 70 .374 + 1 $ .291 - + 1.7 $ .40d + 1

BlO ;;-U235) (Ni310/ijU235 The measured v a l u e s of BB = ( 5 c / 1, c o n v e r t e d t o t h e c o r r e s p o n d i n g " i n f i n i t e medium" v a l u e s ,

a r e shown below.

mix tu re 1 2 3 4

G i n f .996 .945. 1.016 .935

Owing t o s p e c t r a l mismatch an6 l e a k a g e , t h e k- inf of t h e

n u l l r e a c t i v i t y samples i s n o t e q u a l t o u n i t y , b u t

exp k- inf = 1/(1 - f' ) 00

The c o r r e c t i o n f a c t o r w a s c a l c u l a t e d by a 10-group 2-dl

d i f f u s i o n p e r t u r b a t i o n formal i sm (YOXO c o d e ) , as

S/iI = spec t ram a i s m a t c h t e r n

F/H = l e d s a g e - t e r n

The EZ~DF/EIIII va lues f o r u~~~ cap-ture and f i , s s ion were

normalized t o recen t va lues obtained by G.!IiS.

fle.terogeneity f a c t o r s ( g r a i n s t r u c t u r e ) were ca l cu l a t ed

f o r ii2j5, u~~~ znd. Fe ( s o f a r no t f o r C r , due t o uncer-

t a i n t y i n g r a i n s i z e ) .

c a l c exP A comparison or' k-iilf and k-inf i s given below.

mixture 1 2 3 4 c a l c

k-inf .9963 .9903 .9841 .9349

exp k-inf ,9876 . g a l 9 .9683 .9745

s $35 The r e s u l t s of t he exper inent , i . e . t h e r a t i o s Ec/Ff

i n the i n f i n i t e nedicn spec-krum of mix ta res 2 t o 4 , as

i n f e r r e d from the measured va lues presen-ted above and

t h e ou t l i ned ca l cu l a t ed co r r ec t ions , a r e shown below.

Also given a r e t he t h e o r e t i c a l va lues d i r e c t l y ca l cu l a t ed

from the ENDF/~I I I (21 f o r Pe) f i l e .

mixture 2 3 4 (I$"/#= 0 . 4 ) ( N ~ ~ / R ~ = O.d) (&fr/#= 0.4)

experimental .002b2+ 1%; - .00262+20$ - .00143+25$ - ca l cu l a t ed .00433 .00404 .GO5 20

The measured value f o r Pe i s considerably lower thran

t h e t h e o r e t i c a l value pred ic ted by t h e EY\rD?'/X f i l e , i n agreement wi th i n d i c a t i o n s obtained ersewhere.

The s t r i k i n g disngreenent f o r C r induces t o recons ider

wi th a d d i t i o n a l a t t e n t i o n t h i s measurement.

It i s s t r e s s e d t h a t t h e r e s u l t s presented a r e i n a pre-

l iminary form.

TABLE 1 : Atomic concentrations (atoms/cm3 x .lo- 24)

of tbe null reactivity samples

In t roduc t ion - - - - - - - Disposing of a c t i n i d e s o t h e r s than f u e l together wi th

f i s s i o n product waste i n geo log ica l formations involves

p o t e n t i a l s a f e ty r i s k s f o r f u t u r e genera t ions of markind,

and i s connec-ted wi-th d i f f i c u l t i e s : f i r s t l y , of p red ic t -

i n g the t e c t o n i c s of geo log ica l f o r m t i a n s f o r tine

pe r iods up t o hundreds of thousands of g e a r s ; secondly,

of imagining t n e f u t u r e reqv-irements of nankind t o t h e

s o i l - water - air system i n geo log ica l t i n e per iods .

So far, a c t i n i d e s waste may be considered as a p a r t i c u l a r

ca-tegory of r ad ioac t ive work t o be t r e a t e d i n a s p e c i a l way.

I n order t o c l a s s i f y t h e t e c h a i c a l and economical a spec t s

impl icated i n t he choice of d i f f e r e n t d i sposa l p o l i c i e s , t h e

following work has been performed i n t h e f i e l d of r e a c t o r

physics.

Burn-up ca lcu la . t ions providing t h e build-up of a c t i n i d e s

o the r than f u e l , l i k e e.g. protac-tinium, neptunium, a r n e r i ~ i ~ n ,

and curium i n a l l i n t e r e s t i n g c h a r a c t e r i s t i c r e a c t o r t ypes

were done by means of t h e computerprogram CRIGEN. The t o t a l genera t ion r a t e s of h igher a c t i n i d e s i n grams per t o n of f r e s h f u e l i r r a d i a t e d t o t y p i c a l burn-iip r a t e s r e s u l t as :

iT - n>!a Pu-LWR Xagnox AGR SGEPiiII THTR ~ ~ 1 ~ 3 3

6 6 0 5 200 3 5 220 210 2200 1460

The nuc lear l i b r a r i e s f o r t he B r i t i s h r e a c t o r types l i k e

Stea?l-Generating Xeavy Yiater-IJoderated E',cnctor, Advanced Gas-

Cooled Reactor, and Magnox Reactor were generated such

t h a t t h e corresponding l s o t o p i c composition of t h e burnt

fue1 , ca l cu l a t ed by t h e more r e f i n e d burn-up program

HYLAS-11, could be reproduced by ORIGSN wi th in reason-

ab l e e r r o r l i m i t s .

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

of a c t i n i d e s wastes would be t o r ecyc l e them through t h e

r e a c t o r s and t o transmute them t o l e s s hazardeous radio-

a c t i v e species . One of t h e requirements t o be f u l f i l l e d

i f a c t i n i d e s waste r ecyc l ing i n normal power r e a c t o r s i s envisaged c o n s i s t s of keeping small t h e a c t i n i d e s waste

inventory i n the r e a c t o r core .

Corresponding c a l c u l a t i o n s show t h a t i f t h e 1460 grams of

h i g h e r a c t i n i d e s which would be generated per t on of core

f u e l by t h e German NB1 fast breeder design, a r e recyc led

continuously i n s p e c i a l a c t i n i d e s ' f u e l elements, an

equi l ibr ium concerning mass and neutron-physical behaviour

could be reached a f t e r 25 cycles . I n t h e equi l ibr ium s t a t e

t h e a c t i n i d e s waste inventory i n t h e r e a c t o r amounts t o

about 9000 grains per t on of core f u e l ; t h e burn-up r a t e

equa ls t o about 16 '$ per cycle and t h e s p e c i f i c power

(Watts per graix heavy i so topes ) i s double a s high as f o r

normal core f u e l elements.

I n order t o s a t i s f y t h e requirements regard ing t h e hea t

t r a n s f e r , t h e a c t i n i d e s waste f u e l p i n s .have t o be designed

such t h a t t h e i r e f f e c t i v e dens i ty per *mi- t l eng tn of t h e

f u e l element becomes reduced. I n t h i s case aboxt 6 p i n s

of each NAl f u e l element subassembly, which c o n s i s t s of

331 f u e l p ins , would only have t o be rep laced by t h a t

con ta in ing a c t i n i d e s ' waste.

- 6 1 - .

. .

S e n s i t i v i t ~ S t u d i e s - - - - - - - - -. 11 d e t a i l e d s tudy on t h e s e n s i t i v i t y o f

- t h e q u a n t i t y of a c t i n i d e s o t h e r t h m Fd.el b e i n g gene-

r a t e d by o p e r a t i n g d i f f e r e n t r e a c t o r t y p e s ; - t h e long-term r a d i o t o x i c r i s k r e p r e s e n t e d by t h o s z

a c t i n i d e s i s o t o p e s ; - t h e t r a n s m u t a t i o n r a t e o f a c t i n i d e s ' was te r e c y c l e d

t h r c q ' n ;il f a s t b,-e?cier re . ic toy

t o n u c l e a r d a t a h a s been performed.

The u n c e r t a i n t i e s i n b a s i c n u c l e a r d a t a and p o s s i b l e e r r o r s

i n t r o d u c e d by t h e o r e t i c a l approx ima t ions have been d e t e r -

mined and e s t i n a t i o n s f o r the r e l i a b i l i t y o f s h e a'oove-

ment ioned q u a n t i t i e s have been c a l c i l l a t e d .

A s example, i t h a s been shown t h a t t h e t r a n s m u t a t i o n r a t e

o f a c t i n i d e s f w a s t e from a fast b r e e d e r , recycl .ed t h r o u g h

t h e sane r e a c t o r , depcnds i n t h e s a e measu-re on t h e f i s s i o n

c r o s s - s e c t i o n o f Tip-237, .%?1-241, and Am-2b73. '

A t e q u i l i b r i u m c y c l e , however, t h e weight o f t h e s e f i s s i o n

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

r a t e d i m i n i s h e s s t r o n g l y &XI o t h e r c r o s s - s e c t i o n s become

i m p o r t a n t , t o o , as. e.g. t h e c a p t u r e i n >a-241 and Cn-244

and Np-237 and t h e f i s s i o n i n 2u-233, Ccl-242, 'Jn-24.4, and

Cn-245. . .

VII . SSF%GU!UiDS

I n t h e frzme of t h e "Safegdards" progranme, fix-ther

development of t h e s t u d i e s on i s o t o p i c c o r r e l a t i o n s has

been pursued.

"Isotopic-corre la t ion- technique" i s t h a t branch of r e a c t o r

physics which i n v e s t i g a t e s t h e r e l a t i o n s h i p s (commonly

termed c o r r e l a t i o n s ) between accmulat iolz and d e p l e t i o n

of t he d i f f e r e n t i so topes i n nuc lear f u e l s sub jec ted t o

i r r a d i a t i o n . The i so topes which a r e taken i n t o account

a r e those of t h e heavy elements (U, Pu) and those of t h e

f i s s i o n product e lenents .

An extensive t h e o r e t i c a l i n ~ r e s t i g a t i o n has been performed

on those c o r r e l a t i o n s which r e l a t e t h e heavy i so tope

content of t he f u e l a f t e r i r r a d i a t i o n (Pu and U-isotopes)

t o t he f i s s i o n products build-up.

The f i s s i o n products which have been s e l e c t e d f o r t h e

stufiy ware t h e s t a b l e i s o t o p e s of K r , Xe, Nd and t h e radio-

ac-t ive nuc l ides ~ s ~ ~ ~ , C S ' ~ ~ , The reason f o r t h i s

choice l i e s i n t h e f a c t that experimental d a t a on these

iso-topes a r e a l ready ava i l ab l e .

Twenty-two c o r r e l a t i o n s have been formed, r e l a , t i ng t h e

f u e l burn-up o r t h e Pu build-up i n t h e f u e l t o one of t h e

fol lowing i s o t o p i c r a t i o s :

An e x t e n s i v e a n a l y s i s has been made on each c o r r e l a t i o n

f o r d e t e r m i n i n g , i t s depenfience upon v a r i o u s r e a c t o r

l a t t i c e p a r a x e t e r s s u c h as f u e l i n i t i a l emic 'men t ,

modera to r / fue l voliune r a L i o e t c . . Oaly PiiE l a t t i c e s

h w e bee?? con.si:lered, with .a f u e l i n i t i d enric!?~ent;

r a n g i n g from 2 "/o t o 4 '"/o, modera to r / fue l volume

ra- t io r a n g i n g ;ram 1.2 t o 2 . 2 , burn-v.p r a n g i n g from 0 Lo

40.000 ~i%d/.t znd Z i r c a l l o y c l a d d i n g .

A s a r e s - d l t , t h e c o r r e l a t i o n s which a p p s a r more 2 romis ing

f o r n u c l e a r m a t e r i a l c o n t r o l pu rposes have been s e l e c t e d .

VIII . ONE-GROUP CROSS-SZCTION DETE~YTINATION FROM '!IEASURE~ PVZL - C0:POSITIONS OF IRRADIATED FUZL

D e s t r u c t i v e and n o n - d e s t r u c t i v e t e s t s o n i r r a d i a t e d nu-

c l e a r f u e l d e l i v e r s e x p e r i m e n t a l i n f o r m a t i o n o n burn-up

and i s o t o p i c composi t ions . The g o a l o f t h e s e measure-

ments i s a comparison o f e x p e r i m e n t a l burn-up d a t a w i t h

t h e o r e t i c a l o n e s o b t a i n e d from burn-up codes .

A s imple c o n f r o n t a t i o n o f measured and c a l c u l a t e d d a t a

r e s u l t s i n a s t a t e m e n t : "The c a l c u l a - t i o n s a r e good o r

n o t goodw.

I n o r d e r t o a m e l i o r a t e improve burn-up c a l c u l a t i o n s ,

a f u r t h e r a n a l y s i s o f t h e measured burn-up d a t a i s worth-

w h i l e , e s p e c i a l l y when t h e e x p e r i m e n t a l d a t a may be de-

f i n e d as c l e a n d a t a . C l e a n d a t a mean : i r r a d i a t i o n in,

as much as p o s s i b l e , t h e s m e asympto t i c environment w i t h o u t

p e r t u r b a t i o n s from c o n t r o l r o d s a. s.0.

From such d a t a one c a n d e f i n e :

1. One g roup c r o s s - s e c t i o n r a t i o s

v i y r . = -

l Y L T . ~ J

i n which iY i s t h e c r o s s - s e c t i o n f o r r e a c t i o n i

( a b s o r p t i o n , f i s s i o n , a. s.0. ) f o r t h e i s o t o p e y , and X . 1s t h e r e f e r e n c e c r o s s - s e c t i o n o f r e a c t i o n j f o r

j t h e i s o t o p e x .

2. The mean r e a c t o r spec- t run nay be de t e r f i~ ined o u t o f

t h e c r o s s - s a c t i o n r a t i o s by means o f un fo ld ing codes.

A comparison of t h e one-group cross-sect ion r a t i o s fwith

those used i n t he calcula i ; ions , and of t h e mean r e a c t o r

spectrum wi th t h e ca l cu l a t ed one, w i l l psrmit t o t e s t

and improve bas i c d a t a used as input i n burn-up codes.

I n th is way these t e s t s a r e done i n a b e t t e r and more

e f f i c i e n t way than by ju s t confront ing t h e experimental

burn-up and i s o t o p i c composition d a t a wi th ca l cu l a t ed

ones.

Such = a l y s i s i s then a t m o s t e f f i c i e n t , t he more

experimental d a t a a r e ava i l ab l e , covering a wide burn-up range, and as much as poss ib l e d i s t r i b u t e d homogeneously

over t h i s burn-up f i e l d . As mentioned a l ready, t h e burn-

up d a t a should come f r o n the asymptotic environment of

t'ne r e a c t o r .

3 The i s o t o p e s measured ( i n at/cm ) a r e f o r U-fueled " )

r e a c t o r s :

1. from t h e U235 chain :

U235, U236, Np237, ~ ~ 2 3 8 , Pu236 ;

2. from t h e U238 chain :

U238, Pu239, ?u240, Pu241, Pu242

Am241, Am242, h 2 4 3 , Cm242, Cn244 ;

3. f u r t h e r U234 ; 4. burn-up (Csl3'7, Nd148, h e a ~ y elements) .

These d a t a a r e measured normally except Np237 and Pu236.

* ) f o r Pu-fueled and ot'ner fue l ed r e a c t o r s t he l is t has t o be re -e labor l ted i n funckion of t h e i n i t i a l f u e l compos i t i on .

A code i s w r i t t e n and experimental d a t a coming from

s t a i n l e s s s t e e l clkdded PWR f u e l (BU range : up t o

25.000 W#D/I) a r e analysed. Some d i f f e r e n c e s a r e found

wi th a c t u a l l y used cross-sect ion r a t i o s by t h e f u e l

management groups.

The experiments tend t o a s l i g h t l y harder spectrum than

t h e spectrum used i n t h e ca l cu l a t i ons .

To search f o r s y s t a n a t i c e r r o r s I n t h e code, due L J t he

approximations, t h e o r e t i c a l experiments a r e i n course.

I X . EiTZRGY STRATZGIES

?luring t h e perioj. ?'me 1975 - IIay 1976 the f o l l o ~ i n g ac-

t i v i t y r e l a t e d t o t he use of nuc lear energy w a s performed

i n t h e frame of t h e s t u d i e s on t h e Energy System.

1- Completion of t h e p r o g r m e TOTEM f o r t h e eva lua t ion

of e l e c t r i c energy genera t ion p o l i c i e s . Th i s pro-

g r m x e c a l c u l a t e s t h e power s t a t i o n i n s t a l l a - t i o n pol icy

corresponding t o energy demand and load diagram given

as a func t ion of time. The phys ica l c h a r a c t e r i s t i c s of

t h e va r ious t y p s s (up t o 10) of power s t a t i o n s as wel l

as va r ious c o n s t r a i n t s determining t h e pol icy a r e a l s o

given. The programme output i nc ludes Uraniiun, Separa t ive

work and f i s s i l e f a e l den an.'^, expendi tures and inves t -

ments, p o l l u t a n t s production, e tc .

The sys t en i s supposed t o be c losed a s f a r as Plu-tonium

i s concerned. Recycle of Plutonium i n thermal b reede r s

i s automat ical ly provided t o prevent accumulation of

Plutonium when t h e FBR's a r e a l ready int roduced.

2- Use of TOTEPL f o r t h e eva lua t ion of i n s t a l l a t i o n p o l i c i e s .

P a r t i c u l a r a t t e n t i o n was devoted t o -the e f f e c t of uncer-

t a i n t i e s about t h e phys ica l c h a r a c t e r i s t i c s of t h e

r e a c t o r s and of t h e i r f u e l cyc le (de l ays ) .

The e f f e c t of c o n s t r a i n t s on Uranium de l ive ry r a t e and

on t h e reprocess ing capac i ty were a l s o s tudied.

3- Completion of t he programme SITUS f o r t he op t imiza t ion

of t h e l o c a t i o n of power s t a t i o n s , given t h e geographical

d i s t r i b u t i o n and load diagram of t h e power denand, t h e

poss ib l e pa ths of t h e t ransmiss ion l i n e s and va r ious

eco log ica l c o n s t r a i n t s .

NEACRP - L - 155 f

DRAFT - June 1976

REACTOR PHYSICS ACTIVITIES IN FRANCE

JUNE 1975 - MAY 1976

lgth N. E. A . C. R. P. MEETING

JUNE 21St - 2sth CHALK-RIVER

I - GENERAL

After the previous decisions concerning the orders of

1200 MWe nuclear plants in FRANCE for 1976 and 1977, the 1978

programme will deal with 5000 MWe. It is now planned to have

1300 MWe PWR power unit in the future for some plants. Both

French-Belgium PWR plants on operation now, CHOOZ ( 2 300 MWe) and

TIHANGE (1 870 MWe) on industrial operation at full power since

september 1975, are working satisfactorily.Critlcallty for the

first French PWR FESSENHEIM 1 ( = 900 MWe) is expected before the

end of this year, it would be followed about two months later with

FESSENHEIM 2 ( - 900 MWe) and in 1977 with BUGEY 2 ( - 900 NWe) .

Outside the larger size plants for electricity supply,

studies are continuing on average power plants to produce both

steam and electricity. The demonstration plant located at Cadarache

and using integrated boiler, named CAP (Chaufferie Avanc6e Prototype),

went critical on 24 '~ovember 1975 and is now operating at powr-.

Concerning fast breeder reactors., PHENIX rounded the

cape of 3 billions Kwh May 12th 1976 without problem The pl-ant

was rccently operating at 294 MWTh and 264 MWe that is a net

efficiency of 44.5 % : its operates now normally over the nominal

power (250 MWe). From the first commercial operation date (14/7/74),

the global load factor is about 70 %. Control rod balancing techni-

que is used to reduce loca.1 hot spots and increase the power level.

Some pins reached a maximal bum-up rate of 60000 MVJD/T or - 7 % of

the fissile atoms burnt. A lot of specific subassemblies for fuel

irradiation are now loaded in the.core.

The decision concerning the construction of SUPER-PHENIX

(1200 MWe), to be built at CREYS-MALVILLE on the RHONE river, Should

be taken this year at the trinational.lev&], Preliminary design work

starts on the future plant named SAONE and made of two unikin the

power range 1200 - 1800 MWe each. The heterogeneous concept (interpal fertile blanket zones in the core), leading to a doubling time

in the range 11 years, is considered as a possible reference confi-

guration for this future plant. A corresponding important R & D pro-

gramme supports this fast breeder planned development. Performed up

to now in cooperation with ITALY, this programme will now be conrdo-

nated with the German R & D program after the agreements reached on

May 18th 1976 between C. E. A. and NOVATOME on the French side, and G. F. K. and INTERATOM on the German side.

It; must, be emphasized that large modifications appear

in FRANCE during the considered period in the nuclear R & D and

industrial organisation.

First of all, on light water reactors, C. E. A . became

shareholder (30 % ) of

truction. Furthermore

the FRAllATOllC Company in charge of PWK c0r.s-

a cbmmon R & D program on PWR reactors will

be d e f i n e d between EDF, FRAMATOEtE, CEA and WESTINGHOUSE, t h e l a s t

t h r e e p a r t n e r s f i n a n c i n a t h e s e s t u d i e s on an e q u a l b a s i s .

Second, f o r f a s t r e a c t o r s , a j o i n t company, named NOVATWE,

was c r e a t e d on A p r i l 1 9 7 6 by CEA ( 3 0 % ) , CREUSOT-LOIIE ( 4 0 % ) and

ALSTIIOM ( 3 0 % ) f i r m s NOVATO?.E h a s i n c h a r a e advanced r e a c t o r d e s i ~ n

and c o n s t r u z t i o n , t h a t means f i r s t f a s t b r e e d e r r e a c t o r s b u t a l s o

l a t e r on HTR r e a c t o r s . The R&D program on f a s t b r e e d e r s w i l l

c o n t i n u e t o be i n cha roe of C E A , i n c l o s e c o o ~ e r a t i o n w i t h EDF

and NOVATOME.

F i n a l l y , a majoy e f f o r t i s s t i l l c o n t i n u i n ? i n FRANCE

on a l l t h e problems of f u e l c v c l e i n and o u t of v i l e bo th

f o r f a s t and l i o h t w a t e r r e a c t o r s . Concerninn f u e l r e p r o c e s s i n o ,

f o r example, t h e n inh A c t i v i t y Oxide (HAO) p l a n t a t LA HAGUE c e n t r e

s t a r t e d o p e r a t i o n on i r r a d i a t e d l i g h t water f u e l s on May 1 6 , 1 9 7 6

and works s a t i s f a c t o r i l y up t o now f o r t h a t f i r s t camoainn.

I1 - FAST REACTOR PHYSICS

The main a s p e c t s o f t h e f a s t r e a c t o r p h y s i c s oroCram i n

FRANCE t h i s y e a r concerned r e s u l t s q e t from PHENIX o a e r a t i o n

and i r r a d i a t e d f u e l s a n a l y s e s , s t u d i e s on t h e o p t i m i z a t i o n o f t h e

d o u b l i n g t i m e , t h e c o n t r o l r o d i n t e r a c t i o n and i t s consequences f o r

power d i s t r i b u t i o n , n e u t r o n i c s a f e t y problems, r e s u l t s c o n c e r n i n g

t h e f u e l c y c l e i n and o u t o f ~ i l e and f i n a l l y s h i e l d i n v s t u d i e s .

II/1. PHENIX

The p l a n t is o p e r a t i n n on 3. noorninal b a s i s nov. Due t h e

remarkable load f a c t o r , t h e p h y s i c s nrorfram based on i r r a d i a t e d

f u e l a n a l y s e s c a n be followc?? up a s n lanned . The v a r i a t i o n o f

s t a n d a r d f u e l a tomic compos i t ions v e r s u s burn-up f o r c o r e and

b l a n k e t s were measured u p . t o t h e end o f t h e 6 t h ' , c y ~ l e (N 5 4 0 k 0 ~ ~ b K ? / ~ )

These r e s u l t s a l lowed t o check t h e per formances o f t h e burn-u?

codes a n d t h e b r e e d i n g n a i n c a l c u l a t c r ' v a l u e s . F r o m t h e exnerirc!nt ; l l

r e s u l t s a v a i l a b l e up t o now, t h e c a l c u l a t i o n - e x p e r i m e n t comnar:.:;on

l e d t o a r e a l l y good aqreement . The program is s t i l l noing on.

The f u e l manaaement code CAPHE, used f o r t h e n r e d i c t i o n

o f t h e v a r i o u s l o a d i n n s c o n t i n u e s t o he w e l l adanted t o t h e

o p e r a t i o n r e q u e s t s up t o t h e g t h c y c l e even f o r d i f f i c u l t s i t u a -

t i o n s : c o n t r o l b a l a n c i n g f o r example. Very s m a l l d i s c r e n a n c i e s

(<0.1% A K ) were observed f o r t h e f i r s t c r i t i c a l i t v o f each c y c l e .

There was no more problem f o r c o n t r o l rod r e a c t i v i t y wor th o r

r e a c t i v i t y l o s s p e r c y c l e . The new r e s u l t s conf i rmed t h e under- +'

p r e d i c t i o n by c a l c u l a t i o n s of t h i s l a s t pa rame te r by 10 - 10%.

Power and t e m p e r a t u r e c o e f f i c i e n t s and c o n t r o l r o d wor ths a r e

measured s y s t e m a t i c a l l y a t v a r i o u s c y c l e s .

G l o b a l l y , f o r n e u t r o n i c and s h i e l d i n g problems, t h e

c u r r e n t c a l c u l a t i o n s methods seem w e l l adan ted f o r t h e PMENIX

power l e v e l p l a n t s . The most i m p o r t a n t r e s u l t s w a i t e d f o r i n

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

I11 - CRITICAL FACILITIES

The program was devo ted e s s e n t i a l l y d u r i n n t h i s p e r i o d

t o two major p o i n t s :

s t u d i e s of t h e n e u t r o n i c problems o f t h e ?ewer r e a c t o r

PEC f o r C N E N

commercial f a s t b r e e d e r problems s t u d i e d i n t h e s a m

assembly.

A ) The PZCORE program c a r r i e d o u t on WLSIIRCA s i n c e

May 1975 u n d e r c o n t r a c t w i t h CNEN aims a t measur ino t h e n e u t r o n i c

pa rame te r s of t h e i . r r ad ia t . i on f a s t I t a l i a n r e a c t o r PIX, mainly t h e

. . ./. . .

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

and t h e problem o f t h e l o o p l o c a t e d a t t h e c e n t r e o f t h e c o r e .

- Main i n t e r e s t s concerned r e a c t i v i t y e f f e c t s and power d i s t r i b u -

t i o n p e r t u r b a t i o n s . T h i s program was d i v i d e d i n two p h a s e s .

-. . I n t h e phase PECO?E I , t h e c o r e c o n t a i n e d a c e n t r a l

zone r e p r e s e n t a t i v e of t h e PEC c o r e (Pun2-UO -Ma) and a n o u t e r 2 . d r i v e r zone i n Uranium.. Core h e i ~ h t v a s ~ 6 0 c m . P1.1 t h e bs.sic

p a r a m e t e r s o f t h e c e n t r a l zone were s t u d i e d : a a t e r i a l b u c k l i n g ,

r e a c t i o n r a t e r a t i o s , power d i s t r i b u t i o n s . The c a l c u l a t i o n

expe r imen t compar ison on c r i t i c a l mass showed an a b n o r r a l d i s - + c repancy ( E - C = -0.9 - 0 . 1 4 3 A R ) compared t o t h e u s u a l

r e s u l t s o b t a i n e d w i t h t h e CARXP.VAL I11 s y s t e n (on a v e r a c e 2 0 . 2 % )

I t was demont ra t ed t h a t t h i s d i s c r e p a n c y was l a r q e l y d u e t o

t h e c a l c u l a t i o n o f t h e r a d i l l s t a i n l e s s s t ee l - sod ium r e f l e c t o r .

S e v e r a l c o n f i q u r a t i o n s o f c e n t r a l - c o n t r o l r o d were s t u d i e d i n

t h e c e n t r e of t h e c o r e . These c o n f i ~ u r a t i o n s s i m u l a t e ? t h e PEC

a b s o r b e r (B4C) and. f o l l o w e r c o n t r o l r o d s . I t was noted t h a t t h e

C/E d i f f e r e d s i g n i f i c a n t l y f rom t h e r e s u l t s o b t a i n e ? p r e v i o u s l y

on t h e same rod c o n f i q u r a t i o n i n t h e v a r i o u s c o r e s of t h e

MASURCA RZ proqram. The r e a s o n s f o r t h e s e d e v i a t i o n s a r e now

a b e e i n g a n a l y s e d .

Three c o n f i a u r a t i o n s of t h e PZC i r r a d i a t i o n l o o p were

s i m u l a t e d i n t h e c e n t r e of t h e PECQQE 1 c o r e . f i s s i l e c e n t r a l

p a r t o f t h e l o o p and t h e d i l u a n t p a r t were r e p r e s e n t e d . C a l c u l a -

t e d r e a c t i v i t y e f f e c t o f t h i s l o o p were u n d e r p r e d i c t e d by 6 %

w i t h CARNAVAL I11 c a l c u l a t i o n s . C a l c u l a t e d a x i a l power d i s t r i - - b u t i o n s showed a d i s c r e p a n c y o+ 33 a t t h e c o r e b l a n k e t i n t e r f a c e .

R a d i a l power d i s t r i b u t i o n s i n t h e f i s s i l e p a r t were u n d e r p r e d i c t e d

by 43. P r o t o n r e c o i l neu t ron s?ec t rum measureven t s and vamw

h e a t i n g measurements by TLD and i o n i z a t i o n chambers were a l s o

performed i n t h i s l o o p zone . '

. I n t h e phase PECOIE 11, t h e o u t e r zone on t h e

c o r e - r e f l e c t o r i n t e r f a c e was b u i l t w i t h t h e Pu f u e l and t h e

Uranium f u e l used i n t h e c e n t r e . The c o n t r o l s y s t e v o f t h e PEC

r e a c t o r was s i m u l a t e d a t t h e c o r e - s t a i n l e s s Na r e f l e c t o r i n t e r f a c e

w i t h twelve s i m i l a r r o d s l o c a t e d on two r i n n s . The f o l l o w i n n

c o n f i q u r a t i o n s were s t u d i e d :

- end o f l i f e : 1 2 s n d i u s h o l e s

- one a b s o r b e r c o n t r o l rod i n v a r i o u s p o s i t i o n s

- i n t e r a c t i o n o f two a b s o r b e r r o d s i n v a r i o u s p o s i t i o n s

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

Main p a r a m e t e r s s t u d i e d were r e a c t i v i t y wor th u s i n n t h e c o r e - r e f l e c -

t o r b a l a n c i n g t e c h n i q u e and power d i s t r i b u t i o n s p e r t u r b a t i o n s hy

f o i l s and d e t e c t o r s .

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

sodium r a d i a l r e f l e c t o r by a Nickel-sodium r e f l e c t o r was a l s o

s t u d i e d on a f i f t h o f t h e r e f l e c t o r f o r r e a c t i v i t y , c ' on t ro l r o d

worth and spec t rum e f f e c t on t h e r o d . The (rain due t o t h e n i c k e l

r e f l e c t o r appeared less i m p o r t a n t t h a n c a l c u l a t e d by CAP'IPTTAL 111.

F i n a l l y s u b c r i t i c a l i t y peasurements of c o n t r o l rod

wor ths were veasu red f o r t h e f u l l 1 2 rod PEC c o n t r o l sys tem by

v a r i o u s t e c h n i q u e s .

9) F o r f u e l c y c l e i n p i l e a n a l y s e s , t h e h i g h e r e l u t o n i u m

i s o t o p e program was a l s o c a r r i e d o u t i n t h i s c o r e wi.th t h e t h r e e

mixed o x i d e f u e l s w i t h v a r i o u s i s o t o n i c c o c p o s i t i o n s ( ? u ? 4 0 8 % i 4 5 5 ,

11%). That comple tes t h e p r e v i o u s r e s u l t s o f t h i s program f o r t h e

hard s p e c t r a .

F i n a l l y t h e PRE FACIME p r o o r a s a i ~ i n n a t r n e a s u ~ i n ?

t h e e l e m e n t a r y phenomena c o n c e r n i n ? t h e he t e roneneous concep t

( i n t e r n a l f e r t i l e zone) , t h e c o n t r o l rod i n t e r a c t i o n p r o b l e v s znd

t h e s a f e t y problems (Na v o i d ) was s t a r t e d on ?flASV9C~' ' n %v.

F o r f a s t r e a c t o r s , t h e proqram d k a l t main ly t h i s y e a r

w i t h f i s s i o n p r o d u c t s .

R e a c t i v i t y worth r e a s u r e c e n t s by t h e o s c i l l a t i o n

t e c h n i q u e of f u e l s o r samples i r r a d i a t e d i n PHENIX and RAPSODIE

FORTISSIMO t o h i g h bu rn up were p e r f o r ~ e d i n t h r e e c o r e s h u i l t

i n ERMINE. Two t y p e s of i r r a d i a t e d f u e l s were used :

- p u r e 2 3 5 ~ r a n i u r n i r r a d i a t e d i n 9PPSnqLE up t o

100000 . 3 / T

- s t a n d a r d mixed o x i d e PBSNIX c o r e 1 f u e l i r r a d i a t e d

up t o 27000 MT;73/T

These samples w e r e o s c i l l a t e d i n one Uraniuw oxide-

sodiux? c o r e and two mixed-oxide c o r e s t y p i c a l of t h e two zones

o f c l a s s i c a l comnerc i a l p l a n t s . Non i r r a d i a t e d f u e l s of t h e same

compos i t ions and systematic s t a n d a r d samples w i t h v a r i o u s i s o t o p i c

c o m p o s i t i o n s were a l s o measured t o c o r r e c t t h e r e a c t i v i t y worth

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

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

@ ,

by mass spec t romet ry ' . F i n a l l y , a n accu racy of a b o u t -7% on t h e

g l o b a l FP e f f e c t on r e a c t i v i t y i s expec ted f o r commercial n l a n t s . .

Fu r the rmore some s e p a r a t e d f i s s i o n p r o d u c t measurements

were a l s o performed on t h e s e c o r e s u s i n q e i t h e r o s c i l l a t i o n rnethod

o r a c t i v a t i o n t e c h n i q u e s .

The HARMONIE program d u r i n n t h i s p e r i o d d e a l t w i t h

n e u t r o n p r o p a a a t i o n s t u d i e s f o r f a s t r e a c t o r s i n p u r e sodium and

p u r e i r o n m d i a . T h i s proaram i s ?erforx?ed i n cooper? . t ion w i t h

CNEN and c o o r d o n a t e d w i t h t h e T P p ~ r f l p royrap .

F o r t h e bla rr.edium 3 x 3 x 3 m; t h r e e t y p e s o f e n t r a n c e

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

s t e e l r e f l e c t o r s (11 and 37 cn t h i c k ) and one c n n f i n u r a t i o n w i t h

a U02 d e p l e t e d - s o d i u r b l a n k e t 25 C F t h i c k . The neu t ron svec t rum

a t t h e e n t r a n c e o f t h e Ya r red iumms measured e i t h e r by p r o t o n

r e c o i l o r hy f o i l s and d e t e c t o r s . The r e a c t i o n r a t e r a d i a l and

a x i a l d i s t r i b u t i o n s i n t h e N a mec?iuin v e r e measure? hy c l a s s i c a l .

t e c h n i q u e s : f i s s i o n c h a h e r s , a c t i v a t i o n f o i l s , d e t e c t o r s ,

p r o t o n r e c o i l d e t e c t o r , . . .

Same t y p e s of experil-nents were per+ormed on t h e p u r e

i r o n medium 1.8 m t h i c k t o be inc luded i n t h e i r o n benchpark

compar ison . F u r t h e r p o r e s y s t e p a t i c FarnTa h e a t i n n measurements

a x i a l l y and r a d i a l l y were d e t e r p i n e d e i t h e r w i t h TLn o r w i t h

i o n i z a t i o n chambers.

I I I / 4 . IRRADIATED FUELS

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

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

were o b t a i n e d on t h e s p e c i f i c p u r e s a r p l e s cf plu tonium, uranium,

@erici i im and s e p a r a t e d f i s s i o n p r o d u c t s i r r a d i a t e d i n t h e c e n t r e

o f PHENIX up t o 27000 .T'D/T (PROFIT, e x p e r i ~ e ' n t ) . The c o ~ p a r i s o n

c a l c u l a t i o n - e x p e r i m e n t on t h e c a p t u r e r a t e r a t i o s of t h e main

f i s s i l e and f e r t i l e i s o t o p e s of P lu toniur" and Uranium now avails- +

b l e conf i rmed t h e v a l i d i t y o f t h e C>.RYI\VFL I11 s y s t e r t o -3%

e s ? e c i a l i y f o r 2 3 9 ~ u , 2 4 0 ~ u , 2 3 8 ~ , 2 3 5 ~ . Tha t c o n f i r m s t h a t t h e

c a l c u l a t e d P H E N I X i n t e r n a l b r e e d i n ? ~ a i n seems c o r r e c t l y p r e d i c t e d

More r e r ;~ : l t s a r e wa i t ed f o r i n t h e n e x t f u t u r e on heavy

i s o t o p e c a p t u r e s ( 2 4 1 ~ u , 242pu , Am, Cm. ... ) . F i s s i o n r a t e ~ e a s u r e -

ments a r e now be ing performed by a n a l y z i n g Nd c o n t e n t i n t h e

samples .

I V - TIICORITICAL AND ANALYT1CP.L VlOlir(

. Among t h e v a r i o u s a c t i v i t i e s o r i e n t a t e d towards cormer-

c i a 1 p l a n t s , only t h e most i m p o r t a n t p o i n t s w i l l be br ie f1 .y m n t i o n -

ned :

IV/1. Fo r c o n t r o l r o d s , t h e main s t u d i e s concerned

- r o d i n t e r a c t i o n and i n f l u e n c e on power d i s t r i b u t i o n s

- f e r t i l e U02-Na c e n t r a l zone c a l c u l a t i o n s

- measurement t e c h n i q u e s f o r c o n t r o l r o d on povrer r e a c -

t o r s .

IV/2. F s t r o n g 'ff o r t was p l a c e d on *.la v o i d a n a l y s i s

i n v a r i o u s c o n f i q u r a t i o n s : c o r e , b l a n k e t s , r o d s , i n f l u e n c e s of:

FP and i s o t o p i c Pu compos i t ions . A 1 1 t h e p a s t e x p e r i n e n t s on t h i s

t o p i c s are now be ing r e a n a l y z e d .

IV/3. The major work concerned t h e p r e p a r a t i o n o f

t h e V e r s i o n I V o f t h e CARNF.VAL sys t em t h a t h a s t o be o p e r a t i o n n a l

a t t h e end o f 1976 f o r SUPER-PHEXIX enr i chmen t d e f i n i t i o n .

The main improvements w a i t e d f o r i n t h i s V e r s i o n TV

c o n c e r n s :

a) S t r u c t u r a l materials : I r o n , C r , Eli, "10, *"n.

N e w s e v a l u a i i o n s were i n t r o d u c e d i n t h e d a t a set

S e l f s h i e l d i n p f a c t o r s were e v a l u a t e d f o r T ron , - N i and C r .

E v a l u a t e d c r o s s s e c t i o n s were a d j u s t e d +or I r o n , 2

C r and N i on t h e a v a i l a b l e i n t e q r a l expe r imen t r e s u l t s ( L o , B,,

r e a c t i v i t y wor th ) .

b ) Hiehe r p lu ton ium i s o t o p e s : T h i s i s one o f t h e

major improvement.

A f t e r t h e comple t ion .of h i g h e r plutonium i s o t o p e provram, c a p t u r e ,

f i s s i o n and v paramete r s were a c ? j u s t e d on i n t e n r a l measurements

of f o u r t y p e s :

- m a t e r i a l buckl5ng and K m

- f i s s i o n rate r a t io s

- r e a c t i v i t y \worth

- i r r a d i a t i o n on PHExJIX and RAPSODIE ( c a p t u r e ) .

No more problem is expec ted now f o r 2 3 9 ~ u , 24OPu, 2 4 1 1 ? ~ and 2 4 2 ~ u

d a t a .

C ) F i s s i o n p rpduc t s

E new e v a l u a t e d l i b r a r y o f 180 s e p a r a t e d F? i so towes

i s now q e n e r a t e d i n c o o p e r a t i o n w i t h CWEV.

U p t o now t h e 22 most i w p o r t a n t i s o t o p e s a r e a v a i l a b l e .

The e v a l u a t e d d a t a were a d j u s t e d on i n t e q r a l e x p e r i r e n t s on sepa -

r a t e d f i s s i o n p r o d u c t s per formed i n v a r i o u s r e a c t o r s . The c u r r e n t

c o n c l u s i o r l e a d s t o d e c r e a s e t h e p r e v i o u s FP e f f e c t i n Ver s ion TIT

by - 15% f o r t h e a d j u s t e d set .

The f i n a l answer w i l l b e o b t a i n e d when t h e r e s u l t s of t h e a loha l .

FP measurements on ERNINE w i l l b e a v a i l a b l e ( s e p t . 1 9 7 6 ) .

S e v e r a l c t h e r im?rovements t h a t w i l l a l s o be i n c l u d e d

i n t h i s Ver s ion I V were o b t a i n e d o r a r e be in? n repa re i i POW :

e v a l u a t i o n o f t r a n s a c t i n i d e s d a t a : a s t r o n a emphasis

i s p l a c e d on t h i s problem f o r f u e l c y c l e r e q l l e s t s . NP, Cm, 4m

i s o t o p e s were s t u d i e d .

new d a t a f o r Na, Oxygen, Th232, 233 I!, 233pa

new approx ima t ions f o r c a l c u l a t i o n s o f b l a n k e t c r o s s

s e c t i o n s .

F i n a l l y , one o f t h e most i m p o r t a n t improvements f o r

V e r s i o n I V i s p r o b a b l y t h e number of nzw i n t e n r a l exper imenta l .

r e s u l t s a v a i l a b l e from 1973, t h a t l e a d t o a h iqh d e g r e e of

c o n f i d e n c e i n t h e performances of t h i s f u t u r e syste!?.

I V / 4 . The P R O P A V E Ver s ion O sys te r" , a iminq a t c a l c u l a -

t i n g n e u t r o n a t t e n u a t i o n i n s t a i n l e s s s t ee l - sod ium s h i e l d i n ~ s was

used t o a n a l y s e t h e H A R H O V I E expe r imen t s on media w i t h v a r i a h l e

compos i t ion between 0% u? t o 100% sodium. The t r a n s p o s i t i o n o f t h e

r e s u l t s t o SUPER-?HE?I IY c o n f i g u r a t i o n i s now noin? on and

a d j u s t e m e n t o f t h e PROPANE eva l -ua ted c r o s s s e c t i o n s on t h e

r e s u l t s of t h e s e HI?R?,'OnJIE e x e e r i r e n t s i s expec ted f o r t h e end

o f 1 9 7 6 .

- 80 - 111 - LIGHT WATER REACTOR PHYSICS -

The reactor physics activities related to PWR concerned

three main topics :

- the calculation methods. A complete set of codes the NEPTUNE system, is being developped.

- Experimental studies on critical facilities and burnt fuels which are performed in order to tesf the calculation methods and

data.

- Theoretical and experimental works'related to the design or operation of some specific reactors (SENA, CAP, FESSENHEIM, ... ) .

111-1 : Calculation methods

The NEPTUNE system of codes has been developped for complete

calculations of light water reactors. It includes the APOLLO code

for cell calculations and it allows us to perform diffusion, burn up

and fuel management calculations.

The diffusion calculations with two or three dimensions are

realized by a finite elementmethod which permits.to obtain quickly accurate results.

In this NEPTUNE system we can also make one or.two dimen-

sions calculations by the finite difference method when we want very

complete power distributions.

The thermohydraulic code FLICAiscoupled with the three

dimension diffusion calculation of NEPTUNE in such a way that we have

the possibility to take inaccouritcorrect1y the Doppler and water

density effects.

The works presently in progress include a new treatment

of the multicell calculations with a two dimensions transport appro-

ximation, the development of a three dimension calculation of power

distribution by a constructive method, and the study of a coupled

calculation of thermohydraulic and neutrop-kinetics problems ...

In the field of dataan evaluation of the fission and

capture cross sections of 2 4 2 m ~ m has been realized.

- 81 - .For the qualification of methods and data, in addition

to the use of exp6rimental results obtained in our present programs

on MINERVE, AZUR and by the following of start up and operation of

power reactors, an analysis of homogeneous critical experiments rea-

lized in the States and in France has been performed.The main conclu-

sions of this analysis are : 235u - a check of the previously noticed tendencies for

- a test of the 2 3 3 ~ data - a new validation of the 2 3 9 ~ ~ data evaluated by P.RIBON.

These experiments were not sufficient to get a new test of the

2 4 0 ~ ~ data.

An other topic of present studies is the equivalence of

transport and diffusion calculations. A systematic investigation

of the change to be put on the data used for multigroup calculations

with diffusiontheoryisbeing done in order to keep, as well as

possible, the same reaction rate than in a true transport calculation

for a reference configuration.

Furthermore a theoretical study of the equivalence in a

monocinetic theory 1s being carried on.

111-2 : Experimental programs

Experiments related with light water reactor physics have

been carried out on the both f a c i l i t i e s ~ ~ ~ ~ and MINERVE. Some measu-

rements on irradiated fuels have been also completed with a particular

emphasisan gamma-spectroscopy on fuel subassemblies and transactinium

isotopes measurements.

111-2-1 : MINERVE experiments

Light water reactor experiments are performed in the MINERVE

reactor with the "MELODIE" arrangment, that is a two zone coupled core

- the test zone simulates a fraction of a power reactor core. In this one the fuel is composed of oxide pins with zirconium or stainless

steel cladding : this pins are loadedbetween two grids with an

uniform pitch. The number of pinsin the test zone is normally

of the order of eight hundred.

- The driver zone is loaded with an enriched uranium plate fuel, the normal fuel of MINERVE. The control rods are in this zone which is

surrounded by a graphite reflector.

- 8 2 - *

This particular arrangment, very similar to the ERMINE one

for fast reactor experiments, allows us to perform some measuxcments

in various test zoncs with a minimum quantity of fuel and keeping

always the same control devices for the reactor itself.

During this last year the experiments concerned several

configurations of the test zone loaded with a 17 x 17 type PWR fuel;

the most important part of the programme was devoted to power distri-

butions measurements in lattices perturbed with water holes and

poison rods.

A complete simulation of the loading of subassemblies of

the FESSENHEIM I reactor (900 MWe PWR) has been realized and a compa-

rison between y spectrometry measurements on the fuel and fission

chambercounting has been performed in order to test the analysis of

the in core instrumentation.

111-2-2 : AZUR experiments

During this year the main part of the AZUR program is

related to the CAP reactor, that is a small PWR type reactor located

in CADARACHE and which started up in Nov. 1975. This reactor has

been built up for fuel element and component tests. One core of this

CAP reactor will be loaded with a 17 x 17 type fuel and thepurpose

of the AZUR experiments was to check the design calculations of this

core which will contain several kind of poisons (Boron, 'gadolinium

and ha•’ nium) . The measurements in AZUR has been done on the actual. fuel

of the CAP reactor in three steps, respectively with 4, 9 and 16

subassemblies. In the first campaign (4 elements) a particular empha-

sis was put on thetest of cell and multicell calculations, including

relative power measurements through the lattice, conversion factor,

reactivity balance and temperature coefficient between 20 and 90•‹C.

In the other configurations are performed a complete check of some

subassembly loading and axial measurements taking in account the

important height of the fuel (1,8O m) . 111-2-3 : Irradiated fuel measurements

The gamma spectrometry of subassemblies in the storage

pool of the reactor after the shut down for reloading remained one

of our major topics. A ne! campaign has been realized durin~ the

autumn in the SENA rcactor. (cnd of the 5th cycle). Thci expcrirnental

- 83 - apparatus was very siroilsr to thc one used in 1 9 7 4 except for sonic

modifications concerning mainly the Ge-Li detector. About ten fuel

elements were measured.

A code has been written in order to analyse the experi-

mental results, giving the mean power or burn-upin several zones of

the subassemblies. A comparison of the results-with thoseof>in~'core

instrumentation and with 3D calculations using the NEPTUNE system

is being completed.

Inthe field of destructive analysis a new technic based

upon the a spectrometry has been developped for the 2 3 2 ~ content

measurements. The test of the calculation of this isotope is impor-

tant for the prediction of the uranium composition after reprocessing

of the light water reactor fuel if this uranium must be recycled

with or without a reenrichment process. The experimental values that

we obtained, of the order of compared to 2 3 8 ~ for a fuel irra-

diated at 30000 MWd/T and a cooling time more than one year, are

higher than the.calculated values.

LV - HTGR .REACTOR PHYSICS : . ' The study on the highly enriched uranium thorium fuel cycle

with inkegrated block representing the GA design for 1160 MWe reactor

has been going on with the MARIUS IV program.

MARIUS IV is a critical experiment where the central zone is a cylinder (diameter : 1.40 m and 1.40 m long). This imperturbed

zone is surrounded by the buffer zone and the driver core.

At the end of 1975 , the kw was'measured by the progressive

poisonning method.

' The aim of the present program covers the following items :

- buckling measurements,fission rate distributions (foils and fission chambers)

- core reflector interaction (full and partially empty reflector) 0

- control, rod effects (central and excentric rod) 0) 0 - interaction between a fresh and an irradiated bloc (in this case a

the uranium 235 content in the particles is lower). CV 0 sb-,

The experimental results of this MARIUS IV program are a3

used to qualify GA and CCA calculation methods.

- 84 - V - REACTOR PHYSICS FOR SAFETY PROGRAM

The studies about the construction of the reactors PHEBUS,

CABRI and SCARABEE which are designed for safety experimental programs are

carried on at Cadarache.

Tworunsof measurements were performed during this year

in the EOLE liquid-moderated critical facility.~h~ first one was in the

framework of the PHEBUS project (PWR safety);this PHEBE experiment

gave us :

- critical shape to test the validity of the keff evaluation - power distribution to determinate the energy released in the loop during the loss of coolant.

- efficiency of the hafnium control rods (one rod, two, four, six) to evaluate the shadow effect and test the calculation model (trans-

port S4 method and neutron cross section collapsing).

- gamma heating with FLi detectors.

The goal of the second one was to give the physical values

to qualify the calculations of the SCARABEE project. The aim of the

new SCARABEE experiments is to have a fast neutron spectrum in the

center of the loop in order to simulate the critical conditions of

the LMBFR (power gradient in the test pins).

The fast reactor conditions are obtained with a cadmium

filter and an enriched uranium oxide converter.

Allthese experiments are carried out in parallel with the

design calculations and the interpretation of the previous experi-

ments. The ,starting up of CABRI and PHEBUS are prepared.

/l/ k.CORCUERA - ?.!ethodes t h e o r i q u e s pour l e c a l c u l n e u t r o n i q u e

d e s sys t6mes c o e u r - c o u v e r t u r e e t c o e u r - r e f l e c -

t e u r d e s r e a c t e u r s A n e u t r o n s r a p i d e s .

Rapport CEA-3-4726

/2/ P . ROUZAUD - Nouveaux a l g o r i t h m e s d e r i v e s d e l a ~ . e t h o d e

d e s y n t h s s e . P p p l i c a t i o n aux p roh lPnes d e

d i f f u s i o n .

Thsse d e D o c t o r a t d e s?Ecial i te >rent ion ' lathe-

ma t iques - U n i v e r s i t e d e PVYTNCE

/3 / F.BOUTEAU e t a 1 - Exper i ences i n t e g r a l e s d e g r o p a ~ a t i o n d e

n e u t r o n s d a n s les m i l i e u x ac ie r -sodiur r .

Reunion d e s p E c i a l i s t e s d e s e t u d e s d e s e n s i -

b i l i t e e t d e s e x p e r i e n c e s r e p s r e s s u r l e s

p r o t e c t i o n s - P a r i s 7-10 o c t o b r e 1975

J4/ F.ROUTEAU e t a1 - F o r m u l a i r e d e p r o p a p a t i o n d e n e u t r o n s d a n s

les m i l i e u x ac ie r -sodium pour l e s p r o t e c t i o n s

d e l a f i l i P r e r a p i d e . Ibidem P a r i s 1975.

t

/5/ J.Y.3ARR.E, J.ROUCHAP.D - 1m.portance d e s donnees n u c l e a i r e s d e s

i s o t o p e s t r a n s a c t i n i u r n s pour l a nhys ique d e s

r e a c t e u r s r a p i d e s e t t h e r r i q u e s .

R&uniom d e s p e c i a l i s t e s s u r les donnees

n u c l e a i r e s des i s o t o p e s t r a n s a c t i n i u r s , KARLSRUHE 3-7 n o v e ~ b r e 19 75.

/ 6 / F.BOUTEAU e t a1 - S t u d i e s on neu t ron p r o n a n a t i o n i n s teel-

sodium media f o r f a s t r e a c t o r s h i e l a i n c .

>YS Vlinter rneetinn W u 1 FRFNCTSCn

1 6 - 2 1 November 1 9 7 5

/7 / U.BROCCOL1 e t a 1 - F t u d e s d e s h a r r e s d e commande d u r e a c t e u r

PEC d a n s 1' a s s e ~ b l a g e ?QSURCE/PECOPE.

R6union d e s p 6 c i a l i s t e s s u r les t e c h n i q u e s

d e mesures d e s h a r r e s d e corn-mande : a n t i r s a c -

t i v i t E s e t d i s t r i b u t i o n d e p u i s s a n c e ,

CADlrRACBE 21-22 A v r i l 1 9 7 6

/8/ S . T E L L I E 3 - S e c t i o n s e f f i c a c e s mul t in rnupes d e c a p t u r e

r a d i a t i v e e t d e f i s s i o n 242A,

Note CE>.-N-1821 ( 1 9 7 5 )

NEACRP-L-155

SPAIN

REACTOR PHYSICS ACTIVITIES AT JEN-SPAIN*

and

NEUTRONICS OF LASER FISSION-FUSION SYSTEMS

June 1975 - May 1976

P resen ted by G . VELARDE, DIREC - TOR, DEPARTMENT OF TECHNOLOGY

(*) S e l e c t i o n of works performed i n t h e D i v i s i o n s o f

Reac to r Theory and C a l c u l a t i o n s , Fus ion , and Ex-

p e r i m e n t a l Reac to r s o f t h e Department of Techno-

l ogy ( J u n t a d e Ene rg i a Nuclear , Madrid, S p a i n ) .

CONTENTS

I V . -

I n t r o d u c t i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

N o t e s o n Z o r i t a R e a c t o r R e a c t i v i t y F o l l o w ..........

11.1.- B r i e f D e s c r i p t i o n o f t h e Z o r i t a R e a c t o r Co r e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

11.2.- I n t r a n u c l e a r I n s t r u m e n t a t i o n . . . . . . . . . . . . . . . .

1 1 . 3 . - R e a c t i v i t y F o l l o w C o d e s . . . . . . . . . . . . . . . . . . . . .

1 1 . 4 . - R e a c t i v i t y F o l l o w R e s u l t s . . . . . . . . . . . . . . . . . . . R e f e r e n c e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

N e u t r o n i c a n d T h e r m o h y d r a u l i c A n a l y s i s o f a Swim

m i n g P o o l R e a c t o r D e s i g n e d a t JEN f o r t h e R e p u b l i c

...................................... o f C h i l e . . . . . R e f e r e n c e s ...

N e u t r o n i c s o f

I V . 1 . - O b j e c t

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . * - . . .

-. - L a s e r F i s s i o n - F u s i o n S y s t e m s .........

i v e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

I V . 2 . - N e u t r o n i c A n a l y s i s .......................... IV.3.- N e u t r o n i c C a l c u l a t i o n s .......................

I,

I V . 4 . - D s s c r i p t i o n o f C o d e s ........................

I V . 5 . - R e f e r e n c e s ..................................

I . I N T R O D U C T I O N -

A c t i v i t i e s on R e a c t o r P h y s i c s a t J E N , i n c l u d e d i n t h i s

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

r e a c t o r s and l a s e r d r i v e n f i s s i o n - f u s i o n s y s t e m s .

I n t h e t h e r m a l r e a c t o r s f i e l d , t h e t a s k s o f c o r e d e s i g n

f o r c o m m e r c i a l L W R ' s were a s s i g n e d t o ENUSA (Empresa N a c i o n a l

de U r a n i o S . A . ) , a s was m e n t i o n e d i n t h e f o r m e r r e p o r t . The

J E N c o o p e r a t e d w i t h ENUSA and some S p a n i s h u t i l i t i e s i n t a s k s

of t e c h n i c a l s u p p o r t and t r a i n i n g , t h a t c o u l d p r o c e e d e v e n t u a l l y

i n t h e f u t u r e . A t p r e s e n t , main J E N r e a c t o r p h y s i c s a c t i v i t i e s

r e l a t e d t o c o m m e r c i a l r e a c t o r s , a r e t h o s e o f t h e o r e t ~ ~ d l , methods

deve lopmen t and c a l c u l a t i o n a l s u p p o r t t o t h e N u c l e a r S a f e t y

Depar tmen t o f t h e JEN, f o r power r e a c t o r s i n o p e r a t i o n , c o n s -

t r u c t i o n o r p r o j e c t i n S p a i n , which a r e p e r f o r m e d by t h e D i v i -

s i o n of R e a c t o r Theory and C a l c u l a t i o n s . One o f t h e s e a c t i v i t i e s

h a s been t h e o p e r a t i o n f o l l o w o f t h e Z o r i t a 160 Mwe PWR, i n c l u d i n g

r e a c t i v i t y a n d power d e n s i t i e s and b u r n u p d i s t r i b u t i o n f o l l o w ,

which i s summarized i n c h a p t e r 11.

Swimming-pool r e a c t o r a n a l y s i s , which were begun w i t h

JEN-1 a n d JEN-2 r e a c t o r s , h a v e been c o m p l e t e d t o p r o v i d e

t h e o r e t i c a l , c a l c u l a t i o n a l and e x p e r i m e n t a l b a s i s t o Spain. ..

and o t h e r c o u n t r i e s . T h e -. J E N h a s d e s i g n e d a 1 0 ~ w t h swimming-

p o o l r e a c t o r u n d e r c o n s t r u c t i o n i n C h i l e . A summary o f t h e

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

s e n t e d i n c h a p t e r 111.

* . . ~.. . I n t h e f a s t r e a c t o r s f i e l d , t h e J E N h a s p r o c e e d some

a c t i v i t i e s d i r e c t e d t o e x t e n d and improve t h e c a l c u l a t i n g c a p a -

b i l i t y and i t s a p p l i c a t i o n t o d i f f e r e n t a s p e c t s of f a s t r e a c t o r .A

c o r e d e s i g n . I n n e u t r o n i c s , t h e p r o c e s s i n g o f c r o s s s e c t i o n s

f rom ENDF/B-4 was p e r f o r m e d , u s i n g t h e code s y s t e m E T O X - 1 D X

and RESEND. A code s y s t e m f o r c a l c u l a t i o n o f e f f e c t i v e c r o s s

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

f i n e s t r u c t u r e i n m i x t u r e s o f d i f f e r e n t n u c l i d e s , h e t e r o g e n e i t y

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

ETOX-1DX h a s b e e n m o d i f i e d t o t h e t r e a t m e n t o f a n i s o t r o p i c

e l a s t i c s c a t t e r i n g . An a p p l i c a t i o n o f v a r i a t i o n a l t h e o r y f o r

b i l i n e a r w e i g h t i n g o f c r o s s s e c t i o n s i s b e i n g d e v e l o p e d .

The a p p l i c a t i o n o f t h e c o d e s y s t e m ETOX-1DX-CITATION t o c a l c u -

l a t i o n s o f b u r n u p t h r o u g h c o n s e c u t i v e c y c l e s , p o w e r d i s t r i b u -

t i o n s , c o n t r o l r o d e f f e c t s , D o p p l e r a n d Sod ium v o i d i n g e f f e c t s ,

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

s h i e l d i n g , t h e F r e n c h M o n t e c a r l o c o d e TRIPOLI i s b e i n g i m p l e -

m e n t e d . I n t h e r m o h y d r a u l i c s t h e c o d e SWELL h a s b e e n a p p l i e d

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

t o r d e s i g n s , a n d t h e i m p l e m e n t a t i o n o f GAPCON a n d FORE-2 p r o -

grammes was s t a r t e d .

I n t h e f i e l d o f e x p e r i m e n t a l f a s t r e a c t o r p h y s i c s , a n

e x p e r i m e n t t o p u l s e t h e CORAL-? r e a c t o r , w i t h a n i n t e n s e n e u t r o n

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

f o r n o n - s t e a d y p r o b l e m s .

I n t h e f i e l d o f L a s e r d r i v e n f i s s i o n - f u s i o n s y s t e m s ,

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

A d i s c u s s i o n o f t h i s work i s p r e s e n t e d i n c h a p t e r I V .

11. NOTES ON ZORITA REACTOR REACTIVITY FOLLOW -

11.1. BRIEF DESCRIPTION OF THE ZORITA REACTOR CORE.

The reactor core of "JOSE CABRERA" Nuclear Plant is made

up of 69 fuel cluster assemblies, of the 14x14 rod type. In all

the rod channels, but one reserved to the intranuclear instrumen-

tation, fuels and control rods are inserted.

The fuel assembly active length is about 243 cm., the

pitch between assemblies is 19.79 cm., and the pitch between

fuel rods is 1.41 cm.

There are four experimental assemblies in the core. The

normal fuel assembly is composed of 179 fuel rods, and the expe-

rimental one has 135 fuel rods. In each of the control assemblies

there are 16 control rods (80 % Ag, 15 % In, 5 % Cd).

The fuel is U02 withthreedifferent enrichments (fig. 1).

11.2. INTRANUCLEAR INSTRUMENTATION.

As is well known, its principal purpose is to give iiifor-

a /

mation about the neutron flux distribution and the coolant outlet

temperatures in selected channels.

To this aim, the Zorita Reactor nuclear instrumentation

.. .- - consists of 24 thimbles and 2 miniature flux monitors, moving r "

through,the thimbles, along the selected channels. These moni-

tors provide flux readings in 41 points between core and bottom. -

I

11.3. REACTIVITY FOLLOW CODES.

The information obtained with the intranuclear instrumen-

tation is processed by means of three codes we are going to des-

cribe roughly.

1 INCORE: It is a data analysis code written for the treatement

of incore flux and temperature measurements.

The code determines pointwise reaction rate in the flux

thimbles, compares measured reaction rates with expected values

and rejects data according to an input rejection criterid; it

computes relative local power in each fuel assembly and in each

fuel rod chosen, relative quadrant power are calculated too, the

twenty highest values of F~ and F~ with an identifying number 4. A H Q so that hot spot locations in the core can be determined, the

rate at which burnup is being accumulated is calculated for four

axial regions for each fueled area, relative local enthalpy

rise, margin to departure from nucleate boiling, etc.

2 FOLLOW: This code provides an automated reactivity follow proce-

dure by determining the critical boron concentration at nominal

operating conditions as a function of burnup. This critical boron

concentration is found by adjusting the measured boron concentra-

tion for off nominal conditions using input reactivity parameters

3 TOTE: As a part of the analysis of incore flux measurements, the

INCORE code punchs out the rate at which various fuel regions

are accumulating burnup relative to the core average. With this

output from several INCORE runs, a core power history and the

predicted isotopic dependence on burnup, TGTE computes local

burnup, isotopic concentrations and uranium values.

11.4. REACTIVITY FOLLOW RESULTS. .. -

During the first four cycles of the Zorita Reactor, 132

sets of intranuclear flux measurements have been made. Measure-

ments of boron* concentration, temperatures, control groups po-

sitions, etc., were daily realized.

The outputs of these INCORE, FOLLOW and TOTE codes, pro-

vide such an amount of results that cannot be reproduced here

because of its extent; anyway,some of them will be shown.

burnup is the right. one.

The core average EOL burnups for the first four cycles +

were 16052 MWD/T, 7564 MWD/T, 9246 MWD/T and 10432 MWD/T respec-

tively.

We show in fig. 3 the EOL burnup distribution for the first

cycle, and in fig. 4 the second cycle EOL accumulated burnup dis-

tribution. The boron curves (boron concentration VS. core burnup)

are represented in fig. 5 and 6, for the first two cycles. 1

The core averaged axial factor FZ is represented VS. core

burnup in fig. 7 for the first two cycles; its variation with

From each of the 132 sets of flux distribution measurements

made, we have obtain from INCORE a flux map like the one repre-

sented in fig.8.

In all cases, the hot peaking factors values obtained

were found right and all of them below the design maximum values.

11.5. REFERENCES. -

1. W.D. LEGGETT. "The INCORE code". WCAP-7149

- 2. W.D. LEGGETT. "The FOL~OW code". NE-NOA-189.

3. W.D. LEGGETT. "TOTE, A code for totaling local burnup, isotopic and uranium values". WCAP-7309.

2 4 % enriched , ... 2,9'10 enriched 3,6 "10 enriched

[3 lntrumented assembly % Experimental ossembly

Control groupA @ Control group B Scram group

Fig. l ~ ' 9 0 S E CABRERA" REACTOR FIRST CORE

Rod control

Instrumentati6n channel

Fig. 2 :ROD CONTROL CLUSTER (R CC) 8 ~ 0 2 ~ ~ 9 '

U n i 6 n E l c c - t r i c a . V e s t i n g h o u s e

F i g . 3 :FIRST CYCLE EOL BURNUP DISTRIBUTION (MWD/T)

J E N

L l e s t i n g h o u - s e

Fig. 4 .-SECOND CYCLE EOL ACCUMULATED BURNUP DISTRIBUTION (MWD/TX lc3) .

0 6 0 2 ~ 0 9 3

Fig. 8 .-FLUX MAP NUMBER 86 SECOND CYCLE POWER: 509 MWth BURNUP: 6285 MWD/MTU

111. NEUTRONIC AND THERMOHYDRAULIC ANALYSIS OF A SWIMMING POOL

REACTOR DESIGNED AT JEN FOR THE REPUBLIC OF CHILE.

111.1. The reactor of CENE (Centro de Energia Nuclear del Ejgrci-

to de Chile) is an experimental swimming pool facility at

present under construction at Lo Aguirre, 30 km from San-

tiago de Chile. Its construction and design are being

carried out through a colaboration between Centro dc Ener-

gia Nuclear del Ejgrcito de Chile and Junta de Energia Nu-

clear of Spain.

The power of this reactor, 10 Mw, could be extended to 20

Mw. The core is very versatile as far as the possible

different configuration are concerned (Fig. 1). It has an

aluminium holding grid where a maximum configuration of

5x6 fuel elements are fitted, surrounded by 25 graphite

reflector elements. The control system is made up by 4 boral

aluminium cladded blades placed in two paralel planes di-

viding the core into three regions; the blades are movable

along two pair of aluminium boxes. The whole system is inmerc

sed into a cilindrical pool filled up with light water wich

acts as moderator, coolant, reflector and biological shielding

The concrete walls of the pool act as well as a side shielding

- The fuel element of t'he MTR type has 18 aluminium cladded

fuel plates. The fuel region is a dispersion of UA1 into 3 A1 0, the enrichment being 90 % wtd. U-235.

2

.. ~.

111. 2. The analysis of the neutronics (statics) includes the cal-

culation of magnitudes and parameters such as, critical

buckling, iast and thermal neutron spectra; f, ~ I , E and p;

parametric variation of the reactivity versus number of fuel

plates per element; reflector savings; 3-D distributions of

fluxes,power, fision rates and neutron importances; parame-

tric variation of the reactivity versus number of - fuel elements; Integral, diferential and overall reactivity of

the control blades; Burn-up degree; average generation

time; efective delayed neutron fraction, and reactivity

coefficients.

For these calculations the following codes have been used:

1 a) WIMS-D. This code has been developped at Winfrith (UKAEA)

for unit cell calculations. The basic liberary has 14

fast groups, 13 resonants and 42 thermal, which may be

condensed into any number of groups.

2 b) DTF-IV. This is a onedimensional multigroup code which

solves the transport equation in the Sn aproximation.

It has been used for reactivity and flux distribution

calculation.

3 C) DAC-1. This is a onedimensional, S perturbation code. n

In this reactor it has been used for the calculation of

the neutron generation time and effective delayed neutron

fraction.

4 d) TWOTRAN-G-G. This is a two-dimensional multigroup code

in the Sn aproxim&ion. It has beeneused together with

DTF-IV for the flux distribution and reactivity analysis.

The theoretical model just described was previously checked .. . through a comparison with the measurements carried out at

s the JEN-I and JEN-I1 swimming pool reactors of Junta de

Energia Nuclear 1

III. 3. The Kinetics studies carried out for this reactor consisted

of the analysis of pure nuclear accidents, and plant accidents

produced by a variation of the coolant flow. In the first

group are included the accidents due to an uncontrolled

withdrawal of the control blades from both subcritical y

critical conditions at low and full power. The same kind

of accident is produced by the unproper insertion of a

fuel element into a critical or near critical core confi-

guration.

In the secong group, the accidents originated by a total

or partial decrease of coolant flow are analysed.

The theoretical model used for these studies consists of

the following codes:

5 a) PLANKIN. (Plant -- Kinetics). This code originally deve-

lopped to simulate plant dinamics in pressurized water

reactors, has been adapted at JEN to swimming pool

reactor analysis. In this code the core is divided into

10 axial regions whose temperatures are calculated with

a simple model . The neutronics is analysed through the solution of the neutron density balance equation, with

six groups of delayed neutrons and temperature feed-back

effects. - 6 6

b) AIREK-JEN. This code, of the AIREK series, has been used fo-

the analysis of pure nuclear accidents.

. .

111.4. The thermohydraulic analysis deals with the working condi-

tions of the reactor within the proper safety margin to meet

the following design bases:

4.1. The maximum temperature must be below the melting point

of cladding and fuel and within the DNB requirements.

4.2. The cooland flow will be such as to prevent fuel vibra-

tions and hydraulic instabilities.

4.3. Coolant boiling is not allowed at any point. -

The corresponding calculations have been carried out with

the following codes:

7 a) BRANDAL.. This code studies boiling fenomena in plane

channels in steady state. It calculates pressure, coolant

and cladding temperatures, void fractions, heat transfer

in burn-out conditions, etc.

7 b) ZAPADOR.. This is basically identical, to the previous

one, but with the constants adjusted to swimming pool

reactors.

111.5. In the chapter of the biological shielding, the doses at

the relevant points of the external interphases of the

reactor have been analysed; special attention has been given

to the irradiation channels (Fig. 2 ) .

The code Sabine has been used for this purpose. This code

computes the onedimensional neutron propagation in 27 groups,

and gamma attenuation in 7 groups. . ..

- -

111.6. REFERENCES.

1. J.R. ASKEW, et al. "A general description of the lettice

code WIMS". J.B.N.E.S. Oct. 1966.

2. K.D.LATHOP. "DTF-IV, a Fortran IV Program for Solving

the Mu1Y;igroup Transport Equation With Anisotropie

Scattering", LA-3373 (1965).

3. B.M.CARMICHAEL. "DAC-1, A one-dimensional Sn Perturbation

Code", LA-4342 (1969).

4. K.D. LATHOP, F.W. BRINKLEY. "Theory and Use of the General

Geometry TWOTRAN". LA-4432 (1970).

5. C.A.NEGIN. "PLANKIN, Plant Kinetics user's:manual1'. NUS-674,

(1970).

6. A. SCHWARTZ, "Generalized Reactor Kinetics Code AIREK-II",

NAA-SR-MEMO-4980 (1960). 4

7. G. LEIRA. Internal JEN report (1970).

REFLECTOR

Plancha de A l .

-

\ NUCLEO

4 LiPl

\ Ploncho de Al . P lacas de control ( p i e )

Cajeras

Fig.1.- Reactor C.E.N.E. - Nucleo completo - . . .

~ l e m i n t o combustible

Elemento ref Lector

Cesto de irradiocidn

F I G U R A N U M . 9 --

CALCULATIONAL SCHEME FOR FAST REACTORS AT JEN ct/t ( . x i> - L. I 5s 'j

STICHTING REACTOR CENTRUM NEDERLAND Petten, July 12, 1976

Report on

Reactor Physics Activities in the Netherlands

1975-1976

1. General situation in the Netherlands with respect to Nuclear Energy

1.1. Power reactors

Two nuclear power stations are now in operation:

1) Dodewaard, 58 MWe, BWR, GE design, in operation since end of 1968,

2) Borssele, 470 MWe, PWR, KWU design, in operation since October 1973.

There are plans for three 1000 MWe reactors to be built before 1985.

On request of the government three reports were prepared, one on the

risk analysis of the fuel cycle aspects in the Netherlands, prepared by

the combined electricity producers, an other on possible effects on the

health, prepared by the National Health Board, and a third one on the

reactor safety and location aspects, prepared by the National Reactor

Safety Committee. Although those three reports came to the conclusion

that the three reactors would be acceptable, the cabinet decided for

political reasons to defer the decision until 1977 when elections for

a new parliament will have taken place.

The contribution of the Netherlands to the DeBeNeLux project on LMFBR

development is now under heavy criticism. RCN (and the other Dutch

participating organizations) will be allowed to fulfil their obligations

for the SNR-300 at Kalkar, however further activities e.g. on SNR-2

are not yet explicitly decided. Probably those activities will be rather

restricted and may be limited to safety and environmental aspects of

LMFBR's in general.

1.2. Transformation of RCN to ECN

In summer 1975 the cabinet decided that the Reactor Centrum Nederland

should be transformed into Energy research Centrum Nederland. This

transformation will be effectuated this year. An increasing part of the

activities of RCN will have to be reoriented and be more directed to the

so lu t ion of energy problems. Fields on which already some work has begun

are: wind energy applicat ions, d i r e c t conversion by MHD, fusion reac tor

s tudies , energy storage i n f l y wheels, while energy system s tudies a re

being planned.

1.3. KEMA Suspension Test Reactor (KSTR)

In May 1974 the KSTR became c r i t i c a l and i n May 1975 the design power of

1 MN was reached. Up t o now a t o t a l energy of 100 MWh has been produced.

The f u e l cons is t s of 5 pm spher ic p a r t i c l e s of 25% U (90% enriched i n

2 3 5 ~ ) + 75% Th oxyde, i n a H20 moderator. The c r i t i c a l mass i s of the

order of 22 t o 25 kg oxyde. The temperature coe f f i c i en t a t 250•‹C and

60 atm. i s about -60 pcm/OC.

Because of t h i s the f luc tua t ions a t high power a re grea t ly reduced (10 5

25%) compared t o those a t low power (decades). The reac tor w i l l be

operated u n t i l the allowed 1000 MWh have been reached (May 1977).

The system behaved well up t o now. A reference design i s being made

fo r a 250 MWe D20 moderated reac tor which seems technical ly r ea l i s ab le .

The conversion r a t i o w i l l vary from 0.8 t o 1.06 depending on the clean

up of f i s s i o n products to be achieved i n an on-line operation.

2. Reactor Physics a t RCN

2.1. STEK-project

In spring 1975 the f i r s t r e s u l t s of the analysis of r e a c t i v i t y worths of

i so topic f i s s i o n product samples measured i n STEK have been reported i n

a conference paper 1 1 I . That paper gives a shor t ou t l ine of the methods

used to obtain adjusted capture group cross sect ions f o r I z 7 1 , l o l ~ u ,

lo2Ru, and Io4Ru. More extended publicat ions on methods f o r uncertainty

estimation and adjustment calculat ions a re given i n re fs . 12-41. Several

aspects of the STEK work have been reported i n recent progress reports 151

STEK-spectra ------ ----- During the pas t year some changes i n the neutron f l u x and adjo in t spec t ra

of the f i v e STEK cores have been introduced as a r e s u l t of a carefu l

re-examination of ca lcula t ion methods and a l l avai lable experimental

information. The f i n a l l y adopted STEK spect ra a re primarily based upon

calculated spectra , adjusted to reach optimal agreement with a s e r i e s

of i n t eg ra l spectrum measurements. The changes i n C$ and @ + a r e r e l a t ed

to small adjustments in a and a of 235~, for which the KFK-INR group f C

cross section set 161 had been used. These adjustments are mostly in

agreement with the estimated error margins of the 2 3 5 ~ cross sections.

Moreover, the most recent of measurements 171 support the sign of the

adjustments. The differential STEK spectrum measurements have not been

used in this procedure, because these data would lead to unrealistic

adjustments. Previously reported discrepancies between different reac-

tivity worth normalizations 1 I I disappeared as a result of the above

mentioned procedure.

It now seems that a rather long period of developing and refining methods,

procedures, normalizations, and neutron spectrum calculations has ended.

Confidence in these methods has grown appreciably by the results obtained

so far for a series of 23 nuclides: 93Nb, 92,94,95,96,97,98,10OMO 9gTC, 101,102,104Ru 103Rh 104,105,106,107,108,110pd 1271, 133cs, 13gLa, and

14'pr. For these nuclides neutron cross sections (at, a a a el' ny' nn' and on.2n) have been evaluated mainly on the basis of nuclear model calcu-

lations (RCN-2 library). In addition uncertainties have been estimated

for the capture group constants and adjustments using integral STEK meas-

urements have been performed.

It is encouraging to note that for isotopes with reasonably well known

capture cross sections (like 93~b, Io3Rh, Iz71, 133~s, and I4'pr) the

adjustments are small compared to the a-priori uncertainties. For many

other isotopes the adjustments are not very large, i.e. for 92,95,97M0 "TC, 'O1RU, 104'105,106~d, and I3'La. For most other nuclides studied

(i.e. 94,96,08, looM0 302, 104Ru 107, 10891 lopd) is difficult ny

to evaluate and it is not surprising that large adjustments were observed.

For almost all isotopes the uncertainties were reduced considerably by

the use of the integral STEK measurements.

Five pseudo fission product cross section sets (to be used in fast breeder

design and burn-up calculations) have been generated for the following

fissionable isotopes: 235~, 238~, 2 3 9 ~ ~ , 240~u and 241~u. For the fission

product cross sections different recent libraries have been used for

intercomparison, i.e. ENDFB-4, RCN-2 (adjusted), and recent (1975)

libraries from CNEN (Bologna) and JAERI. Group cross sections have been

calculated in 26 groups for 162 different fission product nuclides.

Comparisons have been made for the most important nuclides and some dis-

crepancies between the various evaluations have been traced.

For the different cross section sets, mostly supplemented with ENDF-B

cross sections, pseudo fission product cross sections have been calcu-

lated. The main conclusions are that all recent evaluations give lower

(10% to 20%) capture rates in a fast breeder reactor than the older

evaluations 181, and that the net effect of STEK-adjusfnients for the lumped

23 nuclides is negligible. The contribution of the RCN-2 isotopes to

the total fission product capture effect is about 60%. This percentage

will be increased to about 90% when all remaining nuclides measured in

STEK have been analyzed.

H. Gruppelaar, J.B. Dragt, A.J. Janssen, and J.W.M. Dekker,

Evaluation, uncertainty estimation and adjustment of capture cross

sections for fission product nuclei, Proc. Conf. "Nuclear Cross

Sections and Technology", Washington D.C., 1975, NBS special publica-

tion 425, vol. 1, p. 165 (1975).

H. Gruppelaar, Error calculation of capture cross sections for flssion

product nuclei, Proc. of the Second Int. Symp. on neutron capture y-ray

spectroscopy and related topics, Petten, 1974, Reactor Centrum Neder-

land (19751, p. 760.

H. Gruppelaar, Uncertainty estimates of statistical theory calcula-

tions of neutron capture cross sections of fission products.

Paper contr. to IAEA Consultants Meeting on the use of nuclear theory

in neutron data evaluation, Trieste, December 1975, to be published

by IAEA, Vienna.

J.B. Dragt, J.W.M. Dekker, H. Gruppelaar, and A.J. Janssen, Methods

of adjustment and error evaluation of neutron capture cross sections;

application to fission product nuclides. Submitted for publication

in Nucl. Sc. and Eng.

Quarterly Progress reports on the fast reactor programme, recent

issues: RCN-228 (1975), RCN-239 (1975), RCN-244 (1975).

E. Kiefhaber, The KFK-INR set of group constants, Nuclear data basis

and first results on its application to the recalculation of fast

power reactors, KFK-1572 (1972).

171 R. Gwin, E.G. Silver, R.W. Ingle, and H. Weaver, Nucl. Sc. and Eng.

59 (1976) 79. - 18 1 H. Gruppelaar (come.), RCN-I pseudo fission-product capture group

cross sections, RCN-205 (1974).

2.2. Noise analysis

The noise work has continued in two different areas: temperature noise in

sodium and reactor noise in LWR's (both PWR and BWR). There has also been

some development in the hardware and software used in noise analysis.

Me~h_o_ds-G -an&rrsis A new code system for CDC-6600, named FAST, came into operation for multi-

channel noise analysis. It accepts digitized data of up to 14 simul-

taneously measured signals, and determines auto and cross power spectra

and correlation functions for all desired combinations of the 14 signals.

It uses very efficient algorithms, based on FFT.

A study is underway to investigate possibilities for performing a similar

task by a process computer and partly in hardware, using modern develop-

ments in integrated electronics.

Temperature noise in sodium (in heated rod bundles, simulating an LMFBR

fuel element)

One aim is to determine sodium flow velocities by the cross correlation

technique. After applications in a 4-rod and a 12-rod bundle, measure-

ments have recently been conducted in a partly blocked 28-rod experiment

at Petten (cooperation of RCN and GfK, Karlsruhe). The recirculating

flow pattern behind the blockage could clearly be observed. Also the

temperature noise during boiling has been measured.

In the analysis of 12-rod bundle experiments the emphasis has been on

checking models to describe the exchange of temperature fluctuations

between sodium and heated wall. A set of least-squares fitting programs

has been set-up for this purpose.

For better understanding of temperature noise near heated rods (without

boiling) a single rod experiment in cooperation with the Technical

University of Hannover is in preparation.

Another aim of temperature noise work is the detection of channel blockage,

by measuring the noise at the outlet of a fuel assembly. To this end

measurementshavebeen performed in the outlet of the 28-rod bundle.

Noise in PWR ------------ At regular intervals (begin and end of each core loading) noise measure-

ments are performed at the 470 MWe PWR of Borssele, at the request of

the Directorate of Labour of the Ministry of Social Affairs. Four series

of experiments have been performed so far. Some results have been reported

in ref. 1 11. Measurements are made with several neutron detectors around

the pressure vessel, several pressure transducers in the primary coolant,

and in one case with a displacement transducer inside the core. Neutronic

noise is considered from 0.05 - 25 Hz (above is only white detector noise).

Some conclusions are:

- The neutronic noise spectra always show the same general picture. Some

noise components increase during core life.

- Clear attenuation noise peaks due to core movements are observed (at about 15 Hz and about 12 Hz two movements each) with amplitudes of only

a few pm r.m.s.

- A reactivity noise peak is connected to vertical fuel motion. The effect

is small, but increases strongly during the life of the core.

- Around 2 Hz more complicated effects are found, possibly an interplay

between whole core movements and fuel movements inside the core at a

few strongly damped resonant frequencies.

- The main noise peak in pressure is at 6.5 Hz, and also at several

frequencies above 14 Hz, but then different for different parts of the

primary circuit.

Noise in BWR ------------ In cooperation with KEMA noise measurements have been performed in the

Dodewaard reactor (58 MWe, BWR). Fluctuations have been measured of both

in-core and ex-core neutron detectors, several pressure transducers, flow,

and a clad elongation detector. The interpretation of the results is

still underway, so conclusions cannot yet be given.

Ref _e_r_ens_es-f 2r-122

I 1 I E. Turkcan, Measurements and analysis of ex-core neutron detector

noise of the Borssele reactor (PWR) at full power.

121 Deutsches Atomforum E.V. (DAtF) Kerntechnisches Gesellschaft im DAtF,

D&eldorf, 30 G r z - 2 April 1976, p. 577-580.

2.3. Study of neutron spectrum unfolding procedures

An invited review paper on neutron spectrum determinations with the acti-

vation technique was prepared for and presented at the first ASTM-Euratom

symposium on reactor dosimetry 1 1 I . A special procedure for the determination of self-shielding factors was

developed. A computer program SELFS was made in which self-shielding cor-

rections are determined for each of the 620 energy groups, as used in the

SAND-I1 unfolding programs, thus yielding for each specified foil thick-

ness a set of 620 modified cross sections 121.

The programs SAND-I1 developed by W.N. McElroy (Hanford, USA) and RFSP-J~L

developed by A. Fischer (on leave from Budapest, Hungary) were compared

using the activation detector data irradiated in the STEK facility at

Petten.

Some results have been published last year 131. Now also the unfolding

program CRYSTAL BALL developed by W. Kam and F.W. Stallmann (Oak Ridge,

USA) is in operation at Petten.

The very first results of an intercomparison of SAND-11, REP-J~L and

CRYSTAL BALL have been presented at the seminar-workshop on unfo.lding,

held in Oak Ridge, March 1976 141 . It is expected that RCN will have a research contract with the M A for

the further study of the merits of the three most promising unfolding

codes: SAND-11, RFSP-J~L and CRYSTAL BALL.

Ref srsnsss-rslsvant-I~~-~a_r~81aeh-2121

I 1 I Willem L. Zijp, Review of activation methods for the determination

of neutron flux density spectra, RCN-241 (1976).

121 W.L. Zijp, and H. J. Nolthenius, Neutron selfshielding of activation

detectors used in spectrum unfolding, RCN-231 (1975).

131 W.L. Zijp, J.H. Baard, and H.J. Nolthenius, Neutron spectra in STEK

facility determined with the SAND-I1 activation technique, RCN-232

(1975).

/ 41 W.L. Zijp and H. 3 . Nolthenius, Intercomparison of unfolding procedures

(programs and libraries), RCN-76-059 (1976).

2.4. CTR activities

The main activity was in the frame work of a system design study group

with plasmaphysicists and electrical engineers on a high B (screw pinch) reactor. The blanket turned out to be a very noncritical item. The

studies for the European Tokamak Study Group on blankets with low Li

content will be continued.

2.5. Actinides recycling

A study is being made of possibilities for integral measurements of

actinide cross sections by irradiation of small samples under neutron

spectrum shaping covers in the HFR at Petten.

A preliminary conclusion is that under a 2 cm thick natural boron carbide

shield (supplying a spectrum somewhat analogue to that in a large fast

reactor) transmutation or fission rates in the order of 0.1 to a few %

per year of irradiation seem attainable.

A G I P : BO : CAS : C I S E : ENEL: F I A T : N I R A : TERM: V E L :

SUMMARY OF REACTOR PHYSICS ACTIVITIES IN ITALY IN THE PERIOD JULY 1975 - JUNE 1976

Ugo Farinelli, Editor

List of Contributing Organizations:

AGIP Nucleare, Montecuccolino, Bologna CNEN, Dipartim. Ricerca Tecnologica, Div. Fisica - Bologna CNEN, Dipartim. Ricerca Tecnologica, Div. Fisica - Casaccia Centro Italian0 Studi Esperienze, Milano Ente Nazionale Energia Elettrica, Roma FIAT, Sezione Energia Nucleare, Torino Nucleare Italiana Reattori Avanzati, Genova CNEN, Dipartimento Reattori Termici, Casaccia CNEN, Dipartimento Reattori Veloci, Bologna

For individual contributors please see the references

Casaccia, ~ u n e 1 976

1. THEORETICAL

1 .I Particle Transport (001

A projectional method for Fourier transformed integral equations has been investigated in connection with a one-group neutron transport problem /I/; some ideas on the more suitable functional spaces for the treatment of multigroup problems are outlined in / 2 / . A method combining the OKPL technique /3/ with the approximation theory of /I/ is studied in /4/ for plane multilayer systems. Very accurate reference results for flux distributions and eigenvalues even for high heterogeneities are obtained in /5 / . An enlarged version / 6 / of /7/ illustrates the functional spaces suitable for flux spatial distributions for critical three-dimensional systems and the convergence of iterative procedures connected with the dominance concept; the OKPL method has been extended to anisotropic scattering problems /8/ for three-dimensional geometries. The complex exponential transformation method was mathematical- ly founded in /9/ . The equivalence between the same approximation orders for tensorial diffusion and transport equations was proved in /lo/. A very fast and sufficiently accurate algorithm for solving tensorial diffusion equations is in progress. The emergent radiation of the non-conservative Milne problem is considered in /?I/ in terms of the Chandrasekhar function. Very accurate numerical results are obtained. They also include the Legendre moments of the emergent radiation and the extrapolated endpoint even for highly anisotropic scattering laws.

1.2 Non-linear reactor dynamics (801

Non-linear reactor stability problems have been studied in the frarne- work of nodal dynamics, which is felt to be a flexible and relatively un- expensive tool for such investigations when space dependence of perturbations or feedbacks is important. The problem of variable power feedback coefficients has been solved within a two-node model allowing for arbitrary time delays in the coupling between nodes /12/. The region of admissible power perturbations (region of attrac- tion) has been related to the functional dependence of nodal feedbacks on power, in both cases that such a region does or does not contain all physically realizable perturbations (stability "physically in the large" or not). For an n-nodes reactor (n>2) variable power feedback coefficients are also being considered /13/. The problem of a two-node reactor with power and temperature feedbacks represented by constant coefficients has also been treated /14/. Stability proves to be physically in the large even with destabiliz- ing temperature feedbacks, provided each power feedback is stabilizing and sufficiently large. In other important cases an appropriate region of attrac- tion has been determined.

1.3 Montecarlo Methods (BOI

In the investigation of applications of Montecarlo methods to the solution of the adjoint transport equation, an algorithm has been devised, which substantially improves other existing non-multigroup schemes, especially for the difficult case when resonant nuclides are present [for a first draft of this work see /15/1.

1.4 Generalized diffusion coefficient [CAS)

The work about the development of a generalized diffusion coefficient theory has proceded along the following lines:

a) short range objective: definition and calculation of an axial diffusion coefficient for Na + SS rods immersed in a fast neutron core (since usual 0-homogenization methods resulted not to be adequate1 /16/. Comparisons with transport benchmark calculations have been done in the frame of the CNEN-CEA agreement for fast reactors.

b) long range activities: theoretical elaboration of an iterative method to solve the multigroup transport equation by means of diffusion techniques /17/. A transport code based on this method is presently in completion.

1.5 Generalized perturbation theory [GPTI [CAS. NIRAl

A method was developed for few group constant collapsing using a va- riational principle and GPT /21/. The method is now tested for design oriented problems. GPT was applied both in the nuclide and neutron field for burn-up calculations /la/. A study was performed on the numerical calculation of the generalized importance functions using SN methods /IS/. A code chain was developed to perform non-transport sensitivity analysis both in multiplying and non-multiplying systems [the GIANT-SCOFF code chain].

1.6 Application of GPT to the nuclide field [CASI

A procedure extending the time-dependent generalized perturbation theory [GPTI to the nuclide field has been adopted for cases of interest, namely for fuel burn-up and build-up problems during the reactor life-cycle and for sensitivity analysis of actinide production estimates. The numerical results so far obtained are indicative of the reliability of the proposed method /20/.

2. EXPERIMENTAL

2.1 Nuclear reactor stochastic and noise analysis [BDI

A system for the recording and successive analysis of time series is being prepared. It will consist of an analogic tape recorder/reproducer, an AOC converter interfaced with the SPC-I6 minicomputer. The minicomputer will also be connected with the IBN 370/166 computer. The theoretical activity on the zero power reactor noise analysis has been continued but the activity is gradually shifting towards the use of the spectral analysis methods in both areas of frequency response studies and non-parametric model building.

2.2 Fast neutron spectrometry [BOI

The activity on fast neutron spectrometry by means of Bennett type counters with gamma discrimination has been continued. The electronic system has been interfaced with the SPC-16 General Automation minicomputer and programs for the on-line data acquisition have been prepared. Preliminary spectrum measurements in the central cavity of the RB-2 fast-thermal coupled reactor are in progress. The computer code chain for the unfolding of the data is being tested with the measured data. In particular the joining of ionization spectra measured in the different counters, the w correction and the gamma subtraction in the low energy range are being investigated.

2.3 Advances in reactor noise analysis [CASI

A unified theory of reactor neutron noise analysis techniques has been developed. It describes reactor neutron noise as a stochastic process with neutrons and neutron counts from a multi-detector apparatus as state variables. The reactor system is completely characterized in the time evolution of mean values and fluctuations of its state variables by a multivariate joint distribution.

The generating function of this multivariate joint distribution is shown to generate all the neutron noise analysis techniques developed by a three-decade experience in nuclear reactor experimental physics /22/.

Program DERN has been compiled for calculating the n-th order derivative of a composite function via an original algorithm. Program RAFE, a fall-out code of DERN, has been compiled for calculating a probability profile from a generating function via a recursion algorithm /23, 24/.

Program NORMOS has been compiled for utilizing a maximum-likelyhood method in the best-fitting of experimental data: among the many options, the program contains all the familiar probability profiles and distributions (discrete and continuous1 useful for treating reactor noise data /25/.

2.4 Fast neutron dosimetry [CASI

An IAEA benchmark experiment in the frame of the systematic investigation of activation detectors cross sections has been completed. Reaction rates of 15 category-I and -2 detectors have been measured both in the core and in the reflector of the TAPIR0 fast source reactor. Comparisons between the results obtained from the experiments and from a 2-D transport model show some unexpected inconsistencies. A further calculational effort with improved copper cross sections is required to validate the conclusions of the study.

2.5 Magnetic scattering in reactor (CAS)

During magnetic measurements. performed with the purpose of interpreting experimental reactivity effects, strong perturbations due to eddy currents were observed for the highest magnetic fields. After unsuccessful efforts aimed either at eliminating or at reducing eddy currents, a semiempirical method for taking into account the effect of eddy currents was developed.

Analysis of final magnetic measurements and performance of final reactivity effects are under way.

Investigation of magnetic field lines has been carried out by the LIMA code / 5 5 / , / 5 6 / .

3. LIGHT WATER REACTORS

The activities 3.1 to 3.4 were performed either in the framework of a Cooperation Agreement between ENEL and CNEN and/or under the provisions of a contract awarded to CISE by ENEL.

3.1 Three-dimensional simulation code system (CNEN-ENEL)

The three-dimensional core simulator BACONE /26/ was implemented to reproduce by computational models the thermal-hydraulic and neutronic behaviour of BWR cores versus irradiation time and irradiation conditions. Edit routines are provided to describe the thermal limit conditions of the whole core, the evolution of isotopic compositions, fuel assembly reshuffl- ing, theoretical prediction of instrumentation readings.

An option is provided to calculate fuel cycle length by'~alin~ computation. Data handling for fuel accountability (including history of each bundle) is managed via proper data bases which are used also to account in the computation for the effects connected with previous irradiation cycles.

An interface code BUBA was developed to perform all data processing necessary to feed into BACONE the output coming from a neutronics code far 20 LWR assembly analysis.

Calibration work vs experimental data available from operation history of Garigliano BWR is in progress.

An extension to account for Xe transient effects is under way.

3.2 Two-dimensional LWR assembly neutronics code

3.2.1 Implementation of the AUTOBUS Code /27/ (CISE-ENELI

The AUTOBUS code, which is derived from the BURSQUIO code was developed so as to perform in an automated way all the computations necessary to provide a library for a complete description of a fuel assembly in x, y geometry versus irradiation and void content for both controlled and uncontrolled conditions. The BEVE routine developed by CNEN /28/ was inserted for the treatment of Gd-poisoned rods.

3.2.2 Revision and calibration of the design models (CNEN-ENELI

Work was recently performed on the design methods versus more advanced models (i.e. transport theory DOT and Montecarlo KIM). As a consequence, diffusion equivalent cross sections for the heavy absorbing cell calculated in integral transport theory - basically the THERMOS code - will be used for two-dimensional assembly calculations by AUTOBUS.

Extensive experimental verification of the reliability of the design models was obtained by comparing calculated and measured data for LWR fuel assemblies including both uranium and mixed-oxide-fuelled lattices, also in the presence of Gd-poisoned rods and of plutonium - island configurations, both at room temoerature and in hot conditions.

3.3 Methods for three-dimensional, coarse-mesh diffusion analysis (CZSE-ENEL)

Within the frame of a research contract between ENEL and CISE, work has been continued on the development of the COMETA code /29/.

With respect to 1975, CISE has developed the following areas:

(il acceleration techniques, successfully introduced into the code,

(iil 2-group spectral corrections, which have just been coded on the basis of a simplified model due to Becker [^I,

[iii) coupling of a thermal-hydraulic routine with the neutronics code without affecting the overall computing time.

Studies have been initiated to provide a simplified burnup/void-dependent cross section library in a generalized form that makes no use of pre-assigned functional restrictions.

( - 1 M.BECKER, "Incorporation of spectral effects into one-group nodal simu- lators", Nucl.Sci.Engng., - 59, 276-278 (March 1976)

3.4 Application of Montecarlo methods (801

Following a series of comparisons with experimental results and other theoretical calculations, the Montecarlo cell code KIM has been adopted as a reference code for thermal reactors (both LWR's and HWR's). The checks carried out have shown the validity of the theoretical model especially concerning some new algorithms which have been introduced, such as the geometrical discrete approach.

Calculations of the power distribution in fuel elements of boiling water reactors, with rods containing uranium of various enrichments and burnable poisons yielded results in good agreement with experiments. In the presence of both uranium and uranium-plutonium oxides, there seems to be a small but systematic discrepancy with experimental results, which increases with the quantity of plutonium present in the element. Since other theoretical methods agree with the Montecarlo calculations, the attention is now focused on cross sections (although experimental results are also not free of doubts].

3.5 MUM - 3-DC features (FIAT)

MUM-3DC /30/is a two-group diffusion-depletion code operating in XYZ geometry (with a total of 420.000 space points1 or in XY geometry [with a total of 90.000 plane point) or in Z geometry (with a total of 300 mesh points).

The input by card or tapes is reasonably simple and fuel management can be handled directly by means of combinations of translations, rotations and reflexions of regions. Boron criticality searches are allowed.

The code requests macroscopic cross sections as function of burnup and uses a direct treatment of the non-linear effects of Xe, enthalpy and Doppler.

MUM-3DC is written entirely in FORTRAN-V for UNIVAC 1110. Provision is made to save data permanently on tapes for future reference.

4 . H E A V Y W A T E R REACTORS

4.1 Experiments i n t h e RB-1 r e a c t o r (B01

I n t h e per iod under c o n s i d e r a t i o n , t h e fo l lowing k_ measurements have been performed:

1 ) CLRENE l a t t i c e wi th 19-pins c l u s t e r , 1% enr iched U02; coo lan t d e n s i t y 0.58 g/cm3;

2 ) CIRENE l a t t i c e wi th 19-pins c l u s t e r , 1.15% enr iched U02; coo lan t d e n s i t y 0.58 g/cm3;

31 C I R E N E l a t t i c e with '19-pins c l u s t e r , 1 . I % enr iched U02; coo lan t d e n s i t y 0.41 g/om3.

During t h e l a s t p a r t of t h i s per iod, measurements on t h e l a t t i c e of t y p e 3 ) wi th no coo lan t have s t a r t e d .

4 . 2 Experiments i n t h e RB-3 r e a c t o r (801

The measurements concerning t h e t r a n s f e r f u n c t i o n of t h e r e a c t o r (FURIA experiment) have been accomplished; t h e exper imenta l r e s u l t s obta ined have been e l a b o r a t e d and a t e c h n i c a l r e p o r t (Ooc.CEC(7516) has been publ ished. The r e s u l t s of t h i s experiment have a l s o been used f o r t h e f i n a l i n t e r p r e t a - t i o n of t h e measurements performed on t h e two-phase c o n t r o l rods [BBI planned f o r t h e CIRENE p ro to type r e a c t o r .

A t t h e same t ime, t h e r e a c t o r c o r e of RE-3 has been changed, r e p l a c i n g t h e CEA n a t u r a l U02 f u e l e lements wi th EURATOM ( E C O ) n a t u r a l uranium e l e - ments; a l l t h e s t andard measurements t o c h a r a c t e r i z e t h i s new c o r e have been c a r r i e d o u t . Subsequently, an exper imenta l programme [named PESCIl has been performed: i t was aimed a t t e s t i n g t h e c a l c u l a t i o n method used f o r t h e des ign of t h e ex-core ins t rumenta t ion of t h e CIRENE r e a c t o r . These measurements con- s i s t e d mainly i n de termining t h e r a d i a l t r e n d of neut ron f l u x i n an e x p e r i - mental s i t u a t i o n a s c l o s e a s p o s s i b l e t o t h e one of t h e p ro to type r e a c t o r : t h e agreement between t h e measured t r e n d s of t h e neutron f l u x and t h e c a l c u l - a t e d ones ( through HETROIS and DOT codes1 has proved t o be s a t i s f a c t o r y .

The l a s t p a r t of t h i s per iod has been devoted t o sea rch t h e b e s t co re c o n f i g u r a t i o n needed f o r t h e exper imenta l s tudy o f ESSOR type f u e l elements and t e s t i n g c a l c u l a t i o n methods a s w e l l .

The c o r e c o n f i g u r a t i o n c o n s i s t s of an a n n u l a r f eed ing zone made up with s t andard r e f e r e n c e f u e l elements [ E C O type1 and a moderating i n n e r zone hav- i n g a t i t s c e n t e r t h e ESSOR element under t e s t : t h e a x i a l t r e n d of t h e neut ron f l u x i n s i d e t h e c e n t e r element and i t s v o i d - c o e f f i c i e n t w i l l be i n v e s t i g a t e d . These measurements have s t a r t e d dur ing May 1976.

4 .3 Reactor codes (CISE)

The use of DOT a s g e n e r a t o r of p roper ly "homogenized c e l l " c o n s t a n t s has been demonstrated, a s b r i e f l y r epor ted i n t h e IAEA Bologna S p e c i a l i s t s ' Meeting on Transpor t Theory /31/. Work i s s t i l l under way t o improve t h e f l e x i b i l i t y and t h e e f f i c i e n c y of t h e method, which i s u l t i m a t e l y in tended t o be implemented a s an automated system.

L a t t i c e physics , c o r e and f u e l management CISE-codes f o r a p p l i c a t i o n t o t h e CIRENE system have been q u a l i f i e d ; major improvements i n some modules a s we l l a s i n t h e homogenization p rocess a r e being implemented.

5. FAST REACTORS -

. . n u l l r e a c t i v i t y s i g n a l when compared wi th vacuum, r a c t e r i s t i c s :

I r e f e r e n c s wi th i r o n

g r a p h i t e atoms _ - -. - U-235

810 atoms - = 0.383 + 1% 0.374 + 1% U-235 atoms I - 1 -

5.1 F ~ s t c r i t i . c a 1 - experiments (CAS, VEL, 80 wi th C E A )

i r o n atoms -. = g r a p h i t e a t o m

chromium atoms -- = g r a p h i t e atoms

An e x t e n s i v e experiment:al s tudy has been c a r r i e d o u t i n suppor t of t l ~ e t icutronic des ign of t h e PEC r e a c t o r . The "PECORE" progrmme - implemented i n t h e c r i t i c a l f a c i l i t y NASURCA a t CEN Cadarachs i n coopera t ion wi th C E 4 - comprised f o u r phases:

- Reference core . Noter ia l . buckl ing, r e a c t i o n r a t e ai:d s p e c t r a l index measurements i n a c l e a n , two-region c o n f i g u r a t i o n

//

//

- Test zone. Simulat ion of a c e n t r a l sodium loop wi th t e s t f u e l element: r e a c t i v i t y , spectrum, power d i s t r i b u t i o n and gamma hea t ing measurements

0 .4

//

- Control rods . Ncz:;urements of r e a c t i v i t y , shadowing and f l u x t i l t i n g e f f e c t s . Experiments were made hoth wi th c e n t r a l ab- s o r b e r s and d i l u e n i s [ r e f e r e n c e j and w i t i ~ B4C tubes loaded a t c o r e - r e f l e c t o r i n t e r f a c e i n p s t t o r n s s i m i l a r t o BOL and EOC c o n f i g u r a t i o n s of t h e t e s t r e a c t o r

- R e f l e c t o r E f f e c t s . Comparisons of SS/Na and Ki/Na s e c t o r e f f e c t s on c o r e r e a c t i v i t y , power d i s t r i b u t i o n s and c o n t r o l rod wortiis.

The i n t c r p r e t a t i o n of the experiment bascd on t h e French "forrnu1air.e" CARNAVAL I11 has been completed /32/. Cross-checks of sonie key para:netcrs us ing ENDFA-IV f i l e and CNEN code c h a i n s a r e underway.

5.2 RB-2 the rmal - fas t exper iments (BO, AGIP, CCR)

The work c a r r i e d o u t dur ing t h i s per iod concerned t h e evaluation o f t h e i r o n arid chromium c a p t u r e c r o s s s e c t i o n t o U-235 f i s s i o n c r o s s s e c t i o n r a t l o i n t h e epi thermal spectrum Ln t h e c e n t r a l t e s t r eg ion of t h e RE-2 ..

the rmal - fas t r e a c t o r ( /33/ t o /3S/I . The hon;o~eneous mixtures , put i n t h e c e n t r a l t e s t rsg;.cn, which gave

ave t h e f o :

wi th i r o n

115

0.291 4. 1% .-

0 . 8

//

owing cha-

wi th chronium

112

0.408 + 1 % -

//

0.424

s The c a l c u l a t i o n of R' = U / 6 (S = i r o n , chromium1 has been made by

C . means of t h e balance equa t ions i n t h e " i n f i n i t e medium" ( n u l l l eakage) f o r t h e mixtures without and wi th s t r u c t u r a l materi .al . The mj.xtures xj.th s t r i i c t i l r a l m a t e r i a l have s p e c t r a q u i t e s i m i l a r t o t h a t of t h e re fe re l i ce mixture .

S I n t h e de te rmina t ion of R exper imenta l [ N " ' / N ~ , .B1O / 05) and c a l c u l a t -

C ed parameters a r e f

B l t l s e t . The r a t i o N / N has been determined exper imen ta l ly wi th an e r r o r s m a l l e r than 1 % by means of t h e Nul l -Reac t iv i ty O s c i l l a t o r Technique ( s e e NEACRP-L-120). The main problems connected with t h e i n t e r p r e t a t i o n of t h e o s c i l l a t i n g measurements have been so lved .

The exper imenta l parameter o B q 0 P 5 was obta ined by means of microchambers f of 4,O mm d iameter . Th i s r a t i o hag been measured wi th an e r r o r of about 1%.

I n t h e frame of t h e de te rmina t ion of RS, r e a c t i v i t y worth measurements have been c a r r i e d out , i n t h e same neutron spectrum, f o r i r o n , n i c k e l , chromium, s t a i n l e s s s t e e l AISI 316, Boron and U-235.

The c a l c u l a t e d parameters have been eva lua ted s t a r t i n g from ENDF/R-3 d a t a except f o r i r o n and chromium, f o r which ENDF/B-1 d a t a were used. The handling of nuc lea r d a t a f i l e s was made us ing a Bondarenko t y p e approach.

RS was eva lua ted with an e r r o r of about 15% + 20%. The r a t i o of t h e c a l c u l a t e d va lues t o t h e exper imenta l ones i s aboct 1.66 t 1.53 f o r t h e i r o n 0 .4 and 0 .8 r e s p e c t i v e l y ( c a l c u l a t e d v a l u e s - 0,0043 and 0,0041 f o r i r o n 0.4 and 0 .6 r e s p e c t i v e l y ) . For t h e chromium t h e c a l c u l a t i o n i s under way.

To improve t h e c a l c u l a t e d parameters t o use i n t h e R' e v a l u a t i o n a b e t t e r approach i s under way, using t h e NC-2 code t o handle t h e Nuclear Data F i l e .

I n a l l c a l c u l a t i o n s t h e r e c e n t d a t a publ ished by G w i n e t a l . have been used f o r t h e U-235 c r o s s s e c t i o n s . ( + I

[ + I R . G W I N e t a l . , "Measurements of t h e neut ron c a p t u r e and f i s s i o n c r o s s s e c t i o n s of Pu-239 and U-235, 0.02 eV t o 200 KeV, t h e neut ron c a p t u r e c r o s s s e c t i o n s of Au-197, ID t o 50 KeV, and neutron f i s s i o n c r o s s sec - t i o n s of U-233, 5 t o 200 KeV", Nucl.Sci. & Engng., 59, 79-105 (19761

5 .3 DPUNCT code f o r t h e c a l c u l a t i o n of t h e s c a l a r t r a n s p o r t f l u x by an i so - t r o p i c d i f f u s i o n (801

A code c a l l e d DPUNCT has been w r i t t e n f o r t h e c a l c u l a t i o n of t h e s c a l a r t r a n s p o r t f l u x f o r 2-dimensional geometr ies ( x , y ) and ( r , z l . Th i s code, based on t h e theory repor ted i n /17/ f i r s t performs some c l a s s i c a l d i f f u s i o n c a l c u l a t i o n s t o o b t a i n t h e d i f f u s i o n f l u x and some a u x i l i a r y f u n c t i o n s , end then o b t a i n s success ive approximations f o r t h e pointwise a n i s o t r o p i c d i f f u - s i o n c o e f f i c i e n t s , t h e t r a n s p o r t s c a l a r f l u x and t h e e r r o r f l u x , by s o l v i n g d i f f u s i o n - t y p e equa t ions having a n i s o t r o p i c , point-dependent c o e f f i c i e n t s .

The code i s not y e t f u l l y o p e r a t i o n a l s i n c e i n t h e t e s t phase s e v e r a l numerical problems have shown up, due t o t h e presence of n e g a t i v e c o e f f i c i e n t s which cha l l enge t h e v a l i d i t y of t h e methods normally used f o r d i f f u s i o n e - qua t ions . Moreover, t h e theory i t s e l f has t o be completed f o r some p o s s i b l e phys ica l c a s e s , which w i l l r e q u i r e t h e i n t r o d u c t i o n i n t h e code of some f u r - t h e r f a c i l i t i e s which a r e y e t t o be de f ined .

5.4 Ac t in ide i r r a d i a t i o n i n r e a c t o r s (CAS)

A s tudy was completed on t h e neu t ron ic i r r a d i a t i o n of a c t i n i d e s , and t h e p o s s i b i l i t y of e x p l o i t i n g an a c t u a l f a s t power r e a c t o r was considered wi th s e v e r a l l o c a t i o n o p t i o n s f o r s p e c i a l a c t i n i d e f u e l elements. A compa- r i s o n wi th a thermal r e a c t o r performance, was c a r r i e d ou t /40/. f lu l t ig roup c r o s s s e c t i o n s , i n Bondarenko format, f o r Am-241, Am-243, Cm-244 and Cm-245 were genera ted .

5.5 Heterogeneous f a s t r e a c t o r concept [CASI

A survey of t h e neu t ron ic performances of t h e heterogeneous f a s t r e a c t o r concept was performed. L i fe -cyc le e f f e c t s , breeding r a t i o , Na void and Oop- p l e r r e a c t i v i t y c o e f f i c i e n t s were analyzed f o r s e v e r a l f u e l element hypothes is .

5.6 I n t e g r a l experiment a n a l y s i s (CAS)

Extensive work has been c a r r i e d out f o r t h e a n a l y s i s of t h e ZPPR-2 sodium-void experiments, both i n t h e p l a t e and pin c o n f i g u r a t i o n s . Cen t ra l and o u t - o f - c e n t e r experiments were analyzed wi th 20-d i f fus ion . Use of 2D- a n i s o t r o p i c d i f f u s i o n and 30-di f fus ion i s fo reseen . Severa l he te rogene i ty a lgor i thms i n p a r t i c u l a r f o r t h e s t reaming e f f e c t s , were compared. ENOF/ /B-4 was one of t h e sources of b a s i c d a t a . F i n a l l y , t h e expected sodium void c o e f f i c i e n t i n t h e PEC r e a c t o r was analyzed i n d e t a i l e d c a l c u l a t i o n s /42/.

5.7 Cross s e c t i o n process ing (CAS, N I R A )

A modular s t r u c t u r e f o r c r o s s s e c t i o n process ing, based on ENDF/B d a t a , has been completed /46/. It has a s modules t h e CNEN ve r s ion of R I G E L , ETOE and M C ~ codes. The ETOX code and t h e 10X-STC (1DX modified t o a l low f o r t h e Stacey e l a s t i c removal a lgor i thms /41/1 modules a r e used f o r process ing i n t h e Bondarenko format . A r ev i sed ve r s ion of CALHET, which accounts f o r t h e Benois t s t reaming c o r r e c t i o n t o t h e d i f f u s i o n c o e f f i c i e n t , i s included f o r he te rogene i ty c a l c u l a t i o n s . SUPERTOG provides Pn m a t r i c e s . The modular s t r u c - t u r e provides i n t e r f a c e s wi th t h e ANISN-DOT-MORSE codes, wi th t h e C I T A T I O N code, wi th 10/2D d i f f u s i o n theory s e n s i t i v i t y codes and w i t h t h e GIANT-SCOFF code cha in from t r a n s p o r t s e n s i t i v i t y a n a l y s i s .

5.8 F a s t r e a c t o r dynamics (CAS)

The c r i t i c a l a n a l y s i s of t h e results of t h e benchmark dynamics problems a s proposed by NEACRP /43/ has we l l confirmed t h e r e l i a b i l i t y of t h e NADYP code /44/ adopted f o r performing t h e c a l c u l a t i o n s i n t h e case of one-dimensional problems, whi le divergences. a l s o q u a l i t a t i v e , a r e seen when two-dimensional problems a r e considered. I n o r d e r t o cope wi th t h i s problem, a few modules of t h e code have been c r i t i c a l l y r ev i sed , p a r t i c u l a r l y i n r e l a t i o n t o very f a s t t r a n s i e n t s . The r e s u l t s of t h i s a n a l y s i s a r e i n course of p u b l i c a t i o n /45/.

The conclus ion i s t h a t when i n t h e m e t a s t a t i c method a very s h o r t t ime s t e p i s assumed i n t h e case of a t r a n s i e n t above prompt c r i t i c a l i t y , one should t a k e i n t o account a l s o a f u r t h e r t ime cons tan t (Ross i -a lpha] . By t h i s i n c l u s i o n t h e agreement wi th t h e o t h e r methods becomes very good. Besi- des , a few p a r t s of t h e NAOYP code have been improved and r e - w r i t t e n i n view of i t s modular ve r s ion . F i n a l l y a few " s p l i t t i n g " methods f o r t h e s o l u t i o n of t h e k i n e t i c s equat ion have been analyzed and s e v e r a l numerical t e s t s have been performed. For what concerns t h e f a s t r e a c t o r s , t h e r e s u l t s seem t o i n - d i c a t e t h a t t h e q u a s i - s t a t i c methods a r e p r e f e r a b l e .

5.9 Neutronic des ign of t h e PEC r e a c t o r [VELI

Also dur ing t h i s per iod, nuc lea r d a t a used were obta ined from ENDF/B-I11 and processed t o 27 groups by means of NCxx2. The main p o i n t s d e a l t wi th a r e t h e fo l lowing :

a1 The des ign of PEC has undergone some modif ica t ions , mainly: j u s t one expe- r imen ta l channel i n s t e a d of t h r e e ; t h e use of Nickel i n s t e a d of s t a i n l e s s s t e e l f o r t h e r e f l e c t o r ; a new design f o r t h e c o n t r o l rods . I n t h e new r e f e r e n c e s i t u a t i o n , neu t ron ic c h a r a c t e r i s t i c s have been r e c a l c u l a t e d , i n p a r t i c u l a r a s concerns f l u x and power d i s t r u b u t i o n s , enrichment, c o n t r o l rod worths /47/.

bl The power genera ted i n t h e c o n t r o l rods by neutron capture . e l a s t i c s c a t - t e r i n g and gamma c a p t u r e has been evaluated f o r t h e s i t u a t i o n s of rods f u l l y e x t r a c t e d , f o r rods i n a p o s i t i o n corresponding t o nominal opera t ing c o n d i t i o n s and f o r rods p a r t i a l l y i n s e r t e d .

c l One has evaluated t h e enrichment of uranium elements t h a t would r e p l a c e t h e plutonium elements a t t h e o u t e r p a r t of t h e c o r e i n c a s e it turned ou t t o be impossible t o have a v a i l a b l e i n t ime t h e whole amount of mixed oxide elements necessary f o r t h e f i r s t loading.

6. SHIELDING

6.1 The Removal-diffusion code SHREOI (80)

The 2-dimensional removal-diffusion code SHREOI was completed with the insertion of the nuclear data library of the code SABINE-3, so as to make it self-sufficient /58, 48/.

A 2-dimensional calculation to compare SHREOI with ATTOW-8 was performed for a case in which experimental results were available ( + ) . The results, although not yet fully satisfactory, show that SHREOI is quicker and more reliable than ATTOW. Calculated values for the activation of threshold detectors are in good agreement with measurements. The agreement is less satisfactory for low energy detectors [calculated values tend to be on the low side).

(+I P.BOLDOR1, M.GIORCELL1, "Uso della teoria di rimozione-diffusione nell' interpretazione dei risultati della fase (21 dell'esperienza di scher- maggio CIRENE eseguita presso la ETN del CCR di Ispra", Nota Tecnica CISE n.73.026 (1973).

6.2 Neutron propagation experiments in sodium (CAS)

Neutron spectrometry and spatial propagation studies have been carried out in a sodium block (Ix1x1,5 m3) replacing the thermal column of TAPIRO.

Proton-recoil gas proportional counters have been used - in cooperation with CEA - to gather detailed information in the 20 KeV t 1.5 MeV region. Integral (Au, Na, Mn, U-235) and threshold (Rh, S) detectors have been irradiated in several longitudinal and transversal positions in the Na block. Spectrum unfolding and interpretation of activation measurements are close to completion /49/. A further measurement campaign for the investigation of the epithermal region by means of sandwich resonance detectors is planned for the near future. In the framework of CEA/CNEN cooperation in fast reactor studies, a re-analysis of measurements made in the French fast source facility HARMONIE with a sodium block has been performed /50/; a feasibility study of analogous measurements in TAPIRO has also been completed by comparing spectra, fluxes and geometry factors of the two reactors /51/.

6.3 Neutron propagation experiments analysis (CAS)

The analysis of the iron propagation experiment in the TRIGA reactor (ESIS benchmark) was completed /52, 53/. The ASPES iron benchmark experiment has been analyzed. A systematic work was carried out on the effects of SN approximations, on mesh-spacing both in ID (ANISNI and 20 [DOT-31, on transverse leakage for 10 calculations, source effects etc. A similar analysis for the Fe/Na mixture experiments performed in HARMONIE was carried out together with an intercomparison between the neutron propagation experiments in Na performed both on TAPIRO and HARMONIE reactors /51/. Sensitivity analysis were performed for all experiments. using the GIANT-SCOFF code chain.

6.4 Measurements of gamma-ray fluxes ICAS)

Measurements of gamma-ray dose distributions have been performed through the core and reflector of the TAPIR0 reactor by means of thermo- luminescent detectors (TLD). In particular, LiF encapsulated in stainless steel has been used in the core, and LiF encapsulated in copper in the reflector (which is also copper). The dosimeters have been calibrated by means of standard irradiation facilities (from low energies, using X rays, up to the Co-60 gamma ray energy).

Reference calculations have been carried out on a one-dimensional spherical model of the reactor using the ANISN code and the EL3 and ELI data supplied by ESIS, Ispra, for the (n,y) production and the y transport cross sections. These data do not include the fundamental contributions from inelastic scattering in Cu to the gamma production, so that the comparison with experimental results is significant only in the core region, where the agreement is satisfactory /54/.

These experiments have been carried out in view of the application of this technique to shielding measurements in the frame of the agreement between CNEN and CEA for the development of SuperphBnix.

7. FUSION REACTORS

7.1 Activation and dose-rate calculations for JET [CASI

Calculations of activation and dose-rate in the JET [Joint European Torus1 machine were performed for a reference case at Casaccia, Jnlich and Harwell. Results obtained by these three groups appear to be in acceptable agreement: they are reported and compared in /57/.

At Casaccia these calculations were carried out in three steps:

i Oetermination of neutron fluxes in monodimensional cylindrical geometry with eight energy groups, by the transport code ANISN in S6/P3 approximation with cross sections taken from the ENOF/B-1 library.

ii Calculation of activities due to intermittent operation of the machine [both the case of 100 pulses and the case of an infinite number of pulses were considered); for these calculations 19 activation reactions were used /59/.

iii Oetermination of the dose-rate in a point located on the horizontal mid plane, between two toroidal windings, just outside the vacuum vessel. In this step, averaging procedures were used for the evaluation of absorption self-shielding and geometrical attenuation.

A similar three steps procedure has been started for the final configuration of the JET machine (outline design]. The first two steps have already been completed. Neutron fluxes were calculated in bidimensional cylindrical geometry with eight energy groups, by the transport code DOT in S8/P3 approximation /59/ Cross sections taken from the ENOF/B-1 library were collapsed from 100 to 8 energy groups by the ANISN code. Three meridian planes were used in establish- ing the (R.21 geometries used in OOT calculations. Consideration of these three cases allowed three-dimensional averaging of neutron fluxes within azimuthally continuous components of the machine. For activities calculation 86 activation reactions, leading to a total of 50 radioisotopes, were used.

7.2 Use of a Plasma Focus machine as neutron source for cross section measurements [CASI

An investigation has been initiated on the possibility of using a 40 KJ Plasma Focus machine as neutron source for cross section measurements by the activation method, first with 2.45 MeV OD neutrons and subsequently with 14 MeV OT neutrons. Preliminary results show that for the feasibility of the measurements the foils to be irradiated must be placed as close as possible to the focus (in particular within the vacuum vessel1 and the repetition rate of the discharges must be increased as much as possible (an acceptable lower limit appears to be 1 discharge/minute). It also came out that the number of reactions which can be studied by the machine operating with OT mixture is larger than with pure deuterium filling. A review of threshold reactions satisfying the established feasibility conditions is under way.

The Frascati 40 KJ Plasma Focus machine has been moved to Casaccia, where it is being re-assembled /60/.

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P.DELOGU, Thesis (Physics), University of Naples, (April 19761

F.FRANC0, E.PEORETT1, "LIMA - An IBM 370/165 code for calculating magnetic field lines in a finite-dimensions coil with axial symmetry. Application to the experiment of magnetic scattering in nuclear reactor", Proc. - 5th 1ntern.Conf. on Magnet Technology(MT-51, Rome, (April 19751

A.FRANC0, E.PEDRETT1, "How to use the LIMA code. Calculations of field lines in regions of vanishing magnetic field". CNEN Report RT/FI[761 in m i n t

/57/ The JET Project-Design Proposal, EUR-JET-R5 and NEACRP-A-284.

/58/ A.OANER1, G.TOSELL1, "SHREDI: Bidimensional removal-diffusion shielding code, Foundamentals and results", Nucl.Sci.Engng., to be published.

/59/ E.PEORETT1, L.TONOINELL1, "Bidimensional calculation of neutron flux for JET outline", NEACRP-A-266 (19761

/60/ R.ABBONOANZA, E.PEORETT1. ''On the possibility of using a Plasma Focus machine as neutron source for cross section measurements by the activation method". NEACRP-A-268 (19761.

ERRATA CORRIGE

SUMMARY OF REACTOR P H Y S I C S A C T I V I T I E S I N I T A L Y I N THE

PERIOD JULY 1974 - JUNE 1975

In the previous progress report (NEACRP-L-120) some confusion occurred when labelling the various contributions.

Point 2.1 (Nuclear Reactor Stochastics and Noise Analysis) was wrongly attributed to CAS (CNEN-RIT, Physics Division, Casaccia) when it referred to activities carried out by the Physics Division in Bologna (BO). The contribution on reactor noise from Casaccia was alto- gether omitted.

The activities of point 3.1 (Three-dimensional, Coarse Mesh Diffusion Theory Codes) were carried out at CISE but by miS- take were labelled TERM.

We apologize and we have tried to do better this time.

Reactor Physics Activities in Japan

Period June 1975 to May 1976

by J. Hirota

Fast Reactor Physics

I. Japanese Evaluated Nuclear Data Library, version-1 (JENDL-1) The Japanese Evaluated Nuclear Data Library has been deve-

loped by the Nuclear Data Center in JAERI with the cooperation of the Japanese Nuclear Data Committee. The version-1 (JENDL-1) is aimed mainly to provide the data necessary for LMFBR calculations.

Its compilation was completed at the end of April, 1976.

JENDL-1 contains the data of the followinn nuclides in the ENDF/B-4 format; H, 6 ~ i , 7 ~ i , 'OB, "B, "c, "0, 23~a, 2 7 ~ 1 , Si, Cr, "~n, Fe, Ni, Cu, Mo, 18'~a, 232~h, 233~a, 234u 9 23SU 9 23gU 9

239~p, 239~u, 240~u, 241~u, 241~m, and 28 FP nuclides. For most of these nuclides, evaluation was performed by Japanese evaluators For some nuclides, however, data evaluated in other countries were adopted after examining their reliability. In order to examine the reliability of JENDL-1, various benchmark tests are now in

progress by the Japanese Nuclear Data Committee.

2. Assessment of new version of fast reactor group constants

A new version of fast reactor group constants set, JAERI- Fast set version-11, was produced in a combined basis of the ENDF/B-IV file and the new evaluation of the resonance data for 2 3 5 ~ 9 238~, and 239~u.

In order to assess the overall applicability, benchmark tests

including twenty-one test problems were carried out. Particular

attention was paid to the inelastic scattering cross section of 238~: An optimization based on the least-square fitting of effective multiplication factor and central reactivity worth for the

tventy-one problems was made for the inelastic cross section,

preserving the transfer matrix. It is concluded that tlic :neiastic

cross section of 2 3 8 ~ in the JAERI-Fast set version-I1 is suitable

to reproduce the integral data. A series of analyses were also

made for Doppler coefficient and sodium void measurements carried

out on the MZA, MZB, ZPPR-3, ZPR-3-47, ZPPR-2, and SEEOR cores.

Considerable improvement is seen in the calculated results compared

with the old version.

3 . Enlargement work of FCA and the Assembly VII-1

Under the contract between the PNC and JAERI, the enlargement

work of the FCA was completed in June, 1975. In consequence, the maximum size of assembly which can be built at FCA is now 2.8 m$

x 2.7 mH. During the enlargement work, the following functions

were also provided; temporary removal of the central 3 rows x 3 columns matrix tubes for inserting a test piece such as a fuel. or

control rod subassembly, and driving of two safety or control rod

drawers simultaneously by one drive mechanism.

Immediately after the completion of the work, the Assembly

VII-1 was built. The assembly consists of the Pu fueled sector-type

test region simulating the inner and outer core concentrations of

MONJU and the U-235 fueled driver region. The compositions of the

driver region were determined to realize the following three quan-

tities in the driver region as close as possible to those in the

test region; (i) fundamental mode spectrum, (ii) material buckling

and (iii) diffusion coefficient. The assembly went critical with 0

about 1400 core volume and two different sector angles, 60 and 90•‹, were applied to investigate the validity of the sector-type

experiment. Fission rate distribution in two zone core configula-

tion, density coefficient of core materials, B4C control rods interaction effect, 'OB enrichment effect and anisotropic effect

of diffusion coefficient have been measured on this assembly.

A shielding experiment will be started in the next month

mainly for obtaining informations on the heat generation of the

fuel subassembly stored in the fuel rack of MONJU: Two sodium

tanks (1.5 x 1.5 x 0.8 m each) were already installed on the upper

frame of FCA. The fuel drawers will be inserted in the lower

tank to simulate the fuel subassemblies stored in the fuel rack.

'!he fission rate and y-dose distributions will be mcasui-,:.. .,~i.l:;:

fission counters and TLDs, throughout the inner core, oi.~rer c w - a ?

blanket, shield, sodium and fuel rack region.

4. Methods for estimating reactivity worth of multiple control rods

Two simple methods have been proposed for estimating experimentally the reactivity worth of multiple control rods. In the first method('), the contribution of each control rod to the

multiple-rod worth is obtained by the measured single-rod worth,

to which the effect of a flux distortion due to other rods is corre'cted. The correction factor is obtained from the reactivity worth of the sample measured in the systems with and without an

individual rod. The second method(') has been developed using the higher

order perturbation technique. The interaction effect between multiple control rods can be estimated from a linear combination

of the interaction effect between two or three control rods depending on the degree of the interaction effect. The validity of the method was examined by using experimental results in PHENIX and also numerically for MONJU. Both results were quite satisfactory.

However, for the commercial-size fast reactor, the interaction effect between multiple control rods is too strong to apply this

method directly. A new concept of "quasi-control rod" is introduced to the method for the commercial-size reactor; a pair or a triple of control rods arranged symmetrically with respect to the core center are treated as a "quasi-control rod". The validity

of this technique was examined numerically for a 1000 MWe fast ( 3 ) reactor and its result is quite satisfactory .

(1) Nakano, M. : JAERI-M 6504 (1976). (2) Mitani, H. : Nucl. Sci. Eng., 51, 180 (1973). (3) Mitani, H. : Estimation of the Multiple Control Rod Worth

with Strong Interaction Effect in Large Fast Reactors, to be

published in J. Nucl. Sci. Technol.

5. Numerical study on cell homogenization procedure in sodium

void reactivity analysis

Theoretical analysis of the sodium void reactivity in fast

critical assemblies depends sensitively on details of the calcula-

tional method. However, much is left over about the adequacy of the conventional cell homogenization procedure and about the

applicability of diffusion coefficients by various definitions including the Benoist's method. In order to obtain quantitative solutions to these problems, a numerical study was performed in the NAIG Nuclear Research Laboratory for two simplified reactor models; A) a one-dimensional slab reactor and B) a cylindrical

reactor with infinite height, both being composed of realistic plate-lattices. Rigorous solutions for keff and void worths in

these reactors were obtained with one- and two- dimensional SN

codes for the models A and B, respectively. Results of the comparison between the rigorous and the equivalent-homogeneous solu-

tions are summarized as follows. 1) Void worths are insensitive to the buckling assumed in the course of the cell calculation, but it is not the case for keff. 2) Accurate prediction of keff by equivalent-homogeneous method requires the exact treatment of

the neutron leakage from unit cells in the course of the cell calculation. However, the use of group-independent cell buckling

is practically adoptable. 3) The use of the Benoist's first

approximation for diffusion coefficients overestimates the leakage component of void worths by several and ten percents for the

directions perpendicular and parallel to the plates, respectively. 4) Conventional D (=1/3itr) shows good applicability for void worth calculation.

A part of this work will be published in J. Nucl. Sci.

Techno1 . . 6. Reactivity measurement on far-subcritical fast system

Experimental studies have been made to obtain the reliable

values of subcriticality in the fast systems. Neutron source multiplication, pulsed neutron and source jerk methods were mainly

used for the experiments. The static reactivity p = (keff - I)/

keff is employed as the degree of subcriticality. By introducing the neutron detection efficiency of a detector, the formulations relating the measured quantities to the static reactivity can be derived for each method. The correction factor is then calculated

using the changes in the detection efficiency and other parameters.

Introducing the correction factor, the values measured by each method come to agree within 3 %('I. It is concluded that the systematic discrepancy in the measured reactivities between the static and the dynamic methods is disappeared and the experimental values consistent rigorously with the reactivity defined can be

obtained.

(1) Mizoo, N., et al. : Reactivity Measurement on Far-Subcritical

Fast System, Specialists Meeting on Control Rod Measurement

Techniques, Cadarache, April, 1976.

Thermal Reactor Physics

7. Measurements of large negative reactivity worths of multiple control rods at SHE Experimental and theoretical studies have been made to obtain

the techniques for determining large negative reactivity worths as (1) accurately as practicable .

Measurements of reactivity effect of the experimental multiple

control rods arranged in ring were made by the pulsed neutron method on a heavily reflected graphite-moderated 20 % enriched multiplying system. The atomic ratios of C to 2 3 5 ~ in the core are 2226 and 6628 for SHE-8 and SHE-T-1 respectively. Subcritica- lity was determined in the static reactivity from the measured prompt neutron decay constants by making a reasonable correction

to the well-known King-Simmons formula due to the change of the

neutron generation time estimated by calculation.

New multi-point type formulas were derived which replace the single-point type expressions for methods of "area-type" pulsed-

neutron, source multiplication and rod drop. In the formulas, the reactivity value is derived by integrating all the neutron counting data from every part of the reactor core to sweep out effects

of the kinetic distortion and the spatial harmonics. Experiments

were made on SHE-T-1 to examine the proposed integral versions and to confirm the validity of the King-Simmons formula with a large correction for the change of the neutron generation time.

The numbers of measuring points in the core are 16 and 48 for the

pulsed neutron and rod drop methods, and for the source multiplica tion method, respectively. Discrepancies in the reactivity values

evaluated down to s 35 $ subcritical by the use of the multipoint formulas are within only Q 5 % among the different experimental methods.

The polarity correlation method is successfully applied to the experiment to measure the reactivities from critical to nearly shut-down state (-12 $) on both SHE-8 and SHE-T-1 which have the large neutron lifetime of s 1 ms.

(1) Kaneko, Y., et al. : Measurement of Multiple Control Rod

Worth at Semi-Homogeneous Critical Assembly, Specialists Meeting on Control Rod Measurement Techniques, Cadarache,

April, 1976.

8. Burnup characteristics of Mark-I11 core, VHTR

Burnup characteristics of Mark-I11 core, the Very High-Tempe- rature Gas-Cooled Reactor (50 MWe), have been analysed as a part of the first conceptual design work in the Kawasaki Heavy Indust- ries, Ltd..

Emphasis was put on studying in detail the core power distri-

bution with partially inserted control rods. Block smeared few group constants were obtained mainly by the code DELIGHT-B, the modified version of DELIGHT-2('). Burnup dependency of the shielding factor of the burnable poison was investigated for the

double heterogeneous structure (the burnable poison pin is embedded in the graphite block and the pin is mixture of B4C grain and carbon powder). The shielding factor of the control rod was

calculated by the transport code TWOTRAN based on a super-cell. model. Core burnup was calculated by the three dimensional diffusion code CITATION. 5 energy groups and the Tri-Z model were applied to the 1/6 segment of the whole reactor with a rotational

boundary condition. Mesh number was 14 x 28 x 42 (axial). 7 time steps were taken for the core life of 600 days. At the each time step, more than 3 extents of rod insertions were investigated to keep critical and to have a preferable power distribution. Variable range of the power peaking factor is confirmed and it is

shown that power share among main orificing regions in the core

does not deviate markedly during the core life.

Cl) Shindo, R. and Hirano, M. DELIGHT-2; The Point Reactivity Burnup Code for High-Temperature Gas-Cooled Reactor Cells, JAERI-M 5661 (1974) .

9. Reactor physics activity in DCA for FUGEN

For the investigation of the core characteristics of FUGEN, studies of the reactor physics parameters have been continued by

using uranium and plutonium oxide fuel in DCA (Deuterium Critical Assembly). The recent studies made by using especially 0.87 wt/o

a enriched reactor grade plutonium fuel are as follows: Lattice parameters such as 6 25 , p 28 , 628, 649, 6:: and

thermal neutron flux disadvantage factors were measured in DCA clean core having 25.0 cm square lattice pitch, using coolant of

0 % and 100 % void fractions. Microscopic and macroscopic power distributions in DCA core and material bucklings were also measured. The same kinds of reactor parameters are now being

measured for the DCA core with boron poison dissolved in the D20 moderator. Temperature coefficients of reactivity on both H20 coolant and D20 moderator are scheduled to be measured at the end of this year by raising the whole core temperature up to 80 "C.

Recently, a preliminary experiment was done in the temperature range from 22 OC to 40 OC, obtaining favorable results.

10. Control rod effects in light water moderated Pu02-U02 lattices Control rod effects in light water moderated Pu02-U02 lattices

have been studied with blade type absorbers using a light water

moderated critical facility TCA. The absorber blades were made of Cd-A1 alloy and B4C-A1 mixture. One of the blades was loaded

to separate the rectangular prism lattices which consist of 3.0 w/o enriched Pu02-natural U02 fuel rods. The following

measurements are in progress; reactivity worths using the pulsed neutron technique, power distributions with the y-scanning method, gold wire activity distributions by thermal and epi-thermal

neutrons, and fast adjoint flux distributions by scanning a 252~f source through the lattices. The distributions of power, flux and

adjoint flux were measured perpendicular to the blade. These

values were obtained for five kinds of Cd and six kinds of Ij4C

contents in Al, and for two kinds of water to fuel volume r a t i ~ in the lattices. The measured results are evaluated comparing

with measured results in 2.6 w/o enriched U02 lattices for ,the same absorber blades.

11. Exact and simplified calculations of thermal neutron

spectrum in a control rod array The wing of the cruciform control rod used in the current

BWR is consisted of a linear array of B4C poison rods with stain-

less steel cladding. In the thermal neutron spectrum calculation for this system, the slab approximation (S.A.) has been conventio. nally used. However, Makino, et al. who calculated the reactivity

worth of the infinite array of poison rods by the use of the

collision probability method, pointed out that the S.A. leads to a significant error (J. Nucl. Energy, 23, 1969). In the NAIG

Nuclear Research Laboratory, applying their collision probabili-

ties to the THERMOS code (the Exact Method), the error of the S.A. in the thermal neutron spectrum calculation was assessed. The

comparison showed that the S.A. gives 13 % larger value of the effective absorption cross section in the control blade region. The above error results mainly from the difference of the spatial variation of the flux in the vicinity of the poison rods and the spectrum in each region is almost identical with that of the Exact Method. Thus, an approximate procedure has been proposed: spectrum calculation by the S.A., energy condensation in each region and then one-group flux calculation in the exact geometry.

The numerical tests show that this procedure is satisfactory both in computing time and in accuracy.

Fusion

12. Fission-rate distributions in lithium and hybrid fusion

blanket assemblies In order to investigate the behavior of neutrons in fusion

reactor blankets, a series of relative fission rate distribution

measurements in lithium blanket assemblies have been carried out and a discrepancy between measured and calculated results h a < . l j c c l ~

observed. (1) 9 (21 9 (31

For further investigation of this discrepancy, absolut~ fission-rates of 23SU 3 237Np, 2 3 8 ~ and 2 3 2 ~ h have been measured

by micro-fission-chambers in four types of spherical blanket

assemblies prepared by loading blocks of lithium and/or natural uranium and/or graphite in stainless steel matrix. The four

assemblies are named Li, Li-C, U-Li and U-Li-C Assemblies corresponding to their respective configurations. Source neutrons are

generated at the center of the assemblies by D-T reaction using a 300 KV Cockcroft-Walton type accelerator. An a-monitor was set

up on the accelerator to detect the associated a-particles by 4 T(d,n) He reaction. The absolute neutron yield is estimated by

the counts of this u-monitor. The fission chambers were calibrated by the use of the same accelerator, assuming that the fission cross

sections at 15 MeV referred are correct. The overall experimental errors are mostly less than 10 % and about 7 % on the average.

The results of the measurement were compared with those of one-dimensional transport calculations employing 100-group neutron cross-sections obtained from ENDF/B-IV. The C/E values of 23SU

and 2 3 5 ~ fission-rate in the four assemblies are given in the references. C4)9C5). It is shown that the C/E values of 232~h, 2 3 8 ~ and 2 3 7 ~ p fission-rate decrease with the distance from the

center. A large overestimation of 2 3 5 ~ fission-rate by calculation is observed in the graphite reflector region. The disagreement is found to be mainly caused by the incapability of the adopted multi-

group cross-section production codes in taking account of angular distribution of the secondary neutrons from nonelastic reactions.

(1) Hiraoka, T., et al. : Nucl. Fusion Special Suppl., 1974,

Proc. Symp. Fusion Reactor Design Problems, p. 363, IAEA (1974).

(2) Maekawa, H., et al. : Nucl. Sci. Eng. - 57 335-340 (1975) (3) Maekawa, H. and Seki,.Y. : JAERI-M 6495 (1976) (NEACRP-A-244)

(4) Maekawa, H. and Seki, Y. : Fission-rate Distribution in Li and Hybrid Fusion Blanket Assemblies-Experimental Method and Results, to be published in J. Nucl. Sci. Technol.

(5) Seki, Y. and Maekawa, H. : ibid-Analytical Results and

Evaluation

15. Measurements of angle-dependent emission spectra of U-T

fast neutron from the piles of blanket and shielding materials

Measurements of fast neutron spectra (from 15 MeV to about 0.3 MeV), emitted from the systems of graphite, lead, U02 and Li with the injection of 14.5 MeV near parallel neutron beams, have been carried out by the TOF (3 % 5 m flight path) method based on the associated particle method. in Osaka University. Angle-

dependent emission spectra for 4 angles (0•‹, 4S0, 90' and 135") from the near cubic systems of graphite, lead and U02 were

measured. Forward (0•‹, 45") and backward (135") emission spectra from the slab systems of Li, UOZ, graphite and lead were also

measured. The obtained emission spectra show marked angular dependency, particularly in the region from 14 MeV to few MeV.

Measured spectra for graphite and lead systems were compared

. with the anisotropic transport calculations by the use of ENDF/B- I11 (for lead) and ENDF/B-IV (for graphite). Complicated angle-

dependent structures of the spectra from the graphite can not be explained by the calculation with P-5 components of elastic scatterings, though the backward (135") data show fair agreements. For lead, calculations with P-5 (elastic) and P-0 (inelastic)

scattering components estimate for lowly the spectral data from 10 MeV to about 5 MeV in comparison with the experiments for forward angles (0" and 4S0), while the backward data (135') show fair agreements. It is pointed out that anisotropic (forward) components of inelastic scattering should be taken into account for the high energy region.

(1) Takahashi, A., et al. : Neutronics Experiments of D-T Fast Neutrons by the Application of the Associated Particle

Method, to be published in J. Nucl. Sci. Technol.

14. Monte Carlo method for the analysis of neutral beam injection

into a torus plasma An optimization of the neutral beam injection has been per-

formed by using the Monte Carlo method on the analogy of the

neutron transport. Absorption rates of energentic neutrals were

analyzed for various injection directions and energies. The maximum beam absorption is achieved for the injection directed

roughly to the middle point between the inner boundary of the torus

plasma and the center. The energy spectrum of the neutrals escap- ing from the plasmas was calculated. The analysis shows that the

Maxwellian fitted temperature of the outcoming neutrals can be

used as a measure of the plasma temperature. For the purpose of the calculations mentioned above, new com-

puter routines have been incorporated into the general purpose neutron and gamma-ray transport code MORSE, to deal with the diffuse beam source, detectors for the outcoming neutrals and the axi-asymmetric distribution of neutrals in plasmas.

15. Analyses of spherical implosion process The behavior of the spherically imploding shock wave has been

analyzed by assuming its dynamics as the self-similar motion.

Quantitative results were obtained for density, pressure, temperature and shock strength of plasma gas. Investigations were made

both on the correspondency of the self-similar motion and the

possibility of the existence of the self-similar solution to the

practical implosion process. Additionally the way to produce extremely high compression was examined from the standpoint of the self-similar motion. It is shown that none of imploding process

induced by multi-shock waves can be analyzed under the assumption (1 1 of the self-similar motion .

The effect of neutron heating in laser-induced fusion pellet was investigated revising the laser-induced fusion code MEDUSA. The formulation to take account of the neutron heating was made by the collision probability method.

1 Ishiguro, Y. and Katsuragi, S. : Self-similarity Analysis of Spherical Implosion Process, to be published in JAERI-M

Report.

Shielding

16. Mockup experiments for repairing the shield of MUTSU

A series of shielding mockup experiments have been performed

to confirm the validity of basic calculational methods and to obtain useful design data for complicated shield configulations of the nuclear ship MUTSU.

The repairing plans for the primary shield of MUTSU are as follows:

1) Replacement of mirror insulators in the gap between the

pressure vessel and the primary shield by serpentine rock insulator, clisotile.

2) Addition of polyethylene and serpentine sand shield at the

void space around the control rod mechanism above the pressure vessel.

3) Addition of polyethylene shield in the double bottom under

the reactor container.

4 ) Replacement of the upper shield of ordinary concrete by serpentine concrete shield.

Mockup experiments were carried out on the above items using JRR-4, a swimming pool type reactor. Neutron and gamma-ray measurements were carried out by threshold detectors, BFg-counter, Bonner ball, proton-recoil counter and thermoluminescence detectors. The analysis of the experiments is underway by using the SN codes ANISN and TWOTRAN, with basic nuclear data from the ENDF/B-IV and POPOP4 library. Preliminary results indicate that, for the experiment related to the addition of shielding materials above the pressure vessel, the calculated epithermal flux agrees within a factor of 5 with the observed value. The addition of clisotile between the pressure vessel and the primary shield reduces the

epithermal flux by three orders of magnitude, together with the effect of the addition of polyethylene and serpentine sand shield above the pressure vessel.

17. Two-dimensional calculations for the primary shield of JOYO Two-dimensional calculations for the primary shield of JOYO

have been performed and compared with the previous shield design

calculations. In this calculations, a special care is taken to deal with radiation streaming through the gap between the rotating

plug and the heavy concrete pedestal. The two-dimensional transport code TWOTRAN-I1 was adopted in the calculations. The ?Field geometry of JOY0 was divided into five subsections and treated as a bootstrap problem. The results of the two-dimensional calculations indicate that there is a considerable upward streaming through the gap between the reactor vessel and the graphite

shield. This streaming component raises the radiation level above the slot between the rotating plug and the pedestal.

18. Shield benchmark experiments at YAYOI and sensitivity studies

Neutron flux and gamma-ray dose distributions in iron slab in thickness up to 45 cm have been measured using the Upper Column of the Fast Neutron Source Reactor YAYOI. The neutron flux was obtained with threshold detectors, resonance detectors and gold foils. The gamma-ray dose was obtained with thermoluminescence detectors of 7 ~ i ~ . The measured neutron and gamma-ray distributions were compared with those calculated by the ANISN, DOT and MORSE codes. The cross sections base on the ENDF/B-IV file for neutron and the POPOP4 library for secondary gamma-ray production data.

Sensitivity analyses for secondary gamma-ray cross sections are being carried out using the ROSETTA code system(') for the

reactor shielding benchmark problems proposed by the NEACRP. Adoptability of a coupled neutron and gamma multigroup cross section set (Loon + 30y, P5) as the standard set for core/source and shield calculations in Japan is now being discussed at the working

group of the Japanese Nuclear Data Committee. For processing basic nuclear data, a revised RADHEAT code system will be used which embraces detailed core physics treatment's for resonance, thermal group and heterogeneity.

(1) Miyasaka, S., et al. : Sensitivity and Uncertainty Analysis

for Iron Cross Sections, Specialists Meeting on Sensitivity Studies and Shielding Benchmarks, Paris, Oct., 1975.

PNC N241 76-07 NEACRP ( JAPAN )

Review of Fast Reactor Physics Act iv i t ies

Relevant to LMFBR Programme in PNC, JAPAN

June 1975 - May 1976

T . INOUE

Power Reactor and Nuclear Fuel Development Corporation, Tokyo

1 . Introduction

Development of the fast reactor system has been in progress as

the national project in Japan since the long range benefits expected

from the use of fast breeder reactors f o r energy have long been re -

cognized. In recent years this recognition has evolved into intensive

developmental programs on the Experimental Fast Reactor, JOYO, and

the Prototype Fast Breeder Reactor, MONJU, under the responsibi l i ty

of the Power Reactor and Nuclear Fuel Development Corporation, PNC,

wi th cooperation of industr ies, u t i l i t ies , various research insti tutes

and universi t ies (1).

Construction work of JOYO was in i t iated at the 0 -a ra i Engineer-

ing Center, OEC, of the PNC in 1970 with a target of f i r s t c r i t i ca l i t y

in the end of this year. Most of the research and development works

f o r JOYO a re nearing completion. The design of MONJU has been in

progress since 1967, wi th the support of various research and devel-

opment works at the OEC and laboratories of the manufacturers.

Review of the MONJU program i s now under way by the government, to

be fol lowed by safety evaluation, and construction of the plant to begin

in 1978.

2. E x ~ e r i m e n t a l Fast Reactor. JOYO (2).

Instal lat ion of a l l the components and associated piping works of

JOYO reactor was completed in the end of 1974 and various commission-

ing tests began in January 1975. Sodium was charged into the pr imary

and secondary dump tanks i n the last summer to be followed by the

pur i f icat ion test in the pr imary and secondary loops. 119 core fuel

subassemblies and 220 blanket subassemblies, which i s a sufficient

number f o r composing the in i t ia l core, have already been fabricated.

The core fuel subassemblies w i l l be transported to the reactor s i te

f rom the fabrication plant located at the Tokai Works of the PNC i n

the near future. A l l rad ia l blanket subassemblies, dummy core ele-

ments and re f lectors are already loaded in the reactor vessel.

Fol lowing the high temperature sodium tests and a leak test f o r the a i r -

t ight steel containment vessel, the f i r s t c r i t i ca l i t y test i s cur ren t l y

planned to take place in the end of th is year.

The JOYO reactor has a core region with mixed oxides of uranium

and plutonium, surrounded by a blanket of depleted uranium. The re -

actor has two identical sodium loops, the heat-removal capacity of each

loop being 50 MWt. Each loop consists of a pr imary cooling system,

an intermediate heat exchanger, and a secondary cooling system; heat

i s dissipated into the atmosphere. The reactor, the whole pr imary

cooling system, i t s associated components are contained i n an contain-

ment vessel. Ver t ica l c ross section of the reactor structure i s shown

i n F ig . 1 .

Many of the developmental act iv i t ies related to JOYO are ei ther

completed o r nearing completion, such as physics experiments of the

core with the Fast C r i t i ca l Assembly, FCA, in the Japan Atomic

Energy Research Insti tute, JAERl (3) (4) (5) , fu l l scale engineering

mock-up of major components, fuel i r rad ia t ion experiments using fast

reactors operating in foreign countries.

JOYO has been or ig ina l ly designed as a 100 MWt plant. However,

i n consideration of being the f i r s t sodium cooled fast reactor in Japan,

an output of 50 MWt was adopted fo r JOYO as the f i r s t step and approved

by the regulatory authority in 1970, leaving the target power ra is ing

as the next step.

The present program includes to increase the reactor power to

75 MWt (MK-I, phase 1 1 ) as the second step, without any modification.

I t i s confirmed to be able to operate with reasonable safety margins at

th is second power level. Because many kinds of research and devel-

opment act iv i t ies had been ca r r i ed out in para l le l wi th the construction

of JOYO (5). Prel iminary hearing by the regulatory authority i s i n

progress to obtain the 75 MWt operation licence.

Future program f o r JOYO includes also a plan to modify the pres-

ent core configuration into a more suitable core (MK-II co re ) f o r i r rad ia -

t ion of fuels and materials, af ter successful operation wi th the present

core. With the modified fuels having pins of smaller diameter, higher

neutron f lux w i l l be attainable f o r i r rad ia t ion purposes. Th is MK-II

core w i l l be operated at 100 MWt which w i l l be the th i rd phase step of

reactor power.

Matr ices and main parameters of JOYO MK-I and MK-I1 core a re

shown i n F ig . 2 and Table 1 respectively.

3. P r o t o t v ~ e Fast Breeder Reactor, MONJU (7)

The prototype fast breeder reactor, MONJU, i s now under rev iew

by the government, since last August, which means a special process

i n Japan evaluating the design of the plant, research and development

works, the significance brought by constructing the MONJU plant in the

development program of the fast breeder reactor in Japan. Fol lowing

th is review, safety evaluation w i l l be conducted. Therefore, i t i s

current ly expected to s tar t construction of the reactor in 1978, aiming

at c r i t i ca l i t y i n 1984.

MONJU i s 714 MWt, 300 MWt, l iquid sodium cooled, loop type fast

breeder reactor plant, fueled wi th mixed oxides of uranium and pluto-

nium. The reactor core i s zoned f o r two different processes of

plutonium enrikhment, surrounded by a blanket of depleted uranium.

The expected mean fuel burn-up i s about 55,000 MWD/T as the 1st core

and 80,000 MWD/T as the target. Breeding ra t i o i s expected to be 1 .2 .

The reactor i s equipped with 19 control rods composed of 12

regulating rods, 4 safety rods and 3 back up safety rods, and B4C i s

used as the poison material. Prov is ion i s made f o r instrumentation

on the complete core and port ion of the rad ia l blanket. The design

provides two thermocouples and a f low meter probe f o r each core sub-

assembly and two thermocouples f o r each selected innermost rad ia l

blanket subassembly. Core clamping mechanism i s equipped around

the core preventing the bowing and vibrat ion of the core elements

dur ing the reactor operation and being rel iable f o r the refuel ing.

The heat generated i n the reactor i s transported by three pr imary and

secondary sodium loops to three sets of once-through steam generators

that produce 132 k g / c m 2 ~ steam at 487 OC. The whole pr imary sodium

system and the reactor a re located below the main operating f loor in an

air- t ight containment vessel, approximately 50 m in diameter and 80 m

i n height. Ver t ica l c ross section of the reactor s t ructure and core

configuration of MONJU a re shown in Figs. 3 and 4.

Extensive research and development works relevant to MONJU

a re being ca r r i ed out not only at the OEC, Tokai Works and JAERl but

also at various other research organizations including universi t ies,

national research insti tut ions and industr ia l companies as well as in

overseas faci l i t ies. Concerning steam generator, the f i r s t unit of 50

MW Steam Generator Test Fac i l i t y had operated successfully without

any leakage since June 1974. I t i s now dismantled and under detailed

inspection as or ig ina l ly scheduled. The second unit of 50 MW steam

generatorhas been instal led in the end of last year and started to

operate with the fu l l load condition. Another fac i l i ty completed in the

OEC last year i s the Steam Generator Safety Test Loop, often cal led

by the name of SWAT-3. The purpose of the loop i s to obtain neces-

sary information of behavior of steam generator and i t s associated

devices, at an unl ikely large sodium water reactor accident. Engineer-

ing tests on major components such as core internals, control rods and

refuel ing machines are being made. I r rad ia t ion tests of fuels and

materials are being performed using DFR and Rapsodie.

F o r the reactor physics act iv i t ies on the MONJU core, fol lowing

fu l l scale mock-up experiments conducted in col laboration with UKAEA,

cal led the MOZART programme (e) , par t ia l mock-up experiments a re

i n progress wi th F C A in JAERI. Some shielding experiments a re be-

ing made using a fast neutron source reactor, YAYOI, of Univers i ty of

TOKYO.

Main Darameters of MONJU core a re shown i n Table 2.

4. Fast Reactor Physics Act iv i t ies

1 ) Mock-up Experiment and Analysis of Prototype Fast Reactor

The par t ia l mock-up experiments f o r MONJU have been performed

at F C A (JAERI) since August 1972. A f te r the physics mock-up ex-

periments on VI-1 , VI-2 and VI-3 assemblies were finished, the F C A

matr ices were expanded from 35 rows by 35 columns to 51 rows by 51

columns to begin the engineering mock-up experiments on VII-1 assem-

bly. Th is enlargement work was started in September 1974 and com-

pleted by the end of June 1975. Now the experiment on VII-1 assembly

i s i n progress.

VII-1 assembly i s a sector-type to measure rad ia l ly simulated

physics character ist ics. At present, a study of BqC control r o d

react iv i ty i s going on. The other experiments such as measurements

of sample (especially the higher plutonium) worth, f iss ion product

effect, studies of the plutonium build-up in the blanket and heat produc-

t ion of a fuel assembly i n storage rack w i l l be performed in the near

future.

The analysis of MOZART experiment has a r r i ved at f inal stage

and the calculated integral quantities have been compared wi th the

measured ones. The method of analysis has been set as close to the

design method as possible in o rder to get confirmation in the con-

fidence of design works of MONJU. Effective mult ipl ication factors

a re wel l calculated with an accuracy of higher than 0.5 percent.

The uncertainty due to heterogeneity effect i s s t i l l major source of

e r r o r . The calculated values of reaction rate distr ibut ions i n the

core agree within 3% with the measured ones. In the rad ia l blanket,

however, the maximum discrepancy i s about 12% i n the case of

Pu(n .f) reaction. CaIcuIationaI-to-experimental ra t ios on sodium

void worth f o r the small voided region close to the centre of core

a re 1 .O1 ? 1.1 % and 1.13 f 1.0% fo r the plate and p in geometries

respectively, and these tend to be enlarged f o r the voided regions

f a r f rom the core centre. Overal l calculated values on control

r o d worths agree with the experimental ones within about 5%. No

significant t rend in calculational-to-experimental ra t ios i s observed

against the control r o d position o r the number of control rods involved.

Some resul ts of these mock-up experiment analysis are summarized

i n Table 3.

2 ) Evaluation of Actinide Nuclear Data

The nuclear data of 2 4 1 ~ m , at toin,% r o c r a n , e n r a n r a n and -

v , were evaluated. The evaluation has mainly based on the theo-

re t ica l calculation because the available experimental data a re scarce

except the f iss ion cross section. The recommended value of the

energy range above 1 Kev i s shown in F ig . 5(9). The evaluations of

the resonance parameters of 2 4 1 ~ m , and smooth cross sections and

resonance parameters of 2 4 3 ~ m and 2 4 4 ~ m have been continued.

3 ) Development of Core Analyt ical Method

Evaluation of cross sections of 2 3 5 ~ , 2 3 8 ~ and 2 3 9 ~ u was made

in o rder to obtain a better predict ion of physics parameters f o r large

fast reactors. At f i r s t the sensi t iv i ty of integral propert ies to un-

certaint ies of these cross sections was studied by using the JAERI-

Fast set 1 as a reference group constant set. T o understand general

trends of the sensi t iv i ty, the uniform variat ions of 10% in these cross

sections above 1.0 keV were considered (10). The variat ions in

2350f and 2390f produce uniform changes ( - 5 % ) in kef f ls and the

var iat ion in 238& has dif ferent effects to kef f ls as shown in F ig . 6.

In o rder to obtain more reasonable estimates f o r uncertaint ies of the

cross sections, a stat ist ical analysis (1_1) was made using the data

i n NEUDADA f i le. Based on this analysis, stat ist ical parameters

were calculated f o r the ABBN group structure.

Base on the above studies, evaluation of 2355,, 2395f and 2380r

was performed by the least squares f i t on the effective mult ipl ication

factors, preserv ing ra t ios of dif ferential c ross sections. The

average cross section of 235dt obtained from the stat ist ical analysis

were used as the reference data. F ig . 7 shows the comparison of

the present resu l t f o r 238'3c with the experimental data. Results

f o r f iss ion cross sections of 2 3 5 ~ and 2 3 9 ~ u a re shown in F igs. 8

and 9.

I n o rder to take the mutual and self interference effects into

account, the individual resonance parameters were generated by the

method used f o r the JAERI-Fast set-1. Resolved resonance para-

meters were given based on the BNL-325 (3rd. ed. Vol. 1) and the

ENDF/B-IV f i le . The hyper-fine cross sections were calculated 0

f o r four temperatures 300, 900, 2100 and 3500 K.

The JAERI-Fast set Version II i s recently produced from the

above group constants and other constant which a re produced mainly

f rom ENDF/B-IV.

An performance analysis code system on JOYO (JOYPAC sys-

tem) was developed f o r the purpose of supervising the core nu-

c lear and thermohydraulic behaviors in each operation cycle as well

as f o r checking, and recording of h is tory of burn-up, thermal cycle

and movements on each core subassembly. The code system i s

c lassi f ied into two major par ts such as the simple method and the

detai led method both of which are designated to be used as an off-

l ine computation system.

The simple calculat ion method (SMART) was developed to

calculate the character ist ics of a perturbed core wi th simpli f ied

technique using the resu l ts of the unperturbed core calculated by the

conventional method and data obtained by the character ist ics meas-

urements of JOYO. Th is subsystem i s l inked with a recording sub-

system (MASTOR) to work together.

The detailed calculat ion method (HONEYCOMB) Was

developed to be employed i n e i ther case when there i s any major

change taken place in the core system, o r when needed a detailed

evaluation which can not be made by the simple method. The com-

putable items a re the c r i t i ca l i t y and burn-up of both 2 and 3 dimen-

sions, 3-dimensional r - ray source distr ibution, 3-dimensional

thermal power distr ibution, f low rate distr ibut ion and detailed tem-

perature distr ibut ion i n the designated fuel assembly (FDCAL-3 ) ,

etc. These subsystems can be ei ther separately calculated o r com-

bined a l l together. The code system, JOYPAC, w i l l be used from

the end of th is year when JOYO w i l l reach c r i t i ca l .

In 1971, the accuracies of various physics parameters in

regard to JOYO core were estimated by the extrapolation of the

resu l ts obt3ne-i f rom the mock-up experiments and the i r analyses.

Since 1971, a ser ies of mock-up experiments and analyses f o r

MONJU core, which include the MOZART program as fu l l scale mock-

up experiments and the F C A program as par t ia l one, has been under-

way. Th is evaluation has been aimed to f ind some bases f o r fu r ther

necessary research act iv i t ies in comparison with the requi red accu-

rac ies f o r the core design of MONJU and large fast reactor.

4 ) Research on Shielding

Neutron streaming though the holes penetrating the g r i d plate

shield of MONJU subassemblies was experimentally examined. The

experiments were conducted at the upper column of the fast neutron

source reactor YAYOl (Tokyo Univers i ty ).

The transmitted neutron spectrum was measured with an He-3

counter, while the integral f lux and i t s spatial d istr ibut ion was meas-

u red wi th threshold fo i l s of rhodium and indium as the form of reac-

t ion rate. The resul t reveals that the streaming factor i s about 1 .4.

f o r the ordinary types of coolant passages.

A detailed calculation on the main shield of JOYO was performed

using the two-dimensional transport computer code TWOTRAN-11.

The dose rate around the rotat ing plug obtained by the present analysis

i s estimated to be smaller than that of the previous design calculations.

REFERENCES

) A . OYAMA ,: "Development program of new tpes of power

reactor-fast breeder and heavy water reactor," Intern. Sym. on

Nucl. Power Tech. and Econom., Taipei, Jan. 1975

( 2 ) T . H A W , H. KOSUGl , S. ABE ,: "JOY0 construction and

preoperational test experience", Joint ASME/ANS Intern. Conf.

on Adv. Nucl. Energy Sys., Pittsburgh, March 1976

( 3 ) A. OYAMA, Y. HIGASHIHARA,: "Current status of the fast

reactor physics program in Japan", ANS Intern. Meeting, Washing-

ton, Nov. 1972.

( 4 ) J. HIROTA, H. KUROI, T . IIJIMA, N . MIZOO, K . SHIRAKATA,:

"Recent progress in fast integral experiment and analysis at FCA",

Intern. Sym., Tokyo, Oct. 1973.

( 5 ) S. IIJIMA, A. SHIMIZU, T . INOUE,: "The analysis of FCA

cr i t ica l experiments and i ts applications to JOYO nuclear design1!,

Intern. Sym., Tokyo, Oct. 1973.

( 6 ) T. INOUE, F. YOSHINO, A. IZUMI ,: I1Consideration on safety

margin in JOYO thermal design", IAEA panel meeting, Karlsruhe,

Nov. 1973.

( 7 ) R. MIKI, Y. SUZUKI, Y . NAKAI, Y. NISHIKAWA, K . KAWA-

SHIMA, K . ODAJIMA, H. KITAGAWA,: "Brief description of

planned prototype FBR MONJU of Japano1, BNES Intern. Conf.,

London, March 1974.

( 8 ) C .G. CAMPBELL, J.E. SANDERS, J .L . ROWLANDS,

S. KOBAYASHI ,: "The scope of the MOZART programme and the

general conclusions drawn from it1', Intern. Sym., Tokyo, Oct.

1973.

( 9 ) S. IGARASHI ,: "Evaluation of Am-241 fission cross sections",

IAEA, OECD-NEA Advisory Group Meeting on Transactinium

Isotope Nuclear Data, Karlsruhe, Nov. 1975.

(10) H. TAKANO, A. HASEGAWA, S. KATSURAGI ,: "Proceedings

of AESJ1973 topical meeting of fast reactor physics", A9, 1973

(in Japanese )

(11) A. HASEGAWA, 5. KATSURAGI ,: JAERI-M5536, 1974

(in Japanese ).

Table 1 Design Data f o r JOY0 Core

Power (MW)

Max. Neutron Flux (n/cm2sec)

No. of Core Fuel Assemblies

No. of Blankel Fuel Assemblies

Core Fuel

Pu Content PUOZ/ (PUO~+UO~) (w/o)

U Enrichment u ~ ~ ~ / u (W/O)

Clad Diameter Outer/lnner (mm)

Max. L inear Heat Rate (w/cm)

Max. Burn Up (MWD/T)

CO"t"O1 Rod

No. of Control Rods

Type

Flow Raze (ton/hr)

Reactor Out/ln Temp. ( OC )

MK-I Core

75

3.2 x 1015

68

190

17.7

23

6.3 / 5.6

32 1

50,000

.6

Seal

2200

468 / 370

Table 2 Design Data f o r MONJU Core

Thermal Power

Max. Neutron F lux

Breeding Ratio

No. of Inner Core Assemblies

No. of Outer Core Assemblies

No. of Blanket Assemblies

Core Fuel

P u Content ( l n n e r / ~ u t e r ) ,

In i t ia l Core

Equi l ibr ium Core

Clad Diameter ( ~ u t e r / l n n e r ) , mm

Max. L inear Heat Rate, w/cm

In i t ia l

Equi l ibr ium

F low Rate

Reactor 0u t / ln Temp.

MK-I1 corc -

100

5.0 .. 1015 incl. 6 a s s ' y s

7(for irrad. 1 --

30

12

5.5 / 4.2

37%

60,000

.6

Vent

2200

500 / 370

714 MWt

6.5 x 1015 n/cmzsec

1.22

108

90

172

w/o f i ss i le PU/PU + II 14.9/21 .O

14.8/20.4

6.5/5.56

Table 3 Typical Results of the MOZART Experiment Analysis

1 . C r i t i ca l i t y of MZ-B Core, kef f

C/E = 0.9984 - 1 .OOW

2. Reaction Rate Distr ibut ions

* C/E-values i n MZ-B core

I.C. O.C. Radial Blanket Reaction edge centre inner center outer

28 Chamber 1 .013 1.008 1.036 0.989 1.087 F o i l 1.016 0.991 1 .080 1.065 0.982

F4' Chamber 1.019 1.015 1 .015 0.991 0.908

* Jc/E-1 I values in MZ-C core when control rods a re inserted IC/E-1 1 < 5 % f o r the symmetric insert ion

6 10% f o r the asymmetric insert ion

3 . Sodium Void Worths i n MZ-B Cores

* Radial d istr ibut ion

Ic/E-1 I < 1.0 ? 1.1 % (core centre, plate geometry) < 13.0 f 1 .O% (core centre, p in geometry) < 18.0 f 2.5% (outer core, plate geometry)

, Axia l d istr ibut ion

* Ic/E-1 I i; 1.0 2 1.1'70 (core centre, plate geometry) 48.0 f 4.0% (whole element, plate geometry)

4. Control Rod Worth C/E

I Rod Pat tern case (A)*) Case (6 ) *)

BN (0 ) 0.95 1.051

830 (0 ) 0.96 1 .057

680 (0 ) 1.01 1.082

B90 (0 ) 1.05 1 .072

B80 (Pi ) , B90 (0) 0.96

BN (P2 PhP5 Pb ) 0.94

*) The main difference between A and B i s in the cross section

sets.

LARGE ROTATING PLUG DRIVING UNIT

SAFETY VESSEL

COOLANT INLET-

Fig. 1 VERTICAL CROSS SECTION OF REACTOR A N D CORE

CONFIGURATION OF JOY0

TYPE 3 REFLECTOR

0 CWTROL ROD

FUEL IRRPSI&TlTION ASSEMBLY

MATERIPL IRRPSIATION PSSEMBLY

@ NEUTRON SOURCE

M K - I M K - lI

Fig. 2 CORE MATRIX of MK - I and MK - Il

w

Fig. 3 MONJU REACTOR SYSTEM

The number in each assembly indicates

t h e f low or i f i c ing zone.

F ig . 4 Core C r o s s - s e c t i o n

Energy ( MeV

Fig. 5 Recommended Cross Sections of 241 Am

CHANOEO CROSS SECTIONS

D--R4q

A-Bsl Oi

0-=='u q

Fig. 6 Effect of the variation of 10% in imporiaii?~ cross sections on keff

10 8

6 A Menlove Et Poenitz

4 O.+Fricke et a1 cn z -0- de Saussure et al

% m 2 V H Present results C r )

- * -

z 0 1.0

8

fd 6 u)

cn 4 cn 0 [r 0

2

0.1 1

NEUTRON ENERGY ( KeV 1 Fig. 7

Golrov et al. - - - A Kappeler (70) -*- Blons et al. (70) - + Rjabov (70)

Szabo et al. (71 1 P Perkln et al. (65 ) - K Knoll et a l . (67) - -R Poenitz (70)

- - - Present results ( I ) - --- - * (2)

- - - - -

NEUTRON ENERGY (KeV) Fig. 8

$ Allen et 01. (571 @ Penkin e t 01. (651

'"a, Szobo et 01. (70) s Prince (701

-41- Jomes (70 1 - Farrel et 0 1 . (701 5 Szobo (71 1

- Present results (1)

'/ t 2)

I 1 1 1 1 1 1 1 I I I I 1 1 1 1 1 I I 1 1 1 1 1 1 1

10 to2 lo3 NEUTRON ENERGY ( KeV 1

Fig. 9

REACTOR PHYSICS I N THE UNITED KINGDOM NEACRP-L- (UK) -

C G CAMPBELL J R ASKEW

ADVANCED GAS COOLED REACTORS

1 The two leading s t a t i o n s , Hinkley Point and Hunterston, are current ly working up t o 6W of nominal design output.

2 Analysis of the zero energy experiments undertaken on these r eac to r s before power r a i s i n g has continued, The theore t i ca l s tud ies a re based upon the ARGOSY code with small r e a c t i v i t y correc t ions based upon WIMS comparisons. Especial a t t e n t i o n was focussed on the worth of cont ro l rods both s ingly and i n groups at a va r i e ty of inser t ions . Both c r i t i c a l balances agains t air pressure and t r ans ien t experiments were conducted. B u i l t i n r e a c t i v i t y i n the absence of rods was predicted t o 100 pcm, the difference between reac to r s being 10 pcm. Rod worths a r e predicted within 496, which is comparable with the experimental uncer ta in ty on r e a c t i v i t y scale. Axial reac t ion r a t e p r o f i l e s show systematic tilt discrepancies l e s s than 2%.

3 Testing at power is still i n progress. A t 30% power the r a d i a l va r i a t ion o f r a t i n g over the r eac to r with automatic shaping i n h i b i p d was equivalent t o a 60•‹c range of gas o u t l e t temperature compared t o 50 C predicted.

4 Measurements a r e i n progress on the Hinkley Point B AGR t o determine the f u e l temperature coe f f i c i en t of r e a c t i v i t y and t o monitor its va r i a t ion with burn-up. The f i r s t t e s t s have been made at 18% power during commissioning of the reactor . I n these t e s t s a bank of cont ro l rods was moved out a shor t way ( t o add about 60pcmreact iv i ty) and then re inser ted a f t e r a pause. The neutron f lux and o u t l e t gas temperatures were recorded during the t rans ient . The ana lys i s of f u e l temperature coe f f i c i en t measurements requi res a f a i r l y sophis t ica ted ca lcula t ion of power and f u e l temperature changes during the t rans ient . For t h i s reason the f a u l t study code KINAGRAX is being used f o r the ana lys i s i n which the f u e l coe f f i c i en t w i l l be varied t o give a bes t f i t t o the measured data. Preliminary r e s u l t s of the f i r s t t e s t give temperature c o e f f i c i e n t s which agree with predict ions within the er rors . Errors i n the model and o ther da ta i tems r e f l e c t on the value of temperature coe f f i c i en t derived from t h i s f i t t i n g procedure, so severa l o ther r eac to r t r a n s i e n t s a r e being monitored t o check the adequacy of various aspects of the KINAGRAX modelling and data. These t r a n s i e n t s w i l l include r eac to r t r i p and circula.tor t r i p t e s t s , and the i n i t i a l phase of s tar t-up ( b o i l e r feed t rans ient ) .

HIGH TEMPERATURE REACTOR

5 No agreement was reached on financing a fu r the r extension of the OECD Dragon Projec t at ASE Winfrith beyond March 1976, and the P ro jec t w a ~ terminated. The reac tor has been unloaded, and considerat ion is being given t o the p o s s i b i l i t y of dismantling it. Since its inception i n 1959, the P ro jec t agreement had been extended f i v e times.

- 176 - 6 The col labora t ion exerc ise t o measure the r e a c t i v i t y and i so top ic

composition of i r r a d i a t e d f u e l from the DRAGON r e a c t o r , involving the UKAEA, C U and DRAGON, has continued. The i r r a d i a t e d compacts ( o r i g i n a l l y of 6% enriched U 0 2 coated p a r t i c l e s ) i r r a d i a t e d t o 21,000,

35,000 and 65,000 MWDfie ( 1 .I f i s s i o n s per o r i g i n a l f i s s i l e atom) have been gamma scanned, and o s c i l l a t e d i n the HECTOR r e a c t o r at AEE Winfrith i n two environments; a thermal column of graphi te and a r e l a t i v e l y hard HTR spectrum. The l a s t s e t of samples did not reach the intended i r r a d i a t i o n of 80,000 MWD/Te before the shutdown of the DRAGON r eac to r , and mechanical d i f f i c u l t i e s i n dismantling the f u e l assemblies have precluded measurements on them. Global and l o c d s i g n a l s were measured during osc i l l a t ion . An extensive s e r i e s of c a l i b r a t i o n samples, i n p a r t i c u l a r a s e r i e s of varying Plutonium i so top ic compositions, were a l s o osc i l l a t ed . These measurements a r e now complete and being in terpre ted . The i r r a d i a t e d samples a r e now being subjected t o chemical and i s o t o p i c ana lys i s by CEA, with some confirm&tory a n a l y s i s by AERE Harwell.

7 Following the decis ion t h a t SGHW reac to r s would form t h e next p a r t of the UK nuclear i n s t a l l a t i o n programme, physics measurements i n the ZZNITH reac to r and supporting t h e o r e t i c a l work has ceased. The CEGB's zero energy f a c i l i t y HITHZX a t Berkeley Nuclear Laboratories has continued t o be used t o study the physics of HTR reactors . Zonal r e a c t o r s (annular/ t e l e d i a l geometry and annular / integral geometry) have been b u i l t us ing low enrichment uranium fuel. Measurements have been made of r e a c t i v i t y , power and graphite damage d i s t r i b u t i o n s , r e l a t i v e conversion r a t i o and thermal spectrum; a l l under uniform condit ions, as well a s i n con t ro l rod and gadolimum burnable poison zones. The r e a c t i v i t i e s a r e sys temat ica l ly unprodicted by an amount cons i s t en t with the U K A M ' s Zeni th se r i e s . I n reasonably uniform s tack s i t u a t i o n s , the power d i s t r i b u t i o n s and macroscopic r eac t ion r a t e s i n all energy groups a r e well predicted. However there is evidence t h a t the r e l a t i v e power i n zones of d i f f e r e n t Nc/Nu r a t i o s are underpredicted by up t o s. Adequate modelling techniques have been es tab l i shed t o describe d i s c r e t e burnable poisons i n a v a r i e t y of environmental s i t ua t ions . A l l t h e a l t e r n a t i v e con t ro l rod worths methods t e s t ed have a r a d i a l l y systematic discrepancy, theory-overpredict ing the worth i n r e f l e c t o r adjacent pos i t i ons by up t o 20%. No simple explanat ions have been found t o resolve any of the observed discrepancies. The f a c i l i t y i s t o be used t o car ry out a range of energy and s p a t i a l l y dependent k i n e t i c benchmark measurements t o t e s t a number of a l t e r n a t e codes.

8 The Winfrith Prototype r e a c t o r achieved a load f a c t o r of 92.2% over the Winter peak period from 4 September t o 7 May, 6.6% of the outage being a t t r i b u t e d t o the R & D programme i n support of the commercial reactor . Fuel experience continues t o be good, the leading 36 p i n element i n the r e a c t o r having a t t a ined 20,600 MWD/TeU and the leading 60-pin element 12,800 MWD/T~U.

9 Improvements i n the instrumentat ion and computerized da ta acqu i s i t i on capab i l i t y of the p l an t have continued throughout the year. A WIMS-based l i b r a r y of da t a has been generated f o r the many d i f f e r e n t types of f u e l c l u s t e r s and experimental assemblies i n the r eac to r , and the recomputation of the r e a c t o r h i s t o r y commenced from the s t a r t of l i f e using methods recommended f o r the assessment of the f u l l s c a l e plant.

10 Refinement of the ca lcu la t iona l methods used f o r design has continued, together with fu r the r va l ida t ion agains t experimental data. Monte Carlo s t u d i e s o f a x i a l streaming have resolved the discrepancies i n height c o e f f i c i e n t previously observed, and good agreement has been obtained with experiment i n predic t ing the worth of i n t e r s t i t i a l shutdown tubes. Anomalies of up t o 1016 remain i n the worth of Boron dissolved i n the moderator at high ( ~ 1 0 ppm) concentrations. Analysis of experiments i n the Prototype r eac to r t o confirm the capab i l i ty t o predic t shut-down i n the moderator dumped condition is i n progress.

11 Further development of codes i n the a rea of both sa fe ty and operat ional t r a n s i e n t modelling has continued, i n p a r t i c u l a r centred upon the RELAP and SPLOSH codes. The problem of the appropriate s ingle channel representa t ion of l inked physics and thermal hydraulics t o be used i n the l a t t e r code has been examined. The >dimensional l inked thermal hydraul ics and neutronics code MEKIN is under evaluat ion as a benchmark capabi l i ty .

12 Assessment of predic t ive unce r t a in t i e s i n key s a f e t y computations continues, together with considerat ion of the l o g i c a l b a s i s f o r t h e i r appl ica t ion i n determining operat ional sa fe ty l i m i t s . One aspect of t h i s work is the refinement of co r re l a t ion methods f o r processing da ta from the opera t ional r eac to r and s e t t i n g c r i t e r i a f o r r e j e c t i o n of incons i s t en t measurements.

LIGHT WATER XEACMRS

13 Work has continued t o t e s t the design codes (LWR-WIMS for. l a t t i c e ca lcu la t ions and JOSHUA 111 f o r core ca lcula t ions) on operat ing LWR's.

Most of the work has beep done on Beznau 1 PWR, f o r which the f i r s t four cycles of operat ion have been followed. Apart from the start of cycle 1, the computed r e a c t i v i t e s a r e all about 0.4% s u b c r i t i c a l throughout each cycle. However, subsidary ca lcu la t ions have shown tha t a more accurate representa t ion of the g r i d s would increase the computed r e a c t i v i t y by about O.%, without changing the computed a x i a l power f lux s t ages and t h e i r associated form factors . Further , there a r e indicatbons t h a t the fuel-clad gap temperature r i s e cu r ren t ly assumed t o be 100 C may be too low by a f a c t o r of about 3. Correction f o r t h i s would r a i s e the mean f u e l temperature assumed i n the ca lcu la t ions and hence reduce the computed r eac t iv i ty .

14 The r e a c t i v i t y calculated at the s t a r t of cycle 1 is about 1% super- c r i t i c a l i n the power case. This i s associated with presence of burnable power p i n s which a re grey absorbers. It seems t h a t the e r r o r a r i s e s due t o t h e underestimation of absorpt ions i n the poison, and the problem is cur ren t ly under review.

15 Radial power maps calculated through each cycle using the Albedo matrix opt ion f o r the r a d i a l r e f l e c t o r i n JOSHUA show excel lent agreement (within a few percent) with the experimental maps, channel by channel a x i a l f l u x shapes a l s o show good agreement with experiment through the four cycles.

16 Hot zero power ca lcu la t ions have been performed at the start and end of cycle 1, the s t a r t of cycle 1 r e a c t i v i t y is about 1.3% s u p e r c r i t i c a l , again associated with the presence of burnable poisons. A t both the s t a r t and end of cycle, the boron d i f f e r e n t i a l r e a c t i v i t y worth is underestimated by about I%, while the t o t a l worth of the cont ro l rods is a l s o underestimated by about 1%. The value of end of cycle moderator temperature coe f f i c i en t is overestimated by 10%.

17 An e x t r a option has been programmed i n t o JOSIIUA I11 t o use the CISE s l i p r a t i o . This has been tes ted i n Monticello BWR t o compare i t with the usual Bancoff-Jones s l i p option. The new option increase the voidage i n the core and hence reduces the r e a c t i v i t y s a t i s f a c t o r i l y (from 1.008 t o 1.001 typica l ly) . Use of the subcooled bo i l ing option a l s o s l i g h t l y increases the voidage, y ie ld ing a decrease of about 0.002 i n r eac t iv i ty .

Current work on JOSHUA I11 is designed t o improve and speed up the convergence of subsequent burnup s t e p s a f t e r r a d i c a l changes i n rod pa t te rn ing ( p a r t i c u l a r l y re levant t o BWR1s).

NUCLEAR DATA

Request L i s t

18 The current UK nuclear data reques ts list was i ssued i n August 1975 (~E&-~1369) and was summarised i n t h e last r epor t t o NEACRP,

Progress with ac t in ide da ta

Measurements a r e i n progress t o improve t h e f i s s i o n cross-section da ta of U235, U238, Pu239 and Am241. Measurements of i n e l a s t i c s c a t t e r i n g da ta f o r U238 a r e continuingoand t h e measurements of U238 resonance da ta a t temperatures up t o 2000 C should commence l a t e r t h i s year.

Exis t ing experimental d i f f e r e n t i a l d a t a f o r t h e higher ac t in ides a r e very l imi ted , and t h e p o s s i b i l i t y of ca l cu la t ing t h e cross-sect ions from nuclear theory has been examined. The work is reported i n AERE-R7468. These methods a r e being applied t o Am241, 242111 and 243. I n t h e case of Am241, t h e f i s s i o n da ta below 50 keV a r e high d iscrepant , which i n turn a f f e c t s t h e ca lcula t ion of capture cross-sections. The most l i k e l y behaviour of t h e f i s s i o n and capture cross-sect ions has been chosen by r e f e r r i n g t o t h e ZEBRA i n t e g r a l data.

Recent capture cross-section da ta f o r U238 ded ced from t h i n sample a c t i v a t i o n measurements on IBIS (sample thickness 6 x lo-! atoms/barn) tend t o be lower i n the energy region above 500 keV than t h e evaluated da ta of Sowerby e t al. (Annals of Nuclear Science and Engineering 1 (19741, 409). Previous measurements used th i cke r samples andinvolvedthe d i f f i c u l t cor rec t ions f o r mul t ip le i n t e r a c t i o n and i n e l a s t i c s ca t t e r ing .

The f a s t neutron f i s s i o n spec t ra f o r U235 and Pu239 from Cadarache, Geel, Harwell (U235 only) and Studsvik have been sys temat ica l ly intercompared a s a r e s u l t of t h e S p e c i a l i s t Meeting at Harwell, 14-16 Apri l 1975. The da ta have been intercompared, a f t e r cor rec t ion , fo r sample s i z e e f f e c t s , on t h e b a s i s of t h e Watt formalism which produced prefer red f i t s when compared with Maxwell d i s t r ibu t ions . The sample s i z e cor rec t ions tended t o harden t h e f i s s i o n spectra . Agreement i s reasonable i n t h e energy range 1-10 MeV, but t h e r e is some evidence t o suggest a hardening of both spec t r a with increas ing neutron energy. There is i n s u f f i c i e n t information on t h e shape of t h e f i s s i o n spectrum below 0.5 MeV and considerat ion is being given t o extending t h e range t o lower energies , because of t h e pe r s i s t en t tendency t o underpredict f a s t r eac to r spec t r a a t lower energies. The weighted mean of t h e Watt formalism parameters r e s u l t s i n an average f i s s i o n neutron energy, E, of 2.016 MeV and 2.097 MeV f o r U235 and Pu239 respect ively. The corresponding ENDFB-IV values a r e 1.985 MeV and 2.06 MeV. A complete evaluat ion of a l l ava i l ab le f i s s i o n neutron s p e c t r a l 3ata w i l l be made.

Capture cross-sect ions f o r s t r u c t u r a l ma te r i a l s

23 Measurements of t h e capture cross-sect ions of Fe, N i and C r a r e i n progress on t h e Harwell LINAC, together with t o t a l cross-sect ion measurements on a small-sample f l i g h t path s e t up on t h e synchrocyclotron.

FAST REACTOR PHYSICS

ZEBRA programme

24 The programme of i n t e g r a l da ta s t u d i e s described i n l a s t y e a r ' s repor t was completed i n May 1975. Analysis of t hese experiments, and of work on t h e PFR mock-up core, have l e d t o t h e following conclusions:

86020 170

i Sodium removal experiments with pin-geometry f u e l have so f a r been wel l predicted, and have provided va l ida t ion f o r ca l cu la t ions of voiding indiv idual assemblies. Larger s c a l e sodium voiding has usual ly been accomplished with p l a t e geometry f u e l f o r which t h e leakage components a r e not well ca lcu la ted ; e r r o r s of 15% t o 2096 have occurred. There is a need f o r an improved treatment of leakage i n plate-geometry c e l l s .

ii The i n t e g r a l cross-section of Am241 leading t o t h e bu i ld up of Cm242 has now been measured i n d i f f e ren t spec t r a i n ZEBRA assemblies, and has cons is ten t ly shown t h a t t h e C/E values (using FD5 da ta) i s around 0.64.

iii A f u r t h e r check on t h e i n t e g r a l da ta f o r Am241 capture and f i s s i o n and t h e i n t e g r a l Pu241 f i s s i o n da ta measured i n ZEBRA has been provided from t h e r e a c t i v i t y l o s s i n ZEBRA during a 3 month shut-down. A C/E value of 0.89 + -10 (1s.d.) was found, r a t h e r than a C/E of 0.80 using t h e FD5 values-for Pu241 and Am241 data.

i v I n ZEBRA Core 13 comparative measurements of f i s s i o n r a t e s were made i n t h r e e versions of t h e th i ck r a d i a l breeder sec to r viz. U02/Na, UO/Pu02/Na and U/C/Na. Comparisons with d i f fus ion theory ca l cu la t ions have shown t h a t :

i t h e C/E f o r t o t a l f i s s i o n r a t e tends t o f a l l from about 1.08 neai- t o t h e core/breeder i n t e r f a c e t o 0.98 deep i n t o t h e breeder;

ii i n t h e case of t h e breeder with Pu02 present t h e a t tenuat ion of t h e U238 f i s s i o n r a t e and t h e fa l l ing-of f of t h e C/E va lues is considerably reduced.

v The decay of t o t a l f -energy from Pu239 and U235 f i s s i o n has now been measured from 20 s e c s t o 18 months a f t e r a n i r r a d i a t i o n of about 1 day i n ZEBRA. A preliminary comparison with predic t ions based on Devi l le rs ' and Tobias' da ta show increas ing discrepancies a t cool ing times exceeding 1 month, but fu r the r work is needed before drawing d e f i n i t i v e conclusions.

v i I n t e g r a l measurements of t h e ac t iva t ion of f a s t r eac to r primary c i r c u i t mater ia l s and corrosion products have confirmed t h a t cur rent data a r e general ly adequate, predic t ions agreeing t o within about + 35%. These experiments provide improved one- group da ta should i t be needed, and have emphasised t h e need f o r care i n allowing f o r resonance se l f - sh ie ld ing i n Co-59 - present i n s t e l l i t e c i r c u i t components.

25 A co l labora t ive agreement has been signed with GfK, Karlsruhe ( t h e BIZET p ro jec t ) which provides f o r t h e pooling of UK and German zero power f a s t ' r eac to r f u e l t o allow t h e study of physics and s a f e t y problems i n assemblies i n ZEBRA approaching the s i z e appropriate t o a 1250 MW(E) f a s t reac tor . The main aims of t h e programme a r e t o study:

i t h e e f f e c t s of sodium removal on r e a c t i v i t y

ii t h e d i s t r i b u t i o n of power i n t h e cores and surrounding regions

iii t h e ef fec t iveness and in t e rac t ion of cont ro l rods.

26 The ZEBRA reac to r was shut-down.in May 1975 f o r t h e modifications required t o accommodate cores of t h i s s i ze . The work has kept well t o programme, and reloading is planned t o s t a r t i n August next.

27 The f i r s t core t o be b u i l t w i l l be of a conventional 2-enrichment zone design and t h i s assembly should be c r i t i c a l by the l a t e Autumn 1976.

28 Studies a r e i n progress of t h e proper t ies of l a r g e f a s t r eac to r s of heterogeneous design, which show a s i g n i f i c a n t l y lower r e a c t i v i t y gain assoc ia ted with l o s s of sodium. This a r i s e s mainly because of t h e increased leakage from t h e f i s s i l e t o t h e f e r t i l e zones of t h e core. Other advantages claimed a r e improved breeding c h a r a c t e r i s t i c s and reduced r e a c t i v i t y investment i n con t ro l rods. Such designs apparently have smaller Doppler c o e f f i c i e n t s than conventional cores.

29 The design f o r a heterogeneous assembly f o r study i n ZEBRA, which allows experimental confirmation of t h e impbrtant physics predic t ions c h a r a c t e r i s t i c of a range of poss ib le heterogeneous designs, is being discussed with GfK. The durat ion of t h e ZEBRA programme on conventional core problems and plans f o r changing t h e core i n ZEBRA t o a heterogeneous arrangement a r e under a c t i v e discussion.

I r r a d i a t e d Fuel Exueriments i n HECTOR

30 The programme of o s c i l l a t i o n experiments i n HECTOR on coated-part icle f u e l compacts i r r a d i a t e d i n DRAGON has been completed. I n t h e event t h e maximum i r r a d i a t i o n achieved was about 65 GWD/Te, lower than t h a t o r i g i n a l l y an t i c ipa ted due t o t h e shut-down of DRAGON. The HECTOR measurements included both "global" and "local" s i g n a l s i n t h i s test-region spec t r a (thermal and undermoderated HTR). Complementary experiments were made on a range of c a l i b r a t i o n samples, and a l s o on uni r rad ia ted Fu02/U02 samples (with Pu240/Fu up t o 45%) loaned by CEA. Key reac t ion r a t e s ( including U238 capture) were a l s o measured. The f i n a l s t age i n t h i s experimental programme is t h e chemical and i so top ic ana lys i s of se lec ted i r r a d i a t e d compacts t o be ca r r i ed out over the next few months by CEA. The WIMS-HPS i n t e r p r e t a t i o n scheme has been successfu l ly t e s t e d on both global and l o c a l s i g n a l s from uni r radia ted samples.

WUNREAY FAST REACTOR

31 The f i r s t of a s e r i e s of experiments t o inves t iga t e t h e e f f e c t of reduced cooling and coolant boi l ing was ca r r i ed out successful ly.

PROTOTYPE FAST REACTOR

32 The r e a c t o r has operated at high a v a i l a b i l i t y a t powers up t o 200 MW(T) with e l e c t r i c a l power generat ion up t o 40 MW. Steam has been generated throughout t h e period on one secondary c i r c u i t bu t , i n t h e others, t h e r e has been some generat ion in t e r rup ted by small leaks.

Reactor Physics Division AEE Winfrith

18 June, 1976

R e a c t o r P h y s i c s A c t i v i t i e s i n S w i t z e r l a n d

J u n e 1 9 7 5 t o May 1 9 7 6

R . Richmond

.I. E x p e r i m e n t a l Measurements

E x p e r i m e n t s on t h e e f f e c t o f s t e a m e n t r y i n t o a GCFR l a t t i c e

were c o n t i n u e d i n t h e z e r o e n e r g y r e a c t o r PROTEUS. The s i m u l a t i o n

o f s t e a m e n t r y was a c h i e v e d by f i l l i n g t h e v o i d s p a c e o f a c e n t r a l

t e s t z o n e o f t h e r e a c t o r w i t h expanded p o l y s t y r e n e b e a d s g i v i n g 3

an e q u i v a l e n t c o o l a n t c h a n n e l s t e a m d e n s i t y of 0 .046 g/cm . P . s t h e l a t t i c e h a s h e x a g o n a l r o d d e d g e o m e t r y , t h e f i n a l c o n f i g u -

r a t i o n m o d e l s t h e power r e a c t o r c a s e w e l l

The h e x a g o n a l c e n t r a l t e s t z o n e wh ich i s 13 cm a c r o s s t h e f l a t s was

s u r r o u n d e d s u c c e s s i v e l y by 3 d i f f e r e n t f a s t l a t t i c e e n v i r o n m e n t s .

T h e s e were \

( a ) a t y p i c a l f a s t l a t t i c e ( " c l e a n d r y c a s e " )

( b ) a t y p i c a l f a s t l a t t i c e t h r o u g h o u t which s t e a m s i m u l a n t i s

d i s t r i b u t e d t o g i v e t h e same smeared d e n s i t y a s t h a t i n

t h e t e s t column ( " c l e a n s t e a m c a s e " )

( c l a t y p i c a l f a s t l a t t i c e w i t h s t e a m s i m u l a n t and w i t h

d i s t r i b u t e d boron r o d s t o c h e c k t h e i n f l u e n c e o f a llv a b s o r b e r

on t h e s t e a m e n t r y i n c i d e n t ( " p o i s o n e d s t e a m c a s e " ) .

Measurements were made o f c e n t r a l n e u t r o n s p e c t r u m and r e a c t i o n

r a t e r a t i o s and o f t h e r e a c t i v i t y w o r t h o f hydrogen i n a c e n t r a l

r e g i o n o f t h e l a t t i c e . The a v a i l a b i l i t y o f v a r i e d e x p e r i m e n t a l

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

which t o a s s e s s t h e a d e q u a c y o f t h e c a l c u l a t i o n a l me thods and

d a t a .

The n e u t r o n s p e c t r u m measurmements , which c o v e r e d t h e e n e r g y

r a n g e f r o m 9 keV t o 2 .2 MeV, were p e r f o r m e d o n l y i n t h e p o i s o n e d

s t e a m l a t t i c e . The s p e c t r u m c h a n g e s p roduced by s t e a m e n t r y were

a c c u r a t e l y p r e d i c t e d by c a l c u l a t i o n s b a s e d on t h e U K FGL4 d a t a

s e t .

Good a g r e e m e n t be tween c a l c u l a t e d and e x p e r i m e n t a l v a l u e s was

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

FGL4 d a t a i n one d i m e n s i o n a l t r a n s p o n t h e o r y c a l c u l a t i o n s . I n

p a r t i c u l a r , c h a n g e s i n t h e i m p o r t a n t r a t i o o f U238 c a p t u r e t o

Pu239 f i s s i o n , which c o v e r s a b o u t 70 % o f t h e t o t a l r e a c t i o n s

i n t h e l a t t i c e , were w e l l p r e d i c t e d .

The r e a c t i v i t y measuremen t s i n d i c a t e d t h a t t h e a d d i t i o n o f

hydrogen g a v e a p o s i t i v e r e a c t i v i t y c h a n g e f o r b o t h t h e c l e a n

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

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

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

r e l a t i v e l y l a r g e c h a n g e s i n n e u t r o n l e a k a g e , p r o d u c t i o n , a b s o r p t i o n ,

and m o d e r a t i o n . The s e n s i t i v i t y o f t h e c a l c u l a t e d FGL4 r e s u l t s

t o t h e c a l c u l a t i o n a l p r o c e d u r e s i s s t i l l b e i n g examined , b u t t h e

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

compare w i t h t h e e x p e r i m e n t a l r e s u l t s a s f o l l o w s .

c l e a n d r y c a s e 0.44 0.02 0.16

c l e a n s t e a m c a s e 1 . 4 0 ? 0 .04 2 . 6

p o i s o n e d s t e a m c a s e -2 .47 '! 0.08 - 0 . 8

A s a c h e c k on t h e s y s t e m n o r m a l i z a t i o n i n t e g r a l , e x p e r i m e n t s

and c a l c u l a t i o n s have a l s o been c a r r i e d o u t t o d e t e r m i n e t h e

w o r t h o f d i s t r i b u t e d boron a b s o r b e r i n t h e t e s t r e g i o n . The

measured and c a l c u l a t e d v a l u e s a g r e e d t o w i t h i n a b o u t 10 %.

The c o n c l u s i o n t h a t hydrogen a t t h e c e n t r e o f a c l e a n , c o l d f a s t

r e a c t o r l a t t i c e h a s a p o s i t i v e r e a c t i v i t y w o r t h i s i n a g r e e m e n t

w i t h t h e r e s u l t s o f e x ~ e r i m e n t s i n o t h e r c r i t i c a l a s s e m b l i e s .

P e n d i n g t h e f i n a l a n a l y s i s o f t h e e x p e r i m e n t a l r e s u l t s a n a t t e m p t

was made t o a s s e s s t h e e r r o r s a r i s i n g f r o m d a t a u n c e r t a i n t i e s i n

c a l c u l a t i n g t h e e f f e c t o f s t e a m e n t r y i n an o p e r a t i n g GCFR. T h i s

seemed t o be o f i n t e r e s t s i n c e p u b l i s h e d work on s t e a m e n t r y h a s

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

i n t h e c a l c u l a t i o n s . I n t h e a b s e n c e o f a c u r r e n t c a p a b i l i t y f o r

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

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

t h e d i f f e r e n c e s i n t h e p r e d i c t i o n s g i v e n by d i f f e r e n t d a t a s e t s .

N u c l i d e c o m p o s i t i o n s f o r t h e t e s t c a l c u l a t i o n s were t h e s e o f t h e

p h a s e I A N L GCFR c r i t i c a l a s s e m b l y which mode l s t h e p r o p o s e d G A C

300 MW[el demo p l a n t power r e a c t o r . FGL4 and FGL5 d a t a were u s e d I

w i t h a s i m p l e homogeneous o n e - z o n e model s i n c e t h i s i s a d e q u a t e

f o r c o m p a r i s o n p u r p o s e s . The d i f f e r e n c e s be tween t h e p r e d i c t e d

r e a c t i v i t y c h a n g e s on s t e a m e n t r y depend on t h e s t e a m d e n s i t y and

on t h e c o n d i t i o n o f t h e r e a c t o r . A s a n example , f o r t h e c o l d , 3 .

c l e a n r e a c t o r w i t h a s t e a m d e n s i t y of 0 .015 g/cm i n t h e c o o l a n t

c h a n n e l s t h e r e a c t i v i t y c h a n g e s p r e d i c t e d by FGL4 and FGLS a r e

+ $ 2 . 5 and + $ 3 . 5 r e s p e c t i v e l y . A s i m i l a r c o m p a r i s o n f o r ENDF/B

d a t a e x t r a c t e d f r o m q u o t e d G A C r e s u l t s f o r t h e c o l d , c l e a n un- 3

r o d d e d r e a c t o r shows t h a t , f o r a s t e a m d e n s i t y o f 0 . 0 1 5 g/cm , t h e

p r e d i c t e d r e a c t i v i t y c h a n g e s a r e -$2 .6 and $ 0 f o r ENDF/B-3 and 4

r e s p e c t i v e l y .

A f u r t h e r i n d i c a t i o n o f t h e e r r o r s was o b t a i n e d b y c o n s i d e r i n g

t h e i n d i v i d u a l e r r o r s on t h e r o u g h l y e q u a l and o p p o s i t e t e r m s

mak ing up t h e r e a c t i v i t y change ( i . e . t h e n e g a t i v e e f f e c t o f t h e

change i n k - i n f i n i t y and t h e p o s i t i v e e f f e c t o f t h e r e d u c t i o n

i n l e a k a g e ) . We assumed i n d i v i d u a l e r r o r s o f ' 5 % on t h e s e te rms .

The FGL5 p r e d i c t i o n s o f t h e r e a c t i v i t y change i n t h e h o t , p o i s o n e d 3

r e a c t o r a t a s team d e n s i t y o f O.OISg/cm were t h e n $-1.5z2.1.

V a l u a b l e a d d i t i o n a l i n f o r m a t i o n on t h e s team e n t r y i n c i d e n t w i l l

be g i v e n by t h e c u r r e n t ANL measurements and a KFK p r o p o s a l f o r

a benchmark c a l c u l a t i o n e x e r c i s e s h o u l d a l s o l e a d t o u s e f u l

r e s u l t s .

Meanwh i l e t h e EIR r e s u l t s s u g g e s t t h a t t h e m a g n i t u d e \ ( a n d even t h e

s i g n ) o f t h e r e a c t i v i t y change p roduced b y s team e n t r y i n t o an

o p e r a t i n g GCFR i s s t i l l an open q u e s t i o n . I n t h i s s i t u a t i o n t h e

s a f e t y assessment c o u l d w e l l be based on d e t a i l e d c o n s i d e r a t i i o n s

of t h e maximum r a t e and amount of s team e n t r y r a t h e r t h a n on t h e

a s s u m p t i o n t h a t t h e r e a c t i v i t y change can be shown t o be n e g a t i v e

i n a l l c i r c u m s t a n c e s .

2. C a l c u l a t i o n Methods

The r e s u l t s o f t h e I D t r a n s p o r t benchmark c a l c u l a t i o n s u s i n g

SHADOK and SHCRLDCK w h i c h were c a r r i e d o u t i n t h e f ramework o f

t h e IAEA C o - o r d i n a t e d Research Programme were compared a t a

m e e t i n g i n Bo logna (Dec.1975) , w i t h t h e c o r r e s p o n d i n g r e s u l t s

o b t a i n e d i n F r a n c e (COLINE), I t a l y and Y u g o s l a v i a . The e i g e n -

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

w i t h i n compu te r a c c u r a c y . A t t e n t i o n i s now b e i n g t u r n e d t o 20

met hods.

A t t h e same m e e t i n g a d e s c r i p t i o n was g i v e n o f a n e w l y d e v e l o p e d

2D c o l l i s i o n p r o b a b i l i t y c o d e ( Q P I ) i n wh ich a q u a d r u p o l e P I

a p p r o x i m a t i o n i s used . T h i s i s p a r t i c u l a r l y s u i t a b l e f o r LWR box

c a l c u l a t i o n s and h a s been i n t r o d u c e d a s a module i n t o t h e B O X E R

c a l c u l a t i o n scheme.

A 20 f i n i t e e l e m e n t code ( F I N E L ) w i t h ( x , y ) g e o m e t r y h a s been

p roduced and t e s t e d i n a o n e - g r o u p v e r s i o n . F o r a s o u r c e c a s e

w i t h l a r g e homogeneous z o n e s t h i s c o d e i s t e n t i m e s f a s t e r t h a n

a t y p i c a l f i n i t e d i f f e r e n c e c o d e ( e . g . 2 0 6 ) . A m u l t i g r o u p v e r s i o n

o f t h i s c o d e i s now i n p r e p a r a t i o n and t h e u l t i m a t e a im i s t o

~ r o d u c e a 3D v e r s i o n .

Work on m i c r o - f i s s i o n c h a i n r e a c t i o n s h a s i n c l u d e d s t u d i e s u s i n g

t h e p o i n t k i n e t i c s model . The k i n e t i c e q u a t i o n s w i t h s o u r c e

n e u t r o n s ( i n t h i s c a s e d u e t o f u s i o n 1 n o r m a l l y r e q u i r e a knowledge

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

of t h e a d j o i n t f l u x e s f o r t h e r e f e r e n c e c r i t i c a l s y s t e m . I t h a s

been shown t h a t t h i s c a n be a v o i d e d and t h e r e q u i r e d i n t e g r a l s

o b t a i n e d d i r e c t l y f r o m t h e normal e i g e n v a l u e c a l c u l a t i o n s . I f t h e

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

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

\ t h o s e f r o m t h e f i r s t two o r t h r e e i t e r a t i o n s ) g i v e t h e r e q u i r e d

i n f o r m a t i o n .

STATUS REPORT TO NEACRP (1975 - 1976) NORWAY

REACTOR PHYSICS ACTIVITIES I N NORWAY - June 1975 - May 1976

T. Skardhamar

LWR METHODS DEVELOPMENT AND APPLICATION

The modular code system FMS ( ~ u e l Management System) i s increasingly

being applied by d i f fe ren t power u t i l i t i e s fo r calculat ions on core

management and operation guidance on BWR cores. The accumulated

experiences of these applicat ions of t h e code system is very posi t ive.

During t h e period covering t h i s s t a t u s report the main theore t ica l

developments of the system have been concentrated on implementation

and evaluation of features necessary for PWR applications.

The f u e l assembly burnup code RECORD has been extended t o include a l so

accurate treatments of rod c l u s t e r control absorbers (both Ag-In-Cd and

boron based rods) , and burnable absorber (boron) shim rods. Evaluation

of these models have shown very sa t i s fac to ry r e s u l t s with respect t o

ca lcula t ion of RCC r e a c t i v i t y e f fec t s and shim rod burnup.

Core- follow s tudies have now been performed f o r two operating cycles

of the bf i leberg BWR whose core includes a l s o gadolinium containing

fuel . .Very sa t i s fac to ry r e s u l t s have been obtained as regards

predict ion of r eac t iv i ty , power d is t r ibut ions and control rod e f f e c t s

during burnup. Analysis has a l so been made of one of t h e gadolinium

containing f i e 1 assemblies which has been y-scanned a f t e r burnup. These

r e s u l t s show t h a t t h e l o c a l power d i s t r ibu t ion i n such an assembly i s

wel l predicted by RECORD a l s o during burnup.

The 3-D simulator PRFSTO has been modified with respect t o PWR applicat ions.

Simulation of PWR cores, using l a t t i c e da ta from RECORD, appear t o give

as good r e s u l t s as f o r BWR cores. Further general developments of both

hydraulics and neutronics model i n PRESTO a r e a l so going On.

A new 3-D reac tor dynamics code, RAMONA 111, i s being t e s t e d against

experimental data. Preliminary r e s u l t s a r e sa t i s fac to ry .

A f u e l f a i l u r e probabi l i ty model using load power increments ( f u e l duty

cycling ) as input has been coupled t o t h e power d i s t r i b u t i o n modules of

FMS. Localized f u e l f a i l u r e p robab i l i t i e s may thus be Calculated f o r

a given sequence of reac tor operating data. The model is now being

t e s t e d against f a i l u r e records fo r one BWR and one PWR. Scandpower A/S

of Norway i s responsible f o r t h i s projec t .

Studsvik -

Reactor Physics in Sweden July 1975 - June 1976

Introduction

The theoretical activity in the thermal reactor physics field

is almost entirely oriented towards problems connected with

power reactor design and operation. The experience from the

operation of the LWR reactors in Sweden indicates that existing

code systems work reasonably well; efforts to integrate differ-

ent codes, to make them efficient and more user oriented are

of more immediate interest than new theoretical developments.

On the experimental side critical experiments have been per-

formed in the KRITZ facility and at the Oskarshmn I reactor

irradiated fuel has been gama scanned to determine the power

distribution at the time period close to reactor shut down.

Additional experiments at other reactors are planned.

The activity in the fast reactor field continues to be low.

Thermal reactor physics

Theoretical ------

Through a cooperative arrangement with the State Power Board

and with the private utility groups in Sweden work is con-

tinuously going on at AB Atomenergi to further develop and

improve the cell and assembly code CASMO.

CASMO has been thoroughly tested both against results from

critical facilities - in particular from the Studsvik KRITZ facility - and in power reactor calculations. The agreement with the experimental information is very satisfactory.

The latest improvement of the code concerns the treatment of

boron glass rods and rods containing burnable absorbers.

'The reactivity effect of such rods in a bundle is overesti-

%6 611 01 NYKUPING 1 aterg 64013 SWEDEN 0165 - 800 W nykoping aterg s 33-0067

AKTIEBOLAGET ATOMENERGI

mated in the calculations even when transport theory is used

in the 2D part. A correction has to be introduced that

accounts for the difference between a homogenized pin cell

treatment and a true heterogeneous one.

The work on the large-mesh BITNOD code for reactor core cal-

culations continues as an AB Atomenergi - Swedish State Power Board joint project. The first goal of the work is to

make a 3D PWR version of BITNOD. Although of nodal type the g? code will allow the calculation of,smooth power distribution

in each fuel assembly. As BITNOD is intended for the planning

and following of reactor operation, routines for thermo-

hydraulics, fuel temperature, xenon and burnup are included.

Considerable knowledge is available today about desirable

characteristics of production oriented codes for power reactors.

As BITNOD is created from scratch it will be possible to

efficiently utilize this knowledge and experience in the code.

In fall 1974 some work was made on plutonium recycling and

new studies have recently started. Various design and re-

cycling alternatives, the build up of transuranium elements

etc will be looked into.

Experiments - - - - - During the latter part of 1975 KRITZ was used for an extensive

series of measurements on BWR bundles containing burnable

absorber (BA) rods in various combinations. The main purpose

of the measurements was an accurate determination of the

reactivity effect of BA rods, and of the fission rate distri-

bution inside the bundles. Measurements were performed in 0

the temperature range 20 OC to 245 C. The experiments have

been analyzed in great detail and serve as benchmarks for

the codes in use in Sweden.

No new measurements are foreseen for KRITZ before the end

of 1977.


Decay he% - -

Three sets o f experiments are un lder way in Studsvik a

aiming at improving the accuracy of available data on

fission product (FP) decay heat.

At the Studsvik van de Graaff machine a radiometric method

is used to study the decay of FP from small irradiated

uranium and plutonium specimens. Measurements are in progress

to determine the residual gama radiation from thermal fission

of U-235 over the time interval 10 sec to 35 min after

fission.

The samples are irradiated in a special facility and trans-

ported to the gamma spectrometer by means of a pneumatic

system.

The absolute number of fissions in the sample is determined

by three independent methods: a) by utilizing an absolute

calibrated fission chamber with an active volume of the same

size as the samples, b) by counting the gamma rays emitted

from fission products with well known yields and decay pro-

perties, c) by comparison of the g m a ray yield of uranium

samples irradiated by the accelerator and in the R2 reactor.

The neutron spectrum in the chosen reactor position is well

thermalized and can be determined with high accuracy.

For decay heat determination the gamma radiation from the

fission products is measured with a well-shielded and

collimated NaI(T1) scintillator of diameter and length

12.5 cm. A 4096 channel analyzer is used for recording

the spectra. Sample transportation, irradiation and counting

times are handled by a PDP-15 computer which is also used

for recording the gamna ray spectra. Spectra are auto-

matically stored on magnetic tape for off-line data analysis,

i.e. the transformation from complex pulse height spectra

to energy spectra.


The accuracy of the determination of the total gama energy

is expected to be about 10 96 in the time interval a few

seconds to 30 minutes after fission. Plans are under way

also to include beta decay energy measurements.

The calorimetric measurements briefly mentioned in the pro-

gress report for the last meeting have been further delayed

by the activity at the KRITZ facility. All effort hitherto

has been concentrated on the construction of a calorimeter

with a short time constant and actual measurements are not

expected to start until the end of 1976.

The third set of experiment related to decay heat is part of

a large program for studying the decay properties of fission

products. Irradiated U-235 samples are used in the ion source

of a mass spectrometer allowing individual fission products

to be isolated and measured. The average beta decay energy

for most of the elements in the light mass part of the yield

curve has been determined and measurements are in progress

for the heavy mass part. In a later phase of the program the

average gamma energies will be determined in a similar fashion

The time resolution is from a few seconds and upwards.

The blanket experiments performed in the FRO reactor several

years ago have been finally analyzed and reported. Besides

that the reactor physics activity in the fast reactor field

has been very low during the reporting period.

Only a small effort related to the neutronic aspects of

fusion reactors is made in Sweden. The problems dealt with

at present are connected with the design of the "SPHERATOR"

fusion system developed by Lehnert et al. at the Royal


Institute of Technology in Sweden. This system calls for a

toroidal conductor inside a spherical vacuum chamber. The

neutron and gamma energy deposition in this conductor is I

being studied under various assumption using ANISN code for

determining the neutron and gama fluxes.

Power reactors ------- Five LWR power reactors are now in operation in Sweden (four

ASEA-ATOM BWRs and one Westinghouse PWR). A sixth BWR will start

operation in fall 1976. Six additional reactors have been

ordered.

A joint Swedish-Finnish project has been organized to propose

a conceptual design for a low temperature power reactor for

district heating purposes.

NEACRP L- United S ta tes

Reactor Physics Ac t iv i t i e s i n the U. S. A Report t o the NEACRP

Chalk River, Canada, June 21-25, 1976

P. B. Hermnig, J. W. Lewellen and V. W. Lowery

Introduction

Major a c t i v i t i e s i n the preceding year have centered about the c r i t i c a l

eKperimentS i n support of CRBR and GCFR systems using the ZPPR and ZPR-9

f a c i l i t i e s respectively. Highlights of these programs include completion

of t he reference design EMC, s t ud i e s of the r eac t i v i t y changes due t o off-

normal configurations and s tud ies of the source l eve l f l ux monitor response

f o r CRBR. Measurements of steam entry worths, Be worths and neutron stream-

ing e f f e c t s i n the GCFR provided s ign i f i can t ins igh ts i n t o these phenomena.

The i n s t a l l a t i o n of a l a rge r matrix and f u e l loading machine i n ZPPR w i l l

begin i n January 1977.

A cooperative ERDA-EPRI program was i n i t i a t e d f o r the design of a

Prototype Large Breeder Reactor (PLBR). The par t ic ipa t ing designer-

a r c h i t e c t engineering teams a r e AI-Burns and Roe, GE-Bechtel and w- Stone and Webster.

System optimization s tud ies which have been carr ied out d~ ANL, GE,

HHlL and CE have involved a var ie ty of reactor physics calculat ions . These

have resu l ted i n optimum designs f o r f u e l elements consis tent with goals of

higher breeding gain, improved burnup and improved f u e l r e l i a b i l i t y .

Sens i t i v i t y s tud ies a t ORNL and ANL def ine t he dependence of various

reac tor parameters on the accuracy of d i f f e r e n t i a l cross sec t ion data. They

also def ine cos t benef i t s of improving d i f f e r e n t i a l da ta andlor carrying

ou t add i t iona l c r i t i c a l experiments.

The ENDFIB-IV da ta f i l e has been extensively tes ted and is now processed

f o r rou t ine design use. Development of ENDFIB-v was i n i t i a t e d . ENDFIB-V

is expected t o provide major improvements i n the ac t in ide , f i s s i o n product,

and e r r o r f i l e s .

ZPR-9 - The GCFR benchmark is a t h r ee phase program. Phases 1 and 2 were both

homogeneous single-composition cores with a 53% and 43% void volume f r ac t i on

(coolant volume f r ac t i on i n a GCFR) respectively. Phase 3 i s a

core with composition s imi la r t o previous experiments. Phase 1

March 1975, followed by Phase 2 i n June 1975 and Phase 3 i n May

th ree zone

was begun i n

1976. \The

program wil !1 be complete d a t the end of September 76 and w i l l be followed

by the Advanced Fuel Comparison Cr i t i ca l s . These experiments a r e current ly

being planned and w i l l begin with a carbide benchmark.

Important measurements made i n Phase 2 included steam entry e f f ec t s ,

helium coolant r eac t iv i ty coef f ic ien ts , and cen t r a l conversion r a t i o . The

helium worth experiments were conducted by measuring the worth of the sample

cylinder (pressurized t o 150 and 300 psia) r e l a t i v e t o reference a i r - f i l l e d

cylinders by a l te rna t ive ly posit ioning the sample and reference cylinder a t

the core center. The measurement He worth was 1.5 Ih/gm and the C/E

corrected f o r f l ux d i s to r t i on ranged from 1.31 t o 1.44. The r eac t iv i ty

e f f e c t of whole core steam entry was studied using polyethylene foam s t r i p s

placed i n the void channels of the assembly. Reactor physics measurements

were made i n various density steam environments including cen t r a l spec t r a l

fndices,reaction r a t e t raverses , U-238 Doppler and r eac t iv i ty worths. The

worth of a cen t r a l BqC control rod was measured ( r e l a t i ve t o void) i n the 3 0.0108 g/cm density steam entry environment. The worth increased by 10.3%

compared t o corresponding worth measured i n the reference (dry) GCFR Phase

2 assembly. Recent calculat ions a r e showing a tendency t o overcalcu1.ate

t he steam entry experimental worth.

ZPPR

The Engineering Mockup C r i t i c a l program on the CRBR reference core design

was completed i n June 1976. The program consisted of two phases, one primarily

t o resolve uncertainty i n various safety-related physics parameters while the

other was oriented to confirm design parameters i n a standard EMC program.

Experiments i n Phase 1 measured r a d i a l expansion e f f ec t s , f u e l slumping and

sodium void worths as well as performance charac te r i s t ics of the low l eve l

f l ux monitor. The analysis of Phase 1 core and upper a x i a l blanket void

worth pa t te rn u t i l i zed diffusion theory, with both eigenvalue and perturbation

techniques. The analysis showed r e l a t i ve ly poor agreement with calculat ions

always more pos i t ive than the experiment. However, it appears t h a t the

analyses can be properly cal ibrated by applying adjustment fac tors to both

the leakage and non-leakage contributions.

A program of c r i t i c a l measurements on a CRBR sized core with a l te rna t ing

annular blanket and f u e l zones w i l l begin i n July 1976. The program w 4 1 1 con- \

sist of three par ts involving a clean benchmark assembly of blanket and f u e l

- 199 - r ings followed by a reasonable s h u l a t i o n of probable BOC and EOC conditions.

Following completion of the annular blanket f u e l zoned cores, the ZPPR

matrix w i l l be expanded t o 14'X14'X10'. It i s projected t h a t the matrix

expansion w i l l be completed approximately 6 months a f t e r i n i t i a t i o n . .-

Shielding

The shie lding program a t the Tower Shield Fac i l i t y continued t o support

t he CRBR project . A s e r i e s of experiments were performed t o measure gamma

heating i n a r a d i a l sh ie ld mockup of CRBR. These include TLDs and Ion

chamber measurements i n i ron using i so top i c sources (13'cs, Co, 2 4 ~ a and 5 1 ~ r )

and a r a d i a l sh ie ld mockup of CRBR u t i l i z i n g TSR as a source. The agreement

between experimental techniques was within 3% f o r i so top i c sources and 5%

f o r t he r a d i a l mockup. DOT calculat ions o f ' i s o t o p i c source measurements

were within 3% of experimental r e su l t s . . . An upper a x i a l sh i e ld experiment consis t ing of various thicknesses of

s t a i n l e s s s t e e l , sodium and i r o n was conducted. The experiments proved t o

have s ign i f i can t s e n s i t i v i t y t o d i f f e r en t cross 'sections and energy regions

as t he configurations were changed. Calculations showed good agreement

through 10' of Na and 2' of iron. I n a configuration with 15' of Na followed

by 2 ' of i ron, good agreement of calculat ions with experiments was obtained

through the f i r s t 10-12" of i r on but then began t o deviate , reaching a maximum

fac to r of 3 i n the l a s t 18" t o 24" of the iron.

A pin streaming experiment was conducted primarily i n support of GCFR.

The experiment consisted of 18" s ec t i on of pins approximately 4' long with

void channels between the pins. A f ac to r of 20 dif ference between the O0 and

7' angle was measured, indicat ing a l a rge forward streaming component. DOT

calculat ions t o da t e have shown poor agreement indicat ing a po t en t i a l need

f o r Monte Carlo analysis .

Pipe chase experiments underway a t the TSP w i l l be completed by October

-1976. These experiments a r e designed t o provide information on a t tenuat ion

between the reactors vesse l region and the hea t t ransport c e l l of the CRBR.

Core Optimization Studies ~... ~~ .- ~ ~ ~... ~ ~ . ~ . . . .

Invest igat ions have been car r ied out a t GE, HEDL, ANL and CE t o

a sce r t a in t he key fea tures associated with developing high performance,

i .e., low doubling time, low cost , oxide and carbide fueled LMFBRs. To

iden t i fy these fea tures , trade-off s tud ies have been performed on a l a rge

number of core design fea tures , e.g., p in diameter, p i tch , duct thickneAs,

core height , l i n e a r power, e t c . Work i n t h i s a rea is continuing with

addi t ional e f f o r t s being applied t o improve r e l i a b i l i t y and safety while

maintaining low doubliag times.

Computer Codes and Analytical Methods

I n the U. S. program, advanced ana ly t ica l capab i l i t i e s a r e obtained

primarily by developing computer codes a t nat ional laborator ies f o r subse-

quent use a t i ndus t r i a l a s wel l as laboratory s i t e s . This has led t o the

extensive development of coding standards, spec ia l data formats and data

handling rout ines f o r transmission across functional l i ne s . These

a c t i v i t i e s a id conversion of major codes from developer s i t e s to user

s i t e s within acceptable costs and leve ls of e f fo r t . A supplemental shor t

term a c t i v i t y has been use of remote terminals accessing d i r ec t ly t o

la rge computers a t the development laborator ies . For complex, special

purpose codes of occasional u t i l i t y , t h i s capabi l i ty may be used i n the long-

term a s well.

Computer codes recently completed include M C ~ - ~ , SYN3D, and REBUS f o r

ENDFIB data processing, 3D dimensicnal calculat ions by synthesis-diffusion,

a: in-core f u e l burnup andmmanagement respectively. Work continues with

the MINX-SPHINX data processing approach. M I N X is intended t o process

ENDFIB data t o f i n e groups with pseudo independence of composition, and

SPHINX w i l l perform self-shielding i n the Bondarenko sense, and space-

energy col lapse t o multigroup representations. M I N X i s operating

provisionally although some corrections indicated by code validation t e s t s

a r e s t i l l being made. A more advanced and e f f i c i e n t version is being

developed. SPHINX i s operating a t the or iginat ing s i t e and conversion

t e s t s by other groups a r e i n progress.

More general e f f o r t s , including development, implementation and tes t ing

of s e n s i t i v i t y , 3D transport , and Monte Carlo methods for in te rpre t ing

d i f f e r e n t i a l and in t eg ra l physics data a r e a l so continuing.

Carnegie-Mellon has extended the "Direct Coupled Ray" approximate

method f o r sh ie ld calculat ions i n the "TRANS3" code. Comparisons with the

ANISN and DOT 111 codes a t C-M have shown acceptable TRANS3 accuracy and

rate-of-solution considerably grea te r i n 2D cases f o r the t e s t problems.

Nuclide transmutation and radioactive decay a r e modeled by computer

codes such as CINDER, which i s a generalized code for computing the t e d ~ o r a l

composition of coupled nuclides i n a time dependent f lux environment. The

coupling may be i n any sequential mixture of decay types and p a r t i c l e

reaction. Although the basic code has been i n use f o r about ten years,

recent LASL versions of CINDER maintain currency with programming languages,

computer hardware, and the ENDFIB dais base.

Topics i n applied mathematics a t e explored a t MIT. Tentative so lu t ion

s t r a t eg i e s f o r appl icat ion t o ana ly t ica l reactor problems recently reported

a re :

. Comparison of d i r e c t and i t e r a t i v e methods f o r the solut ion of

f i n i t e element equations

. Selection of f i n i t e element functions f o r three dimensional

problems

. Comparison of FEN and f i n i t e dif ference

. Non-linear methods f o r solving f i n i t e element equations . Collocation methods

. Time dependent response matrix methods

Programs f o r evaluating computer codes i n the areas of A-+s processing,

ZPR analyses, and "large reactor" analyses have been i n i t i a t e d . Comparative

calculat ion of reference cases by various methods a r e underway i n the area

of da ta processing and ZPR analyses. A reference problem has been defined

f o r l a rge reactor analyses.

Actinide Burnue

Technical invest igat ion of ac t in ide product recycle and burnup i n

LMFBR conditions have continued a t GE. It is concluded t h a t ac t in ide

recycle i n LMFBRs of fe rs an a t t r a c t i v e reduction i n the long-term storage

requirements of the act inides , and does not adversely a f f ec t the performance

of t he LMFBR. The eff ic iency of t h i s approach i s dependent on processing

and fabr ica t ion of the recycled act inides . Studies t o fur ther evaluate

t h i s approach a r e continuing. -.

Materials Dosimetry - -

Major current a c t i v i t i e s a r e support of EBR-I1 i r r a d i a t i o n t e s t s and

preparation f o r FTR t e s t i ng , including a character izat ion e f f o r t t o be con-

ducted during i n i t i a l FTR operations. Characterizing the i r r ad i a t ion

environment i n FTR w i l l encompass measurements with diverse ac t ive and',

passive instrumentation techniques t o provide baselines f o r the largely

passive monitoring su i t ab l e f o r ac tua l t e s t s a t prolonged f u l l power.

"Passive" connotes, e.g., sensing f o i l s i n contras t with "active" f i s s ion

chambers.

Those parameters re levant t o materials dose information (f lux h is tory ,

s p a t i a l var ia t ion , spec t r a l indicat ions) and temperature and i ts var ia t ion

(react ion r a t e s , gamma f i e l d s , flow) w i l l be addressed.

The inter-laboratory react ion r a t e program (ILRR) has continued

extensive invest igat ions of f i s s i o n chamber, radiochemistry, f o i l activa-

t ion , helium mass spectrometric measurements, and t rack etch techniques.

I n i t i a l measurements were made i n the Coupled Fast Reactivity Measurement

(CF'RMF) i n Idaho. More recent s tud ies have included the BIG-10 a t LASL,

Sigma-Sigma a t Mol and the Intermediate Spectrum Standard Neutron Field

(ISNF) and Cf sources a t the National Bureau of Standards.

Nuclear Data

Measurements have continued or, a var ie ty of da ta important to reactor

d n . Precision nubar measurements as a function of energy and cross

sect ions of higher ac t in ides a r e underway a t ORNL. Scat ter ing cross sec-

t i o n measurements f o r reactor materials a r e continuing a t ANL up t o 10

MeV. Capture cross sect ions f o r some of the major f i s s i o n products a r e

being measured up t o 200 KeV a t RPI. The ENDFIB-IV data f i l e has been -

extensively tes ted on a var ie ty o f . f a s t and thermal benchmarks. Several -

improvements over the ENDFIB-111 data f i l e a r e noted f o r major applications

of i n t e r e s t . Development of the ENDFIB-V data f i l e i s scheduled f o r com-

p l e t ion i n ear ly 1978. Major improvements a r e planned i n the ac t in ide ,

f i s s i o n product, standards , and 'error f i l e s .

REACTOR PHYSICS ACTIVITIES IN THE FEDERAL REPUBLIC OF GERMANY

compiled by

H. Kiisters

GENERAL

Budget - - - - - - - 1976 - - -

The budget of th les nisterium fiir Forschung und Technologie has been

approved in September 1975. It amounts to about 2.8 billion DM, which is

3.9 % less than envisaged for 1975 /I/.

Energy research and technology requires 1.36 billion DM. This last item

(in Million DM) is being distributed according to: Reactor development: 280;

Reactor safety: 64; Uranium supply: 35; development and control of fuel

cycle: 135; Nuclear waste: 46; research centers (GfK, KFA, IPP, GKSS): 610;

non nuclear energy research 167.

At 26 th Febr. 1975 the utility RWE took over from KWU the 1200 W e ?WR

plant BIBLIS A. Since then this plant could demonstrate an availability of

? 88 %. Not planned reactor shut offs had been caused mainly by conventional

components. Thus it could be proved that such large units can be operated

without any major disturbances,as expected. The second unit, BIBLIS B, went

critical at March, 25th, 1976. Zeropower tests have been completed. 30 Z of . nominal power was reached in April 1976.

- Three topics governed the public discussion with environmentalists:

a) The occupation of the construction place Why1 in the upper Rhine valley

by the concerned public from February to November 1975. The start of the

construction will be postponed to November 1976.

b) Two fatalities occurred in Nov. 1975 during repair of a slidevalve in the

primary circuit of the 252 MWe plant near Gundremmingen in Bavaria. Hot

steam of about 2 5 0 ~ ~ with very low radioactivity scalded two mechanics.

As a consequence, much sharper regulations were established for main-

tenance and repair of nuclear components 121.

C) The siting problem for a combined reprocessing facility and waste disposal

area in the northern part of Germany raises strong emotions. The nuclear

community was addressed by the government that any delay in the procedure

to forward a satisfactory safety report and financing regulations might

delay the licensing of reactor plants 131.

HTRs ----- The advanced thermal reactor concept THTR 300, which is being constructed at

Uentrop in WestEalia, is a pebble-bed high temperature reactor with thorium

as converting material. The construction is being delayed by about 14 months,

mainly caused by additional safety requirements by the licensing boards

(e.g. airplane crash, earth quakes, additional emergency core cooling, secon-

dary shut down system). The start up is now foreseen for 1978. Eio decision

has been taken up to now for a follow-up plant; there are industrial interests

especially for the production of process heat.

FBRs ---- The fast reactor prototype SNR 300, being constructed at Kalkar north of

Duisburg, is delayed by about 20 months, caused by additional safety re-

quirements (especially in connect ion with core destructive accidents) . The total costs will amount to about 2.3.10~ DM. About 60 % of the additional

costs (750 Mio D3) are due to an increase in prices and wages. According to

schedule SNR 300 will go into operation for electricity production by

midst of 1981.

For a near commercial plant SNR 2, the electric power has been discussed to

be between 1300 MWe and 1500 MWe. Research and development work will include

heterogeneous core designs with internal breeding zones, also to minimize

the coolant loss reactivity.

Inrernatiznal-Co_o~zraEio_n Memoranda have been signed by the German and French Governments to amalgamate

the research and development work for fast reactors; similarly the industrial

activities will be pooled. Ways are open to include the present partners

Belgium and the Netherlands into this new joint venture. Existing contracts

(e.g. with USA, UK) will continue. Especially the folow-up reactor development

(1200 MWe and larger) will strongly be influenced by these agreements.

Rekfences /I/ Atomwirtschaft 1211975, p. 633

/ 2 / Atomwirtschaft 1211975, p. 587

/3/ Atomwirtschaft 511976, p. 235

- 206 -

REACTORPHYSICS ACTIVITIES IN THE NUCLEAR RESEARCH CENTER KARLSRUHE

1. Experimental Investigations

1.1 Analysis of Critical Assemblies (SNEAK)

1.1.1 Corrections to Reaction K a t e s S i p P t h e e ~ ~ ~ ~ - ~ r . j & ~ c a 1 Assembligs

SNEAK-7A and 7B due to Re-calibration (E.A.Fischer /I/) .....................................

Soon after the report on assemblies SNEAK-7A and 7B /2/ was completed, the

results of a new calibration of the 239F'u-fission chamber relative to the

235U-chamber by Korthaus /3/ became available. This recalibration, which is

in good agreement with a careful determination of the 2 3 9 ~ ~ mass by low

geometry a-counting leads to the conclusion that the 239~u fission rates

quoted in 121 should be increased by 4 %. The reaction rates in 2 3 5 ~ and

238~ are not affected by this new fission standard. The corrections are

already included in the benchmark specifications.

In this contribution, those experimental results in assemb1ie.s SNEAK-7A and

7B which have to be corrected relative to Ref./Z/ are gathered. The correction

affects Spectral Indices, the Reaction Rate Balance, and the Beff Measurements with a 252~f source. Some of the experimental errors are smaller than those

quoted in 121, because the uncertainty in the calibration has been reduced.

Spectral Indices and Leakage (Compare Table I11 in Ref. /2/

Leakage / Fission rate 2 3 9 ~ ~

* corrected value

Apparent worth of a 252~f-~ource (Table VI and VIII in Ref. 1 3 )

The worth of fission neutrons, as obtained with a 252~f-~ource, is nearly

4 % lower than in 121, because the 252~f worth is normalized to the fission

rate in the central cell. The new results, to be compared to Table VI in 121,

are then shown in the following table.

Central Reactivity Worths of Fissile Materials, and of 252~f Neutrons

wg

SNEAK-7A SNEAK-7B kxperiment CIE (KFKINR) Experiment CIE (KFKINR~

Note that the discrepancy of about 7 % in C/E between the central fissile

material worths and the 252~f worth, which is present in Table VI of 121,

reduces to an average of 3 % with the new calibration. Thus, it almost

disappears within the experimental error.

In addition, the 252~f worth was used, in combination with the measured

normalization integral, to obtain experimental values for Beff. The results

are given in the following table; the Beff values obtained by noise analysis

are also included.

0-,, with a 252~f-~ource, and by noise analysis

BvF, cm 3

SNEAK-7A F v - - 3 252~f-~xperiment 473 + 2.5 % 40560 + 1.5 % 0.00395 + 3 %

Noise analysis 0.00413 ? 6 %

Standard Calc. U)

KFKINR 424 40000 2.95 0.00359 0'. 7

SNEAK-7B 6 N

252~f-~xperiment l I60 f 2.5 L 92700 i 1.5 % 0.00429 ?; 3 % 0

Noise analysis 0.00450 + 6 % 'a cX)

Standard Calc. KFKINR

I048

Integral measurements of delayed neutron fractions in fast reactor spectra

by two different techniques were carried out. The worth of a calibrated 252~f

spontaneous fission source, together with absolute fission rates, and with

the normalization integral obtained from fission rate mapping, gives experi-

mental values for Beff of a critical assembly. These measurements were performed in three Pu-U-oxide fueled and in one U02 fueled assembly. The

pile oscillator technique was used to determine relative yields of 235u,

2 3 8 ~ and 239~u. The results confirm the evaluated delayed neutron yields

by Tuttle, with a slight bias towards a higher 239F'u yield. With these

dada, the central worth discrepancy disappears for SNEAK xeasurements.

Table 1 shows the experimental results, and the comparison with calculations.

The Beff for the different assemblies were obtained by a consistent calcu-

lation, with weight factors obtained by the KFKINR set. As mentioned before,

~eepin's data lead to an underestimation of both the measured Beff and the ratios. The discrepancy is reduced with l'omlinson's data, and virtually

disappears with Tuttle's combined data. Thus, the SNEAK experiments confirm,

in general, the Tuttle evaluation except that they indicate a slightly

higher 2 3 9 ~ ~ yield.

The agreement can even be slightly improved if Tuttle's "fast" value for

239~u is used, keeping the "combined" values for uranium. The latter data

would probably emerge if one included the integral SNEAK measurements in

Tuttle's data evaluation.

One more comment should be made on the comparison of measured and calcu-

lated Beff values. One can conclude on the delayed neutron yields from

such a comparison, provided that the weight factors for Ref•’, as calcu-

lated with the. KFKINR set, are reliable. However, one expects that the

calculated weight factors are in error due to the underestimation of the

fission ratio o / a in the core, and also due to the overestimation of •’8 f9

the 238U fission in the blanket, assuming that the foil measurements in the

blanket are reliable. In order to assess the influence of these differences,

an attempt was made to estimate corrections to the weight factors, using

the experimental fission ratios in the core, and assuming an overestimation

of 15 % of the 2 3 8 ~ fissions in the blanket. Table I shows that this

correction of the weight factors causes at most a I X change in the B eff' Therefore, the conclusions drawn are not affected by uncertainties in the

weight factors.

It is of interest to compare central worth measurements in SNEAK with cal-

culations, using Tuttle's combined yields. The results are shown in Table 2.

Experimental data are from /2,5,6/. The central worth discrepancy dis-

appears entirely for the fissile isotopes, which confirms again the delayed

neutron data used. The consistent underestimation of the 'OB worth may be

due to errors in the calculated spectra. The quality of prediction of the

2 3 8 ~ worth seems to depend strongly on the core composition.

Table 7.: R a t i o s of C a l c u l a t e d and :.?easured C e n t r a l Wo?:ths i n S>?E.lq-Assemblies

( T u t t l e ' s ? e ~ . c i e d Delayed ?>?eutr& ILiel6s)

2 3 5 . "! Assembly w i t h i n n e r core zone s i m i l a r t o LXFBR, and o u t e r d-Euelec? d r i r z r zc-c

235U-fueled b, S i r n i l z r t o 9 R , b u t

Some of the results have been reported already in the 1975 progress report.

Here we will mention two aspects: measurements on Plutonium and Americium

reaction rates and on absorber worth.

For the measurements of fission rates in Plutonium and Americium isotopes

miniature fission chambers have been developed. The results are compared

with calulations, using KFKINR set and ENDFIB-111 data; they are summarised

in - Tab. 3 and Tab. 4. For the analysis Keepin's B-values were used. --

The reactivity worth of 4 absorber materials (Ta, B4C, BI0(90 %), Eu203)

have been investigated 181. Larger deviations from theory have been ob-

served for enriched B C (there is a 30 % error, which might partly be due 4 to the mathematical model used) and for Eu203(20 % underestimation). The

latter disagreement very probably is due to cross section uncertainties.

For the other materials the deviations are less than 4 %, if I "rod" is

being used; the deviations are between 5 and 10 %, if 12 rods are investi-

gated.

Tab. 3: Measured fission rate ratios and material worth in SNEAK-9C (8 % and 20 % Pu240 content in test zone, resp.)

Referencecore 1 o-Error (8 % 2 4 0 ~ ~ ) (20 % 240~U) %

P (mS (gr)

(Referencecore) lo-Error z

I + a

(Referencecore) lo-Error

ENDF-B-111 KFKINR %

Tab. 4: Comparison of calculated and measured reaction rate ratios

Referencecore

K F K I N R

1.049

0.927

0.941

1 .O49,

1.388

20 %TZ4Opu Zone

K F K I N R

1.054

0.925

0.928

1.043

1.384

Referencecore

K F K I N R

Interlaboratory comparisons between GfK, RCN, CEN Mol and NBS have been

made during the past two years, of techniques that are currently applied

for the measurement of fission rates and uranium-238 capture rates in a

number of zero-power fast assemblies related to the LMFBR program. This

effort has involved the exposure of absolute fission chambers and of

activation fulls, to the MOL-CZ central neutron field. Long term flux

level monitoring accuracy of better than f 0.5 % in Mol-ZZ has been

achieved. The perturbation of the neutron field by the access hole has

been studied extensively. Uncertainties in measured reaction rates estimated

by each laboratory relative to flux monitors are between + 1.5 % and + 3.5 %.

Interlaboratory agreement for 235~, 238~ and 2 3 9 ~ ~ fission rates is in the

range t 0.5 % to f 1.3 %. Poor agreement is obtained for the 238~ capture

rate measurements and further interlaboratory effort are recommended in-

cluding complementary experimental techniques 1 9 1 . The work is being continued on fission rates of Np237, Th232, U233 and the higher Pu-isotopes.

Capture rate measurements for Th232 have started.

Detailed information on the physical properties of large fast breeder cores

is expected from the common BIZET program in ZEBRA which is now being

planned jointly with the Reactor Physics Division at the UKAEE Winfrith.

A total fissile material stock close to 2 tons will be used to do experiments

in a core of 2,50 m 0. The investigations will concentrate on

Na-void maps

Power distributions

Control rod interaction.

Fuel loading will start in July of this year, first criticality of an all-

Plutonium core will be reached in October and the experiments will start

in November 1976.

The calculation of the neutron induced y-field near a reactor core is

rather complicated. Therefore a simple iron block (100 x 100 x 87 cm)

with a central Cf-fission source has been built. Measurements have been

performed on neutron spectra between 30 keV and 3 MeV and the y-spectra

(with a Si(Li) semiconductor) between 0.5 and 3 MeV at different positions

in the block 1101. The calculations were based on transport coded, using

y-production cross sections taken from ENDFIB-IV. The measured y-spectra

show a large deviation from the calculated ones, as can be seen from the

figure on page 15.

y-spectra also have been measured in the SNEAK-9C core and in a Li-sphere

at SUAK with a 14 MeV neutron source 1111.

Further measurements have been done on neutron leakage spectra from iron

spheres (some results were already discussed in 1975) 1121 and on neutron

spectra in a Li-sphere. A separate report is given at this meeting 1131.

1.3 Reactivity noise analysis

This subject will be covered in a separate paper to this meeting /IS/.

Iron p i l e a t SUAK

Measurement 20 cm from source

Calculation

40 cm from source

y-Energy (eV)

Comparison of measured and calcu-

la ted y spectra i n an iron p i l e

1.4 Neutron cross section Measurements

Capture and total differantial cross sections of the isotopes Cr50/52/53;

Fe54/57 and Ni62/64 have been measured in the range from 5 - 300 keV at the van de Graaff accelerator. Meanwhile also the capture data for Pu 240

and Pu 242 have been determined. At the cyclotron differential elastic

scattering cross sections have been analysed for 10 angels between 20•‹

and 150' for the isotopes 0, Si, Fe.

At the lead slowing down spectrameter the a-values for U 235 in the range

between 200 eV and 20 keV have been measured and compared with various

data sets 1141. The following figure contains the main results of interest.

+ Thiswork

- KFKINR - KEDAK

Comparison of measured a values with evaluated a-data 8602u296

2. Theoretical Investigations

2.1 Nuclear Data

A full documentation of the re-evaluated nuclear data contained in the

KEDAK-3 library will be completed in the second half of 1976, together

with the description of the necessary data management and retrieval codes.

Also the documentation of the improved group constant processing code

system MIGROS-3 will be available soon. During revision of KEDAK the

graphical representation of the data was very helpful; the first part 56 (non fissile materials) was published 1161. Neutron capture data for Fe ,

~ i ~ ~ , ~i~~ and ~i~~ have been analysed in the resonance region 1171; a

reevaluation for the capture data of structural materials in fast reactors

will be completed end of this year. The experimentally found disagreement

in production of He in neutron irradiated steel with larger Ni-contents

is being studied with regard to the cross-section uncertainties. Further

improvements in the calculation of pre-equilibrium processes were done

by inclusion of direct interaction of neutrons with nuclei. A survey is

given in ref. 1181.

A new expansion of R-matrix (Wigner-Eisenbud or Reich-Moore) expressions

for the collision matrix is derived which is of the Kapur-Peierls (or

Adler-Adler) type. The resulting cross-section representation combines

built-in unitarity with convenience for Doppler-broadening calculations 1151

2.2 Code Development

The modular program system KAPROS (erlsruher Prograrnmzystem) is now available

for reactor calculations 1201. The contents of KAPROS, the modules, will

be extended systematically during the next years.

During the development of KAPROS the following goals have been emphasized:

I.) Flexible coupling of modules into so called KAPROS proce-

dures, allowing the system to adapt to changing engineering

or physics tasks.

2.) Efficient data transfer from module to module with utili-

zation of all free parts of core storage.

3.) Possibility of long term data storage.

4.) Simplified input preparation.

5.) Utilization of the system statistics for improvement of

modules.

6.) Provision for sophisticated error handling.

7.) Provision for restart facilities.

In the sense of KAPROS, "module" means a medium sized program unit for

solving engineering or physics problems. Modules have to be written in

FORTRAN, or at least observe the standard linkage conventions of the IBM

operating system.

System components - -------- ------ KAPROS has been designed for an IBMl370-168 computer running under an ope-

rating system with MVT (multi tasking with a variable number of tasks).

Most of the KAPROS nucleus is written in FORTRAN, only a few subroutines

are in ASSEMBLER. The nucleus as an overlay structure needs 70 K Bytes

in the job region. KAPROS makes use of direct access peripherical devices.

The figure shcss a KAPROS-jokbcated in ies region in a MVT enviroment. All

libraries and peripherical devices which may be used during the job are

also presented.

KAPROS resources

module library

stotislics olher users library

KSUT

user specified files

intern01 l i fel ine KAPROS archive

restor1 lifeline 110- buffers ( j o b to job1

scrotch lifeline other users lmodule to module1

interactive storoge overlay

by o KAPROS job

Nodule execution in KAPROS is performed by dynamical structures 121 121 141 built a object time which exceed the possibilities of predefined overlay

structures.

As a consequence several different load modules are executed during a

KAPROS job. Normally each load module contains several system routines

supported by the operating system, for example the IBM FORTRAN 110 routines.

In order to avoid the confusion resulting from simultaneous use of these

routines in different modules, and to keep the modules short, only one ver-

sion of each routine is used, located within the KAPROS nucleus.

Each module is allowed to call another module - even itself in a recursive way - by calling the system routine KSEXEC. To perform the module sequencing in KAPROS the standard link conventions of IBM FORTRAN are obeyed with

one exeption: KSEXEC partly written in ASSEMBLER is used in a reenterable

form.

The units of data whose handling is supported by KAPROS are the d a t a

b 1 o c k s . Data blocks are linear arrays of data of arbitrary length, to which a name has been assigned. The name consits of a 16 character

literal constant and an integer constant. The internal structure of the

data blocks is left to the modules and is of no concern to the KAPROS nucleus.

The collection of all existing data blocks of a KAPROS job is called the

1 i f e 1 i n e (synonymous to the data pool or data base of other program

systems). The lifeline extends over that part of the core region of the

KAPROS job, which is not occupied by the KAPROS nucleus, the active module

or the OS buffers androutines. Once this space is used up any further data

blocks are stored in a temporary direct access dataset. These parts of the

lifeline are called internal lifeline and scratch lifeline respectively.

Thus KAPROS jobs with few data blocks may be completely wound up in the

core region avoiding inputloutput operations. If necessary a data block

may be transferred from the internal lifeline to the scratch lifeline and

vice versa. The internal and the scratch lifeline are only existent until

the end of the KAPROS job. However, it is possible to extend the lifeline

to a permanent direct access dataset common to all KAPROS jobs called

restart lifeline. Data blocks are kept in the restart lifeline dataset

up to 7 days after the start of the producing KAPROS job.

Further work was devoted to the development of two-dimensional and three-

dimensional diffusion codes in triangular- and hexagonal geometry. The

coupling of the Monte Carlo Code CAMCCO source iteration scheme with

a coorse mesh procedure was improved 1211.

The main objective in the field of code development was dedicated to im-

provements of the code system CAPRI-2/KAD1S, which is capable to describe

the transient analysis in fast reactor safety studies. Special interest

was given to include fuel movement in an hypothetical accident situation,

to describe melting, transport and freezing of cladding material and to

model the behaviour fission gas during excursion. The safety related

reactor physics problems are given in an additional paper 1221 to this

meeting.

2.3 Reactor Applications

In the frame of the licencing procedure for SNR 300 many parametric studies

have been performed for reactivity and flow coast down accidents. Trans-

porttheoretical investigations showed that for a simplified model of two

approaching blocks of fuel (simulating gross slumping behaviour after the

~rimary excursion) yields similar results in terms of energy release as

diffusion theory. For a large fast reactor of 2000 MWe reactor physics

and thermohydraulic studies have been carried out for a conventional two-

zone core concept. Further studies will include unconventional core designs

to improve breeding and to minimize sodium void reactivity.

In the field of incore- and excore fuel cycle problems a review on

the activity inventory in thermal and fast reactors was given.

A briefing summary, including also the results of the IAEA specialists

Meeting on Transactinium Nuclear Data, held at Rarlsruhe, is given

in a separate paper 1241 to this meeting.

Heterogeneity effects arising in fast thermal test loops have been

investigated in order to get a check for physics predictions of

irradiation experiments performed in MOL 7 C loop of the Belgian

BR 2 reactor. A short review is given in a separate paper 1251.

References

~tz, Physics Invc iK-7A and 78, KFH

/2/ E.A. Fischer, P.E. Mc Grz Critical Assemblies: SNE!

I31 E. Korthaus, private communication (1974)

141 E.A. Fischer, submitted for publication in NSE

151 G. Jourdan, SNEAK 9B, KFK-2012 (1974)

161 M. Pinter, SNEAK-9A, KFK-2028 (1974)

171 W. Scholtyssek, H. Fries, KFK-Report 127513

181 G. Henneges, KFK-Report 127611

191 M Pinter et al. "Interlab. Comparison of Absolu Capture Rate Measurements in the ~ol-11 Sec

Standard Neutron Field", Conf. on Nucl. Cross Se

f Two Pu-Fu

ion R anr SS.

ary Intermediate-Energy ons and Technology,

Washington, March 1975

I101 S.H. Jiang. H. Werle, KFK-Report 127513

/Ill H.E. Korn, KFK-Report 2211 (1975)

llLl H. Werle, H. Bluhm, G. Fieg, F. Kappler, D. Kuhn, M. ~alovi.5, KFK-Report 2219 (1975)

1131 M. Fritscher, F. Kappler, D. Rusch, H. Werle, H.W. Wiese, Determination of Neutron Spectra and Cross Section Sensitivity of Tritium Production in a Li-Sphere.

/I41 C.S. Yen, KFK-Keport 2191

1151 Contributet Papers by M. Edelmann, J. Ehrhardt, P. ~ o ~ p i , F. Mitzel and W. Vath to the Topic Power Reactor Noise

/I61 B. Goel, KFK-Report 2233, NEANDC(E) 170/U

1171 F.H. Frohner, KFK-Report 127514

1181 H. Jahn, Proceedings of the IAEA-Consultants Meeting, Triest, Dec. 1975

I191 F.H. Frohner. Proc. CINN-Conf

1201 11. Bachmann, G. Buckel, h'. Hobel, S. Kleinheins Proc. of Conf. Comp. Math. in Nucl. Eng., Charleston (1975)

1211 H. Borgwaldt, KFK-Report 127511 (1975)

1221 H. Kiisters, Contributed Paper to the Topic "Review of Reactor Physics Problems related to LMFBR Safety.

1231 H. Kiisters, M. Lalovit, Proc. of IAEA Spec. Meeting on Transactinium Nuclear Data (TND), Karlsruhe, Nov. 1975, see also KFK 2283 (1976)

I241 H . Kiisters, M. Lalovit, Results and Main Conclusions of the IAEA- Advisory Group Meeting on TND, separate paper to this meeting, see also KFK 2283

I251 H. Kiisters, Homogenisation Problems in Fast and Thermal Reactors, this Heeting

1261 R . J . Tuttle, Nucl. Sci. Eng. 56. 37 (1975

I n s t i t u t f u r Kernene rqe t ik

U n i v e r s i t y o f S t u t t q a r t

1. Transmiss ion p r o b a b i l i t y method f o r neu t ron t r a n s p o r t

c a l c u l a t i o n s i n non-uniform l a t t i c e s (M. Mesina)

A method f o r s o l v i n g t h e two-dimensional mul t igroup t r a n s p o r t

e q u a t i o n w i t h a n i s o t r o p i c s c a t t e r i n g has been developped. 'The

he te rogeneous sys tem i s d i v i d e d i n t o homogeneous s u b r e g i o n s ,

i n which angular-dependent mean s o u r c e and f l u x d e n s i t i e s a r e

d e f i n e d . Working w i t h an i n t e r f a c e f l u x d e n s i t y coup l ing tech-

n i q u e , t h e e n t i r e neu t ron t r a n s p o r t can be d e s c r i b e d by gene-

r a l i z e d t r a n s m i s s i o n p r o b a b i l i t i e s o n l y , w i thou t e x p l i c i t u se

o f e scape p r o b a b i l i t i e s . T h i s method i s a p p l i c a b l e t o geometr i -

c a l l y compl ica ted subreg ions (even concave o n e s ) , as f a r a s

t h e s u r f a c e con tour i s approximated by a polygon.

2. P r e d i c t i o n of r a d i a t i o n damage i n s t r u c t u r a l m a t e r i a l s

o u t s i d e t h e r e a c t o r c o r e ( G . Hehn)

Computer s t u d i e s were done t o improve t h e p r e d i c t i o n of r a d i a -

t i o n damage o u t s i d e t h e r e a c t o r c o r e f o r a PWR and a f a s t reac-

tor . The aim was t o show, how measurements of t h e neu t ron f luen -

c e and c a l c u l a t i o n s of t h e energy s p e c t r a could be combined b e s t . -

Outs ide t h e c o r e t h e r e i s a s t r o n g l o c a l v a r i a t i o n o f t h e neu t ron

f l u e n c e accompanied by a p p r e c i a b l e changes i n neu t ron energy spec-

trum. With one- and twodimensional S N - c a l c u l a t i o n s t h e c o r r e l a -

t i o n s w e r e d e t e r m i n e d between s u r v e i l l a n c e samples a t a c c e l e r a t e d

p o s i t i o n s n e a r t h e c o r e and t h e expec ted damage i n t h e p r e s s u r e

v e s s e l . Displacement c r o s s s e c t i o n s were c a l c u l a t e d f o r t h e i m -

p o r t a n t n u c l i d e s i n s teel from ENDF/BIV d a t a . The neu t ron f l u x

s p e c t r a were weighted a p p r o p r i a t e l y t o g e t t h e d i sp lacemen t r a t e s

a s a f u n c t i o n of neu t ron energy. With t h i s s p e c t r a l q u a n t i t y co r -

r e l a t i o n s a r e shown and p o s s i b l e u n c e r t a i n t i e s a r e d i s c u s s e d f o r

b o t h r e a c t o r t y p e s . Our r e s u l t s show, t h a t c a l c u l a t i o n s of d i s -

placement c r o s s s e c t i o n and neu t ron f l u x s p e c t r a need h igh compu-

t a t i o n a l e f f o r t , i f e r r o r s of 30% i n t h e d i sp lacemen t r a t e shou ld

be avoided.

3. The Application of the Finite Element Method to two- and

three dimensional reactor physics calculations (H.P. Franke;

E. Sapper)

The finite element method (FEM) is applied to time indepen-

dent two- and three dimensional multigroup neutron diffusion

equations.

This method treats these equations as series of inhomogeneous

one-group equations with sources arising from fission and

group-to-group scattering. The formalism is incorporated

into the computer codes FEM-2D and FEM-3D to solve criticality

problems by power iteration techniques. Both codes use semi

automatic mesh generation modules. FEM-3D generates tetrahedral

finite elements with 4 or 10 nodes. The approximate solution

is described by complete lagrangien polynoms of first and

second degree. To improve convergence behavior of eigenvalue

iterations various well known acceleration techniques are

implemented.

To solve the group equations direct methods as Cholesky de-

composition or iterative techniques as Conjugate gra6ient

method have been investigated. For these purposes special

matrix handling and storage schemes are essential and imple-

mented. A reduction in problem size is achieved by special

finite elements which may contain more than one material

region.

Accuracy and computing costs have been demonstrated by cal-

culations for the IAEA benchmark problem for two- and three

dimensions.

The results obtained by FEM-2D and FEM-3D are quite promising

and support further efforts to develop the method.

4. Muleigroup Cross Section Libraries form EMDF/B-IV (M.Mattes,J.Keinert)

The following libraries have been generated during 1975/76 and are

available on request:

- EURLIB-3 Coupled neutron-gamma library including

KERMA-factors for shielding calculatibns.

100 neutron groups (14.92 MeV to 0.00001 eV),

energy boundaries are contained in CSEWG 239-

supergroup structure.

20 gamma groups (14 MeV to 20 keV)

P5, 18 nuclides

- GGC-4/GAM-I1 99 neutron groups (14.92 MeV-0.414 eV) - - content: ;; x-iides

- MUFT 54 neutron groups (10 MeV-0.625 eV)

content: 17 nuclides

- TAEWI- 1 26 extension of our thermal library to 126 neutron

groups (3.053 eV - 0.00001 eV) by taking into acccunt the CSEVG 460 supergroup structure and the thermal

LASER energy boundaries.

content: H in H20 for 8 temperatures

(293.6 K - 623.6 K, P5) graphite for 10 temperatures

(293.6 K - 2000 K, P5)

- Benchmark Analysis homogeneous assemblies (1)

heterogeneous assemblies

fusion reactor blanket studies

Literature:

(1) J.Keinert, M.Mattes: ATKE - 26, 174 (1975)

- 229 -

a i k a l i s c h Technische Bundesanstalt, Braunschweig

P rob lems r e l a t e d t o burn-up i n t h e r m a l and f a s t r e a c t o r s .

Burn-up d e r i v e d f rom t h e h e a t o u t p u t

H . Ramthun

Burn-up o f f u e l e l e m e n t s can be e s t i m a t e d f rom t h e work ing p e r i o d s

and t h e r e l a t e d power o f t h e r e a c t o r ; i t can be d e t e r m i n e d by means

o f r e a c t i v i t y measurement s o r i t can be c a l c u l a t e d f rom t h e a c t i v i t y

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

A new p r o c e d u r e t o o b t a i n r e l i a b l e burn-up v a l u e s of 2 3 5 ~ and 2 3 9 ~ u

i s t h e measurement o f t h e h e a t o u t p u t . T h i s method h a s been d e v e l o p e d

and t e s t e d a t t h e P h y s i k a l i s c h - T e c h n i s c h e B u n d e s a n s t a l t , Braunschweig ,

f i r s t b e g i n n i n g w i t h s e c t i o n s o f t r u e f u e l p l a t e s i n an a d i a b a t i c

c a l o r i m e t e r and l a t e r on w i t h MTR f u e l e l e m e n t s i n a c a l o r i m e t e r

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

FMRB. With a rough knowledge o f t h e " i r r a d i a t i o n h i s t o r y " ( e . g . t h e

m o n i t o r e d power p r o f i l e o f t h e r e a c t o r ) a s p e c i a l program computes

from t h e measured h e a t o u t p u t t h e burn-up v a l u e , t h e b r e e d e d 2 3 9 ~ u ,

t h e a c t i v i t y of 42 f i s s i o n p r o d u c t s and a l o t o f a d d i t i o n a l i n f o r m a t i o n

on t h e s t a t e o f t h e f u e l .

The r e s u l t s o b t a i n e d f o r h i g h l y and low e n r i c h e d u ran ium have been

compared w i t h t h o s e of gamma r a y s p e c t r o s c o p y and o f mass s p e c t r o s c o p y

( a f t e r d i s s o l v i n g t h e sample i n t h e c a s e o f t h e above-ment ioned s e c -

t i o n s ) and t h e y w e r e i n good a g r e e m e n t . I n v iew o f t h e burn-up v a l u e

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

p r o c e d u r e s b e c a u s e it a v e r a g e s a u t o m a t i c a l l y w h i l s t t h e gamma r a y

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

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

Burn-up determination by means of reactivity measurements

R. Hollnagel

In order to obtain the 2 3 5 ~ burn-up in MTR-fuel elements, the reactivities of cadmium absorbers, of fuel element sections and dummies of these sections without 2 3 5 ~ were measured relative to a reference fuel element in a selected core position in addition to the reactivities of the investigated elements.

In principle, knowledge of the shapes of the neutron flux as well as of the importance function in the measuring position,

of the approximate burn-up profiles and, for obtaining the

fission product poisoning, of the rough history of the investigated elements is necessary. The axial shape of the product of

flux and importance for 275~-f uel follows from the measurement of two 235~-sections in different core heights, that one for thermal capture from two Cd-sheets. These products should. be

transversely flat in the selected position to avoid uncertainties. The burn-up and its profile of one fuel element were measured by the gamma-spectroscopic method. The quoted reactivity temperature coefficient of the FMRB was reproduced rather well by control rod movement to compensate for a slight drift in the pool water temperature during the measurements.

Reactor Noise Measurements

E. Viehl

Previous studies of noise analysis at the FMRB were carried out with the aim of isolating the kinetic reactor parameters of our coupled-core system. This was done by using the two-point rewtor kinetic equations from which analytical

expressions were derived describing the measured auto- and cross-spectral densities of the zero-power noise / I / . The actual investigations are based upon this model and concern the intermediate range between zero-power noise and power- noise, paying particular attention to the influence of the

primary cooling system.

Several measurements have been made at different power levels

of the two cores, the volume current of the primary coolant flow being varied too. The resulting noise spectra show a broad

band noise at low frequencies, apparently explained by tempera-

ture fluctuations in the primary coolant flow. Peaks in the spectra indicate vibrations of the control rods. The coherence

functions in the medium power ratio range lie essentially below the graphs obtained without external excitation of the system.

The interpretation of these measurements is not yet complete

and requires new measurements, including a direct analysis of the input quantities such as the temperature fluctuations.

/ I / Viehl, E.: Zero-Power Noise Analysis in a Reactor with 'Pwo Weakly Coupled Asymmetrical Fission Zones.

Nucl. Sci. and Eng. - 56 (1975) S. 422

Determination of flux density profiles of thermal neutrons in the core of the "FNRB" (Experimental Research Reactor, ~raunschweid

K. Knauf, J. Wittstock

The per at ion of a reactor for research purposes often requires

that the spatial distribution of the flux density of thermal neutrons be determined for a part of the core or even throughout the whole core. This is realised at the FMRB with the aid of wires of an Dy-A1 alloy which are suspended perpendicularly

between the plates of the MTR-elements. In this way the flux dansity distribution was investigated for a series of vertical sections through the core. The procedure in detail is as -

follows: a number of the 600 mm long wire probes are activated :.a the measuring plane and the activity is then measured for eaoh wire at 30 points using a wire scanning arrangement. By interpolation betwean all the measured values the measuring

plane is covered by an equidistant net of activity values, i.e. thermal flux density values. A contour line diagram is derived

from the net. Profiles obtained in this way of the thermal flux density across a section through a control rod element for five different positions of the latter, are of particular interest.

Criticality of Nuclear Transports

H.-H. Schweer

Efforts have been made to calculate the reactivity of unreflected and reflected critical uranium- and plutonium-spheres. These calculations were carried out in order to show the reliability of our available cross section sets and numerical methods used for the solution of the Boltzmann neutron transport equation. Moreover we are trying to fill the gap between the ENDF/B library and the secondary cross section libraries used in our programms. Depending on the material employed in nuclear transports, we may

use diffusion theory,the transport method (SN approximation) or the Monte Carlo method to calculate the reactivity.

86020220

Neutron S p e c t r a from Thermal F i s s i o n of 2 3 5 ~ and 2 3 9 ~ u

H. Kluge, K . Weise, H.W. Z i l l .

We have been r emeasur ing t h e the rma l n e u t r o n f i s s i o n s p e c t r a

o f 2 3 9 ~ ~ and 2 3 5 ~ . I n t h e f i r s t s t a g e t h e r a t i o ji39/jE 23 5

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

n e u t r o n f i s s i o n o f 2 3 9 ~ ~ and 2 3 5 ~ w a s measured a t t h e through-

g o i n g c e n t r a l beam t u b e o f t h e Exper imen ta l Resea rch R e a c t o r

Braunschweig (FMRB). Two p r o t o n r e c o i l s p e c t r o m e t e r s were used ,

one w i t h a gaseous r a d i a t o r f o r t h e ene rgy r ange below 1 . 3 MeV,

t h e o t h e r one w i t h a s o l i d r a d i a t o r f o r e n e r g i e s i n t h e range

above 1 . 2 MeV. The measurements show t h a t t h e f ' i s s ion n e u t r o n

s p e c t r a a r e n o t p u r e Maxwel l ians .

INTERATOM, densberg

In the last year besides the physics activities for the protopype SNH 300,

the properties of a near commercial breeder with a conventional core design

have been determined in cooperation with GfK-Karlsruhe and BelgonuclLaire,

Brussels. Preliminary studies were performed including internal breeding

zones (heterogeneous core design). A detailed design for a 1300 W e plant

is under way.

In connection with the design of HTR-pebble bed reactors special calculations

have been done to determine the neutronic features of the empty space above

the core. The basis is a compled Monte-Carlo diffusion theory method,

developed at the IKE of the University of Stuttgart.

A special coarse mesh method has been introduced into the calculation of

2d and 3d diffusion calculations. The feature of this method is that the

transverse leakage can be linearly interpolated in space between two neigh-

bowing points on a coarse mesh line / ] / .

REFERENCES

/ I / J. Lieberoth, Reaktortagung DUsseldorf, p.75 (1976)

Review of the recent activities at GKSS, Research Center Geesthacht

The following code developments for LWR,reactors have been done in the

frame of the nuclear ship project at the Research Center Geesthacht:

1. The code GELS as a PWR spectral code solves the integral transport

equation for the whole burnup in cylindrical geometry taking into

account all strong absorbers occuring in PWR reactors of present design.

It has been checked by comparison to the experimental data of the

Obrigheim PWR reactor. The main advantages of this code are high speed

and easy managability. A detailed report on this code is included at

the NEACRP meeting June 1976.

2. The FLARG-type LWR box code LEIWAR has been extended in order to use it

as reactor simulator. It includes now burnup interpolation, thermo-

hydraulics and other conventional feedback characteristics, Xenon

transients and its compensation by control rods and soluble boron poison

with given priority strategies. Its computing time has been strongly

reduced by'introduction of a coarse mesh rebalancing routine. The LEIWAR-

simulator is applied at the optimization of core power distributions at

load follow cycles and at the optimization of the trim program with

respect to the whole core lifetime.

3. For 2- and 3-dimensional calculations the integral transport codes

BOX2DMG and BOX3DMG using transmission probabilities in large boxes have

been developed and tested at special benchmark problems.

KRAFTWERK UNION, KWU ERLANGEN

The activities in the last year concentrated more and more on safety

related aspects. Special nodal coarse mesh methods have been developed,

using the nodal expansion technique (NEM) l - 4 The nodal coupling

coefficients are determined by 1 dimensional diffusion equations (power

expansion series with Galerkin weighting) and are iterated in the

process of solution; the accuracy is determined by the degree of the

expansion series and by the transversal currents. For the description

of transient reactor behaviour, an improvement of boundary conditions

by using a similar formulation as the response matrix technique, could

be achieved. The determination of spatial flux-fine structure was obtained

with good accuracy in relative large areas by subsequent interpolation

with higher expansion terms 1 5 1 .

A 3 dimensional dynamic code IQSBOX has been developed and tested. Es-

pecially helpful were the various benchmarks 1 6 1 .

The nuclear data library and the processing codes for cylindrical pins

in a quadratic lattice have been improved; special attention was given to

the calculation of control rod characteristics.

2 dimensional burn-up codes were modified by taking into account the axial

flux shape in a second homogenisation process. By this procedure good

agreement could be found with actual 3 dimensional calculations.

In the area of thermohydraulic design the reliability of the W3-R-Grid

correlation for the complex KWU-spacer grid configuration could be shown

in comparison with DNB-experiments in FRIGEN (GKSS).

Calculations of LWR-lattices including Pu-pins require a finer energy and

space mesh. This was taken into account in the corresponding methods. For

the Obrigheim PWR, no significant increase in rod power or decrease in

control rod worth shows up.

References

I H. Finnemann

A Consistent Nodal Method for the Analysis of Space-Time Effects in Large LWR's. Proc. of the Joint NEACRPICSNI Specialists' Meeting on New Developments in Three-Dimensional Neutron Kinetics and Review of Kinetics Benchmark Calculations MRR 145. p. 131 (1975)

/2/ F. Bennewitz, H. Finnemann. H. Moldaschl

CONF-750413, Proc. Conf. on Comput. Methods in Nucl. Eng., April 15-17, 1975 Charleston, South carolina

131 H. Finnemann, M.R. Wagner

The Nodal Expansion plethod: A New Computational Technique for the Solution of Multidimensional Neutron Diffusion Problems. IAEA Specialists' ~eeting on Methods of Neutron'Transport Theory in Reactor Calculations, Bologna, Italy, 3-5 Nov. 1975.

141 F. Bennewitz, H. Finnemann, M.R. Wagner

Higher Order Corrections in Nodal Reactor Calculations. Trans. Am. Nucl. Soc. 22, 250 (1975)

151 K. Koebke

Berechnung lokaler Flus- und Leistungsverteilungen durch nach- trzgliche Interpolation nodaler Grobmaschenverfahren. Reaktortagung Diisseldorf 1976, Paper No. 120, Proceedings p. 79-82

161 W. Werner, H. Finnemann, S. Langenbuch

Two- and ~hree-~imensional Kinetics Benchmark Problem. Erscheint in Trans. Am. Nucl. Soc.. Toronto (1976)

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