GENERAL DISTRI13UTION

'4

NEA COMMITTEE ON REACTOR PHYSICS

&

REACTOR PHYSICS ACTIVITIES IN NEA MEMBER COUNTRIES

October 1982-September 1983 NEACRP-L-266

OECD NUCLEAR ENERGY AGENCY 38 boulevard Suchet 75016 Paris

NEA COMMITTEE ON REACTOR PHYSICS

REACTOR PHYSICS ACTIVITIES I N

NEA MEMBER COUNTRIES

October 1982 - September 1983

OECD NUCLEAR ENERGY AGENCY 38 Boulevard Suchet, 75016 P a r i s

17040 Copyright OECD, 1984

REACTOR PHYSICS ACTIVITIES IN NEA MEMBER COUNTRIES

This document is a compilation of national activity reports presented at the Twenty-Sixth Meeting of the NEA Committee on Reactor Physics. held at Oak Ridge National Laboratory. Tennessee. U.S.A., from 17th to 21st October 1983 .

Australia ...................................... 1

Austria ........................................ 4

Belgium ........................................ 7

Canada ........................................ 13

Denmark ........................................ 18

Finland ........................................ 27

France ........................................ 35

F.R. Germany ................................... 48 Italy ........................................ 70 Japan ........................................ 77 Netherlands .................................... 94 Norway ........................................ 102 Spain ........................................ 109 Sweden ........................................ 121 Switzerland .................................... 129

e United Kingdom ................................. 138 United States of America ....................... 150 . JRC-Ispra ..................................... 157

AUSTRALIA REACTOR PHYSICS ACTIVITIES IN AUSTRALIA

(October 1982 - September 1983)

D.B. McCulloch

Australian Atomic Energy Commission Research Establishment Lucas Heights Research Laboratories, Sutherland 2232, Australia. .

1. METHODS OF CALCULATION t e - 1.1 AUS Modular Scheme

Only limited further development of AUS has been undertaken this

year. The 3-dimensional neutron diffusion module POW-3D has been

completed by the inclusion of perturbation routines; a 1-dimensional

diffusion module AUS100, suitable for calculations of spatially de-

pendent group-condensation spectra, has been added to the scheme to

complement POW-3D; bilinear weighting has been included in the

MIRANDA cross-section preparation module. . 1.2 Monte Carlo Point Reaction Rate Estimation

The use of next flight estimation of point reaction rates in

calculations with the Monte Carlo code MORSE has been investigated

with particular reference to a typical fusion blanket experiment of

a central source in a metre cube of lithium carbonate. (1) The method of Iida and Seki , which involves importance

sampling to determine whether a distant collision should be used,

was found to give substantially improved results. The infinite

variance problem for collisions very near a detector is eliminated

by analytically averaging the contribution for collisions within a

certain radius of the detector. It was also found advantageous to

select the group into which the neutron was scattered, rather than t

include the contribution of all possible groups to the next flight

estimation. . 2. FUSION REPXTOR BLANKET NEUTRONICS

2.1 Validation of Group Cross-section Library

Fusion neutronics calculations at the AAEC are based on group

cross-section data from AUS.ENDF200G which is a 200 neutron, 37 .. i photon group library derived from ENDF/B E. To validate this -

library and the chanqes to AUS system modules which were required

to include photon and kerma factor data, a number of comparisons

have been made with published calculations of conceptual fusion

reactor blanket designs.

The STARFIRE study was chasm Ear the ~najor investigation, 'L'he

published results include tritium production levels and energy

releases for a large number of 1D scoping calculations using ENDFIB ?'7 - data. Though reasonable agreement ( ~ 4 % ) with total energy release

results was obtained, an unexpected difference of about 7% was ob-

tained for tritium production.

The comparison was therefore extended t.o three other studies

for which published tritium production results using ENDE'/B -- were

available, vix. the CTR standard blanket, the European Economic Com-

munity blanket concept for INTOR, and DEMO. Agreement to about 1%

was obtained for these systems.

This attempt to validate AUS system results by comparison with

published design studies is considered less than satisfactory, becau"e. no single self-consistent set of results for a comprehensive range of

parameters appears to be available. An international benchmark calcu-

lation enabling the comparison of many aspects of fusion blanket

neutronics, particularly those not amenabk to direct experimental

verification, would be very useful.

3. RESEARCH REACTOR (HIFAR) NEUTRONICS

The computational models of HIFAR previously developed for use

with the AUS scheme have been applied to routine HIFAR reactivity and

burn-up calculations in an indirect manner by extracting flux factors

and reactivity coefficients for use in a simple fuel and reactivity

accounting program, HIFUEL.

The HIFUEL approach is similar to the semi-empirical method used

for operational purposes over many years via the HIBURN program , (2 ' . but uses calculated instead of measured parameters. This has the

advantage of replacing experimentally based data (e.g. flux depres-

sion factors, reactivity coefficients) from a number of sources, which

are often difficult to reconcile with confidence, with a single self-

consistent data set. It also allows comparison of routine and more

complex, special-purpose HIFAR calculations in a consistent manner.

HIPUEL calculations were run for a sequence of BIFAR operating

programs spanning several years, and the results compared with the

corresponding operational data. HIFUEL consistently predicted react-

ivity changes to better than 0.5% in keff, and fluxes to better than

10%.

REFERENCES 1. Iida, H. and Seki, Y. (1960). Reduction of computational

time for point detector estimator in Monte Carlo

transport codes. NSE - 74, p213.

2 . McCulloch, D.B. and Trimble G.D. (1969). A method of estimating fuel burn-up and higher isotope production

in the reactor HIFAR. AAEC/TMSO8.

AUSTRIA - 4 -

REACTOR PHYSICS ACTIVITXES IN AUSTRIA October 1982 - September 1983

compiled by

F. Putz

1.1 Reactor Computations

An additional shut down system consisting of small

boronated shut down spheres and provided for future

HTRs with pebble-bed cores has been studied at ITP/TU Graz (Institut fur theoretische Physik der

Technischen Universit%t Graz) jointly with the

Institute of Reactor Development of KFA Jalich (FRG). For the neutronics analysis a problan dependent . 60 group cross section library had been developed. In order to take into account the selfshielding of

the shut down spheres it was necessary to develop various cell models and to investigate the inEluence of the geometry underlying the cell models. Criticality calculations using the code CITATION-2D are in good

agreement with experimental results gained from ex-

periments performed on multisphere configurations at

the Siemens-Argonaut reactor of TU Graz.

1.2 Transport Theory

At ITP/U Graz (Institut fur theoretische PhysSk der Universitat Graz) a study is in progress of the passage

of neutron pulses through stratified media using the

method of multiple collision probabilities.

Reactor Kinetics

Based on the theory of the well known TWIGLE-code

a reactor kinetics code is being developed at bFZS

(8sterreichisches Forschungszentrum Seibersdorf)

including an improved hydraulics model. Application

of a coarse mesh rebalancing method leads t0.a

reduction of computer time. An exponential transform reduces the number of time steps, the increments

of which are calculated by a predictor-corrector

method. The code is provided for the simulation of

reactivity transients.

Nuclear Data

The existing data bank of all available neutron-nuclei

scattering lengths has been updated at A1 (Atominstitut

der rsterreichischen UniversitSten, Wien) by including

the recent values from the literature. Altogether 960

values for the various elements and isotopes have been

collected and referred.

EXPERTMENTAL STUDIES

C 2.1 Burn-up Determination

An important activity at A1 in the preceeding year was

the inter-comparison of non-destructive techniques to

determine spent fuel burn-up. Application of ionization

and fission chambers has been compared with the use o f

thermoluminescence dosimeters (TLD) and solid state

nuclear track detectors (SSTD). After careful selection

of optimal TLDs and SSTDs and of various types of con-

verter foils such as depleted, natural or enriched

uranium these d e t e c t o r s have been exposed i n t h e f i r s t s t age t o spen t TRIGA f u e l elements and i n a second s t age t o spent f u e l elements from t h e Halden bo i l i ng water r eac to r . I t could be shown t h a t TLDs

and SSTDs o f f e r some considerable b e n e f i t s compared t o i on i za t i on and f i s s i o n chambers such a s easy t rans - po r t a t i on and handling, simple app l ica t ion , p a r a l l e l exposure t o many f u e l assemblies a t t h e same time and low c o s t s . I n no way these two d e t e c t o r types may replace chambers but they can be used a s add i t i ona l back-up methods i n case t h a t rou t ine measurements show dev ia t ions from t h e expected burn-up value.

2.2 Reac t iv i ty Measurement

Measurements of t h e s u b c r i t i c a l r e a c t i v i t y have been c a r r i e d o u t a t t h e Siemens-Argonaut r eac to r of TU Graz using t h e inverse k i n e t i c s method and t h e source j e rk method respec t ive ly . The r e s u l t s gained wi th these two methods agree s a t i s f a c t o r i l y .

REACTOR PHYSICS ACTIVITIES IN BELGIUM

September 1982 - August 1983 J. DEBRUE (S.C.K./C.E.N., MOL)

FAST RGACTORS

The belgian participation (BELGONUCLEAIRE and SCKICEN) to the RACINE programme in the

French zero power reaCtOrMASURCA at CADARACHE concerned mainly the analysis of

the gamma heating experiments. The experimental data of CADARACHE, WINFRITH and

MOL were normalized on the basis of a common calibration irradiation. There is

general good agreement between the TLD results of the three participants and also

between TLD and ionization chamber, although no firm conclusion can be drawn before

the correction of TLD signals for neutron sensitivity, non-saturation of fission

product activity and cavity effect are applied.

The influence of cell heterogeneity has been calculated with UK-DeBeNe methods.

The absolute normalization of the neutronic fluxes necessary to compute absolute

gamma doses has been investigated, namely through a campaign of intercalibration

of fission chambers at SCKICEN.

The participation of BELGONUCLEAIRE to the CEA studies on the subcritical approach

- for the first loading of SUPERPHENIX began with a survey of different possible

loading policies, calculated first in 1D and later in 2D geometry. The objective 1

is to assess the reactivity, counting rates and source multiplication correction r

factors at various stages of the approach. A detailed programe of measurements

in MASURCA was prepared (RACINE-RlS) to simulate features of the SUPERPHENIX core.

The development of the computer programme SUPERCAPHE, intended to support the

management of the core, blanket and control subassemblies of SUPERPHENIX, was

pursued. The tests concerning the reactivity and power distribution calculated

with the 3D-version of the module GESTXON were carried on. *

THERMAL REACTORS

The main reactor physics activities concern the validation of the calculation

methods applied in pressure vessel surveillance programmes. The achievement of

this objective is based upon systematical comparisons of calculations with mea-

surements performed in research and power reactors, in the frame of a cooperative

agreement with US laboratories (HEDL, ORNL, NBS) sponsored by the NRC.

The validation work related to research reactor experiments makes use of measure-

ment data obtained in PCA (Pool Critical Assembly, ORNL), PSF (Pool Side Facility

of the Oak Ridge Research reactor ORR) and VENUS (critical facility at SCK~CEN, Mol).

The investigated parameters are the neutron propagation from the reactor core up

to the vessel wall, the damage exposure in the vessel and the gamma heating in the

surveillance capsules and in the vessel wall.

The experiments in PCA and PSF are driven by MTR-type reactor and involve compara-

tively simple, finite-slab geometry. The reactor configuration built in VENUS is

an engineering mock-up which simulates rather closely the main features of a power

reactor : low enriched fuel pins, staircase-shaped core boundary, core baffle,

barrel (see figure). In this assembly, the azimutal part of the fast neutron flux

distribution outside the core is significantly sensitive to power rating details

in core boundary fuel pin rows ; the peak-to-peak azimutal variation along the

cylindrical barrel is about a factor 2.

The measurements performed in VENUS from January to June 1983 are summarized in

Table 11. The measurement points are located in a 45" sector. Detailed measurements

in the steel pieces (baffle, barrel) were performed with miniaturized detectors.

Reference neutron and gamma-ray fields developed at BR1 were used for validation

and standardization of the techniques.

The theoretical analysis is made in two steps. Calculation of the power distribution

in the core and determination of the fission sources (X-Y geometry). After conversion

of the X-Y source distribution in a R-8 distribution, neutron transport calculation

(S8 - P3) in order to determine reaction rates and neutron spectra at the measurement locations outside the core. A 17 group cross section set, derived from the ORNL x

171 group set (VITAMIN-C), is used for these calculations. This work is under

progress.

Fast neutron fluence measurements performed around the BR3 reactor core during

previous cycles have been complemented by activation analysis of scraps taken from

the thermal shield at the mid-plane level. On the other hand, material sampling in

the longitudinal weld of the vessel above the core has been performed to assess

the chemical composition of this material ; a coincidence technique was applied

to determine the copper content (% 0.2 wt %) in the samples after irradiation in

BR1. This work is part of a dosimetry-metallurgy programme conducted to better

define the embrittlement of the BR3 vessel steel. . P FUSION

The application of a delayed neutron counting technique has been investigated and

tested in order to measure the neutron flux density at several locations around

the JET torus. Fissile targets irradiated at these locations will be measured 3 with He counters embedded in a moderator at the terminals of pneumatic rabbits.

The irradiation conditions occurring in the (D, D) phase of JET operation (neutron

energy of 2.45 MeV) have been simulated at the BR1 reactor ; the sensitivity and

different measurement parameters have been studied.

In support to the irradiation of fusion materials in BR2, the adjustment of the

(dpalhelium production) ratio in steel is looked for in order to fit better the

irradiation conditions of the first wall in a fusion reactor. Irradiation stra-

tegy and spectrum tailoring in BR2 are examined, namely through the irradiation

of small nickel samples.

Table I

Measured p a r a m e t e r

Gamma d o s e d i s t r i b u t i o n ( R o l l s Royce) (CEGB )

Number Number Technique used of of

( L a b o r a t o r y ) ' p o i n t s r u n s

hmma spec t rum Compton s p e c t r o m e t e r l lJanusl l (HEDL)

2 2

TLD d e t e c t o r s

Gamma dose p e r t u r b a t i o n induced by t h e Compton s p e c t r o m e t e r I'Janusl1

TLD d e t e c t o r s (SCK/CEN ) 1 2

F a s t n e u t r o n spec t rum P r o t o n r e c o i l c o u n t e r s

( 4 d i f f e r e n t t y p e s ) 3 25 ( sCK/CEN )

F a s t n e u t r o n spec t rum and p e r t u r b a t i o n s induced by p r o t o n r e c o i l c o u n t e r s

Thermal n e u t r o n f l u x Dy a c t i v a t i o n d e p r e s s i o n i n s i d e t h e f u e l ( SCK/CEN )

Nuclea r emuls ions (HEDL )

16 9

A x i a l b u c k l i n g i n d i f f e r e n t c o r e and r e f l e c t o r r e g i o n s , f o r d i f f e r e n t n e u t r o n energy r a n g e s

- - Thermal n e u t r o n f l u x I Dy a c t i v a t i o n

d i s t r i b u t i o n (SCK/CEN) [ I 7 /

I

M i n i a t u r e f i s s i o n chambers 17 5

(NBS) ( SCK/CEN

2 3 5 ~ f i s s i o n chambers F i s s i o n r a t e d i s t r i b u t i o n / (NBS L SCK/CEN)

2j8u f i s s i o n chambers (En 2 1.5 MeV) 3 9 ( * )

(NBS & SCKICEN)

F a s t n e u t r o n f l u x d i s t r i b u t i o n

( f i s s i o n chambers)

2 3 7 ~ p f i s s i o n chambers (En > 0.6 MeV) 60 ( * I

(NBS SCK/CEN)

Table I (continued) -, .

Technique used 0 f

(Laboratory) Measured parameter

a s t neutron f l u x d i s t r i b u t i o n (SSTR = So l id S t a t e Track Recorders)

' a s t neutron f l u x d i s t r i b u t i o n ( a c t i v a t i o n )

Axial power d i s t r i b u t i o n

Radial power d i s t r i b u t i o n

SSTR 4 t h 2 3 7 ~ p f o i l s 7 r l SSTR w i t h 2 3 8 ~ 0 2 f o i l s En 7 - 1.5 'MeV (HEDL)

SSTR with 2 3 5 ~ 0 ~ f o i l s f o r co r r ec t ion on 2 3 8 ~ ~ 2 I 8 3

(HEDL)

lh a c t i v a t i o n (En > 0.8 M ~ v ) (HEDL & SCK/?EN)

?

S a c t i v a t i o n (En

Fuel a c t i v a t i o n ( SCK/CEN ) I I I

SSTR i n s i d e f u e l ( HEDL I 6 l 2 /

- ~

EXPEXI!.IE.'4T . _ _ I W-P'LS - F ~ F N L L ~ M B ~ ~ ~

D A T E : I r o r n l ~ U L 8 U r r 23.06. 83

CZNFIGUXAilSi .1 X " 29 1 U-

DENSITY 3 0 0 5 : -~~..NQtiL- . . p - BORON CONCENTRATION 1p.p.m 1- S i A R T - U F SOURCE @ - YES . _

N E U T R O N G E N E R A T O R @-BE---- A B S O R B I N G R O O I fix__. PYREX ROO 0 L8

F U E L REGIONS TOTAL : I 2600 - 58 1 pins C O L O U R R U U B L i l LO R L G l U S II(YES -

O L S I G I I A T .CUSI::Pu OR O F DR T l G I T E D

[j] ~ c ~ e o r ~ u s r m n s r . 7 C S L L

CANADA

REACTOR PHYSICS ACTIVITIES IN CANADA

(October 1982 - September 1983) by

P.M. Gamey

Chalk River Nuclear Laboratories Chalk River, Ontario, Canada

SUMMARY

Support of the design and operation of the current CANDU PHW reactors continued through the development, validation and application of various computer codes, both by Ontario Hydro (OH) and CANDU Operations of Atomic Energy of Canada Limited (AECL). Initial design studies for a 300 MWe CANDU PMWR were started by AECL.

The reactor physics program within the Research Company of AECL (AECL-RC) is largely associated with the development of Advanced Fuel Cycles for the CANDU PHWR. The major activities within this area were the development and validation of lattice cell codes, assessment of the characteristics of the CANDU PHWR for both the thorium fuel cycle and the LWR spent fuel (the Tandem Fuel) cycle, and system studies showing the impact of such fuel cycles on uranium requirements. An experimental lattice physics program in the ZED-2 zero energy research reactor at CRNL is in hand for (Pu,U)02 fuels, and (Pu,Th)Oz fuel is being fabricated for a similar program to start in 1984.

Code development and application continued within AECL-RC in support of the research reactors NRX, NRU and WR-1. Further experiments were undertaken in ZED-2 in support of these reactors.

Further measurements were made of the neutron yield of targets irradiated with 100 MeV protons. Measurements in a graphite-thorium assembly irradiated with 14 MeV neutrons were completed and evaluated.

Development by AECL-RC of the low power heating reactor SLOWPOKE-3 continued and the program was expanded to evaluate small high temperature reactors (SHTR) for production of both heat and electricity with a thermal power less than 20 MW.

CANDU PHWR Support

Measurements were made of the reactor physics characteristics of the four 600 MWe reactors (Point Lepreau, Gentilly-2, Wolsungl and C~rdoba) that were commissioned during this period. Good agreement with calcul Lion was obtained.

Improvements were made t o the i n t e r p r e t a t i o n of the in-core s e l f powered f l u x d e t e c t o r s i g n a l s i n terms of the g loba l f l ~ x / ~ o w e r d i ~ t r i b u t i o n ( l , ~ ) .

An extension t o t h e model expansion method was i n i t i a t e d t h a t a l lows more a c c u r a t e c a l c u l a t i o n s e s p e c i a l l y c l o s e to absorbers (3 ) .

The c a p a b i l i t y t o s tudy t h e dynamics of t h e primary hea t t r a n s p o r t system tak ing i n t o account both the neutronic-themohydraulic coupling and the r e a c t o r r e g u l a t i n g system has been developed and app l i ed t o t h e CANDU PHW system.

Advanced Fuel Cy-

Development and assessment of Advanced Fuel Cycles f o r the CANDU r e a c t o r continued wi th in AECL-RC.

An experimental program of measurements on the c h a r a c t e r i s t i c s of (Pu,U)O? . f u e l was s t a r t e d i n ZED-2. Planning is i n hand f o r a s i m i l a r s e r i e s of ( ~ u , T h ) 0 2 l a t t i c e experiments t h a t w i l l s t a r t i n 1984. These experiments a r e l a r g e l y of t h e " s u b s t i t u t i o n " type toge the r wi th r e a c t i o n r a t e measurements. Development of improved methods t o i n t e r p r e t these s u b s t i t u t i o n experiments con t inues .

The NJOY system i s being used t o prepare ENDFIB-V multigroup d a t a f o r the l a t t i c e c e l l codes LATREP, SOLACE, WIMS-CRNL and RAHABIOZMA. Fur the r improvements have been made t o these codes and new vers ions i ~ s u e d ( ~ 9 ~ ) Several of t h e s e codes cont inue t o be t e s t e d a g a i n s t thermal t e s t l a t t i c e s and i n genera l good agreement has been obtained with exper iment(6) .

The code, FISSPROD, t h a t c a l c u l a t e s the f i s s i o n product concen t ra t ion , i s being u dated wi th c r o s s s e c t i o n s from ENDFIB-V. A new v e r s i o n was issued( ') . A new v e r s i o n of the monopole heterogeneous source-sink code MICRETF was i s s u e d ( 8 ) .

Fur the r development of methods t o so lve the neutron Lransport equat ion con t inued(9) .

Assessment of the core c h a r a c t e r i s t i c s of thorium f u e l l e d CANDU r e a c t o r s continued. This work was, however, temporar i ly suspended t o eva lua te the core c h a r a c t e r i s t i c s of a CANDU r e a c t o r when opera t ing on the Tandem Fuel Cycle ( P C ) . In t h i s c y c l e LWR spent f u e l , a f t e r removal of f i s s i o n products and the higher a c t i n i d e s , i s r e f a b r i c a t e d i n t o CANDU f u e l bundles a f t e r d i l u t i o n , i f r e q u i r e d , by e i t h e r n a t u r a l o r deple ted uraniun. This i s a j o i n t study with the Korean Advanced Energy Research I n s t i t u t e . Similar s t u d i e s a r e i n progress wi th EPDC (Japan) on the recyc l ing of the separa ted uranium from reprocessed LWR f u e l .

Research Reactors

A program t o develop a s u i t a b l e low enr iched uranium (LEU; 20% U-235 i n U) f u e l f o r t h e NRU and NRX resea rch r e a c t o r s has been i n hand f o r t h e pas t s e v e r a l years. The prime candidate f u e l is a U-Si-A1 i n A 1 d i s p e r s i o n , a s t h i s would a l low t h e c u r r e n t geometry and U-235 loading t o be r e t a i n e d . Good exper ience has been obtained t o d a t e i n the i r r a d i a t i o n behaviour of t h i s f u e l . Deta i led r e a c t o r physics c a l c u l a t i o n s a r e now i n process t o i d e n t i f y

its impact on the reactor core characteristics. Previous but less detailed calculations indicated that due to the well thermalized spectrum in such reactors, the impact of the increased U-238 would be quite small.

The cores of these two reactors are highly heterogeneous and thus require detailed reactor physics codes to allow the core characteristics (flux distributions, reactivity worths, etc.) to be accurately calculated. Development continues of improved methods, concentrating largely on the use of a combination of fine mesh, albedos, and discontinuity factors in hexagonal geometry multi-group diffusion theory. Further experiments in ZED-2 in support of these reactors were undertaken(10$ ll).

Advanced Systems

The measurements of the reactor vield from'thick targets when irradiated with - 100 MeV protons from the McGill University Cyclotron was extended to incl.ude iron, copper and thorium. The previous results for lead and lithium were used in the evaluation of a conceptual thermal neutron source based on an intermediate energy proton accelerator(12).

A workshop was held at CRNL in September 1983 to discuss the present status of the codes, data and conceptual designs for the targetlblanket of an Accelerator Breeder.

An experiment to assess the neutronics characteristics of a fusion reactor blanket assembly containing graphite and thorium was completed. The experimental results have been used to validate codes and cross section libraries used in such studies(13).

New Applications

A uranium oxide LEU core is under development for the 20 kW SLOWPOKE-2 research reactor. It had been hoped to utilize the U-Si-A1 in A1 dispersion fuel under development for NRX. However, as reliable manufacture of such fuel with the higher uranium concentration required for SLOWPOKE-2 could not be guaranteed, the evaluation is now focused a U02 pin concept. Reactor physics calculations to identify the core characteristics are currently in progress.

4 Reactor h sics characteristics of the 2 MW SLOWPOKE-3 heating reactor were

P l h ) reported . An experimental program to evaluate the thermolhydraulic characteristics of this concept is in progress. Calculations pertaining to its reactor physics and dynamic characteristics continue(15).

A further program is in hand to establish r'.e technical and economic ability of higher power ( 20 MWth) reactors for du. '. heat and electricity production. Reactor physics characteristics of several concepts have been established. These were an organic pool cc.ncept and two designs in which there were either organic on pressurized li,ht water cooled CANDU fuel bundles in tubes in a D20 moderator.

References

"Rational Mapping (RAM) of In-Core Data" R.A. Bonalumi and N.P. Kherani, CNS/ANS International Conference on Numerical Methods in Nuclear Engineering, Montreal, September, 1983.

"Possible Use of Measurements to Amend Cross Section Tables (PUMA)", N.P. Kherani and R.A. Bonalumi, CNSIANS International Conference on Numerical Methods in Nuclear Engineering, Montreal, September, 1983.

"The Use of a Finite-Difference Scheme to Improve the Accuracy of the Modal Expansion Method" A.R. Dastur and P.S.W. Chan, CNS/ANS International Conference on Numerical Methods in Nuclear Engineering, Fontreal, September, 1983.

"LATREP", J. Griffiths, Atomic Energy of Canada Limited, Report AECL-7603 (1983).

VIMS-CRNL A User's Manual for the Chalk River version of WIMS", G.J. Phillips, Atomic Energy of Canada Limited, Report AECL-7432 (1982).

"Testing ENDFIB-V Data for Thermal Reactors" D.S. Craig, Atomic Energy of Canada Limited, Report AECL-7090, Addendum 1, 2 6 3 (1983).

"FISSPROD-3 An Expanded Fission Product Accumulation Program using ENDFIB-V Decay Data" W.H. Walker et al., Atomic Energy of Canada Limited, Report AECL-6973 (1982).

"Micrete Version 4.1 User's &mual and Program Description", R.A. Judd, Atomic Energy of Canada Limited, Report AECL-7679 (1982).

"Improved Solution of Integral Transport Equatir- Across a Plane Boundary" M.S. Milgram, CNSIANS International Conference on Sumerical Xethods in Vuclear Engineering, Montreal, September, 1983.

"ZED-2 Experiments on the Effect of a Co Absorber Rod on an NRU Loop" G.. Arbique and P.M. French, Atomic Energy of Canada Limited, Report AECL-7515 (1983).

"Experiments Performed in ZED-2 in Support of the Irradiation of (Th,Pu)02 Fuel (BDL-422) in NRU" R.T. Jones, Atomic Energy of Canada Limited, Report AECL-7918 (to be issued).

"Characteristics of a Thermal Neutron Source Based on an Intermediate Energy Proton Accelerator" M.A. Lone et al., Atomic Energy of Canada Limited, Report AECL-7839 (1983).

"Neutronics Evaluations of Activations in Graphite-Thorium Assemblies, 14' ' MeV Neutron Sources: Comparisons with Measurements" S.A. Kushneriuk and P.Y. Vong, Atomic Energy of Canada Limited, Report AECL-8060 (1983).

14. "Reactivity Calculations for the 2 MWth SLOWPOKE-3 Heating Reactor" J.D. Ir i sh , Canadian Nuclear Society 4th Annual Conference, Montreal, June, 1983.

15. "Large-Signal, Dynamic Simulation of the SLOWPOKE-3 Nuclear Heating Reactor" C.M. Tseng and R.M. Lepp, Atomic Energy of Canada Limited, Report AECL-8107 (1983).

., DENMARK

Ris0 National Laboratory

Recent Reactor Physics Activities in Denmark

by

Hans Neltrup

1. Design calculations on - a silicium - doping rig at DR 3 --

Doping silicium crystals by neutron absorption has become an

activity of considerable economic significance at the DR 3

reactor.

The irradiation of the 4 100 mm by 400 mm silicium mono-crystals

takes place in vertical experimental tubes extending into the

heavy water reflector of the 10 MW Dido-type reactor. Quality

requirements by the customer demands that the irradiation should

have a high degree of homogeneity across the crystal. In the

radial direction this is obtained satisfactorily by a device that

rotates the crystal around a vertical axis during irradiation. In a the vertical direction constant inflow of neutrons is aimed at

by stepwise changes in the absorber thickness of a SS. absorber

tube surrounding the rig.

The design of this absorber tube was supported by a combination

of measurements and reactor physics computations.

In order to be able to handle the reactor physics the reactor-rig

configuration was converted as shown in Fig. 1 into a cylinder

symmetric one, in which a realistic representation of the rig

in the centre was surrounded by an annular reflector, an annular

Romogenised reactor core followed by an outer reflector. The

system was completed by axial reflectors in top and bottom.

The flux in this system was calculated by use of three group

diffusion theory. By varying the group 3 absorption cross

section in top and bottom reflector the thermal, group 3 flux

along the rig axis was fitted to the corresponding flux measured

in the actual rig position. The kind of fit obtained in this way

is illustrated in Fig. 2.

Once a good fit is obtained the effect of varying absorber

configurations can be examined. In fig. 2 the flux flattening

effect of a number of absorber configurations is illustrated.

The reactor physical aspect of this problem and the partici-

pation in the RERTR project has triggered an increased interest

in the DR 3 operational staff for a full 3-dimensional reactor

physics model for the entire reactor. One object of such a

model would be an independent evaluation of the reactivity

worth of the different experiments, which is traditionally

monitored semi empirically by a complicated inter-calibration

system.

Test reactors have by nature small dimensions and a complicated

structure, which has made reactor physicists reluctant with

regard to realistic modelling. One particular characteristic of

the DR 3 reactor is the "signal arm" type coarse control rods.

In order to be able to model theese the 3-dimensional diffussion

theory program DC4 has been modified so that triangular meshes

can be introduced. The modified program is operational and has

been tested, but no realistic calculations have been performed

so far.

2. Neutron diffusion theory

An investigation of multigroup n eutr CU on diffusion theory with Cy

emphasis on the keff spectrum, has been initiated. The aims of 0 0 this investigation are the usual ones (distribution of eigen-

values in the complex plane, comuleteness and regularity of

generalized eigenfunctions, positivity of fundamental mode). 0\ ' T I , . . ?. ,., \ ' ' he models considered are realistic in the sense that albedo

boundary conditions and ragged reactor boundaries are taken

into account.

3. Response matrix program for fuel boxes

The interactive effect of gadolinium poisoned pins in the bench-

mark formulated by NEACRP has been examined at the beginning of lilfe by use of the REPRO-FLUS0 program.

A mutual flux tilt of 12% across the poison pins in the diagonal

connecting direction was found. A 4-6% reduction in the total

absorption in the poison pins compared with the absorption in a

single poison pin in a 3 by 3 pin configuration was found de-

pending on the kind of normalisation used.

4. Determination of pin power distribution from coarse mesh

solution

A method, the superposition method, which calculates the local

pin power histories for the individual assemblies was investi-

gated . It consists in developing the solution to the hetero- genous assembly problem with boundary conditions derived from

the global solution after a set of precalculated solutions

(base solutions) to a number of assembly problems.

Typically the base solutions would have white boundary condi-

tions on three sides whereas X.J/ 4 could have a given shape, constant, linear or parabolic along the fourth side. Here X is

the eigenvalue of the problem. This type of base solution proved

to be very efficient even when only a few were used.

In an earlier test of this method a number of representative

regions of a reactor core, each surrounded by suitable driver

and/or reflector zones was selected and calculations carried out separately on each of them.

With a view to certain objection to this procedure a quarter

core was calculated in a subsequent investigation pin by pin as one large fuel assembly. This calculation constituted a re-

ference solution with which the superposition method and other

current methods such as the flux normalization method could be

compared. Two different quarter cores shown in Fig. 3 one with

strong flux gradient and one more homogeneous were investigated.

The results showed that with the superposition method the error in local pin power could be kept within 5-8% using only 8 base

solutions, whereas the normalization method could lead to errors as large as 4 0 % .

A second investigation was concerned with the power distribution

during burnup. The overall quarter core calculations were per-

formed as illustrated in Fig. 4 pin by pin as one large assembly

and the burnup for each assembly was carried out with the

correct spectrum by use of the "Flux-luppe" method. The obtained

assembly power distribution was compared with the one obtained

by the following iterative scheme. Each assembly was burned in

the flux from a detailed heterogeneous calculation with white

boundary conditions. The assembly power level distribution was

taken from a fine mesh overall solution as above. At the end of

each burnup step new homogenized fine mesh cross sections were

fed back into the subsequent fine mesh calculation. The scheme

is illustrated in Fig. 5. The inv~stigation showed that if the

error in pin power should be kept below 5% during the lifetime

which is a desired goal there has to be a coupling between the . pin cell depletion and the global calculation. If up to 10%

errors can be tolerated the global flux tilt effect can be I . neglected in the assembly depletion calculations.

Fig. 1

t Height [crnlZ

Radius Icml R A

DR3-RIG 7V3 wHh &rod and 2mm a-ahtold

- 24 -

Figure 3. Benchmark problem definition.

Test 1

Lay-out of the assemblies.

@ 0.97 WIO U-235

@ uo - - qa 1.8, - - 0 247 - .

- 25 - Figure 4 Core-depletion approximation.

Assembly calculation with zero net current boundary conditions. 1'

& 133 assembly calculations with the boundary conditions derived

from the global solution. I

Purpose: Generation of the homogenized cross sections for the

63 pin cells, water gaps, and control rod blade in each assembly.

k

Purpose: Generation of the homogenized cross sections for the

63 pin cells, water gaps, and the control rod blade in each as- I

Overall calculation on "pin cell level".

8379 homogenized pin cells

+ non-burnable regions.

Purpose: Determination of the average assembly powers and the

detailed shape of.fluxes and currents on the boundaries of the

semblv. _J

7

133 assembly depletion calculafions with the boundary conditiq

derived from the overall solution. I

tions for the 8379 pin cells.

no ...J

L -

Stop.

Furpose: Determination of the deplete* homogenized cross sec- I I

- 26 -

Figure 5 Assembly-depletion approximation.

Assembly calculation with zero net current boundary conditions m @ @ @ ' Purpose: Generation of the homogenized cross sections for

the 63 pin cells, watergaps, and the control rod blade. I Overall calculation on "pin cell level" 1

8379 homogenized pin cells

+ non-burnable regions.

Purpose: Determination of the average assembly powers. I

boundary conditions. Each assembly is depleted with the average

assembly powers found in the overall calculation.

I Purpose: Determination of depleted homogenized cross sections

for the 8379 pin cells.

FINLAND

REACTOR PHYSICS I N FINLAND

STATUS REPORT TO THE NEACRP 1983

Compiled by R .Hoglund

Techn ica l Research Centre of F in land

1. GENERAL

With 38% of t h e e lec t r i c power used i n F in land i n 1982

produced by t h e fou r o p e r a t i n g nuc l ea r r e a c t o r s , and 36%

du r ing t h e f i r s t h a l f of 1983, t h e r e a c t o r phys i c s a c t i v i t i e s

a t t h e Nuclear Engineer ing Laboratory of t h e Technica l

Research Cen t r e of F in land (V'I'T) have mainly aimed a t t h e

s a f e and r e l i a b l e o p e r a t i o n of t h e s e r e a c t o r s . The

development of t h e computer code system h a s t hus been

r e s t r i c t e d t o minor improvements u s u a l l y i n connec t ion w i th

work c a r r i e d o u t f o r t h e power companies and t h e r a d i a t i o n

p r o t e c t i o n a u t h o r i t i e s . A l i c e n s e agreement wi th S tudsv ik

Ene rg i t ekn ik FIR h a s made t h e CASMO cel l code a v a i l a b l e t o

VTT, t h u s r e s u l t i n g i n a mre complete code system f o r t h e

BWR c a l c u l a t i o n s a l s o .

2 . CELL CALCULATIONS

VTT made a t t h e end of t h e y e a r 1982 w i th S tudsv ik

Energ i tekn ik AB a l i c e n s e agreement, which g i v e s it t h e r i g h t

t o use t h e CASMO f u e l assembly burnup program and some

a u x i l i a r y programs (MICBURN, CASPOL, MOVEROD and CLIB) i n i t s

own a c t i v i t i e s and f o r in-core f u e l management and o t h e r

s e r v i c e s t o F inn i sh customers. The programs were d e l i v e r e d a t

t h e beg inn ing of February 1983 and t h e y were i n s t a l l e d on

VTT'S awn CDC computer. A f t e r some s u c c e s s f u l t e s t

c a l c u l a t i o n s , s e v e r a l sets of group c o n s t a n t s have a l r e a d y

been produced w i th CASMO f o r f u e l a ssembl ies o f TvO'S b o i l i n g

wate r r e a c t o r s .

3. CORE CALCULATIONS

The BOREAS code system has again been used extensively to

analyze the burnup cycles of the two TVO reactors and the

suitability of different kinds of fuels for these reactors.

According to test calculations corresponding to critical core

conditions, the effective multiplication factor is, getting

closer and closer to 1 as the equilibrium cycle is

approached, being about 0.998 during the fourth burnup cycle

against 0.993 to 0.994 for the first couple of cycles. The

main reason for this change might be found in the cell data

on which the calculations are based. During a single burnup

cycle, the variations in k eff are about 0.002 to 0.004.

Furthermore, some studies have been made on the 400 MW

version of the SECURE heating reactor. The calculations have,

as usual, required some development work on BOREAS and its

auxiliary codes. Since the CASMO burnup code is now in use at

VTT, the links between CASMO and BOREAS will have to be

improved in order to make the transfer of data as smooth as

possible.

As a rather new application of BOREAS, shut-down margins have

been calculated for a cold core with all control rods except

one fully inserted. Due to the lack of suificient test

material, the reference level of keff is still somewhat

uncertain, but the calculated value for a critical core seems

to be slightly higher in the cold case than in the hot one.

The data condensation code BROAD that produces two-group

parameters for one-dimensional dynamics calculations has been

further developed and used for different applications.

Essentially it creates neutron flux weighted average cross

sections at a number of axial levels in the reactor core

using the results of the corresponding BOREAS (steady state)

calculation.

4. PRESSURE VESSEL IRRADIATION

Development of REPVICS (REactor Pressure Vessel Irradiation

Calculation System) has been finished and the system has been

tested by comparison of calculated and measured 54 ~ e ( n , ~ ) = ~ ~n

reaction rates. This comparison, which was reported at the

6th ICRS in Tokyo in May (paper 9-4), shows good agreement.

The W T (S ) results tend to lie slightly above the measured N values with the maximum difference amounting to about 25%.

The HEXANN (Monte Carlo) results do not show any systematj.~

deviation large enough to be distinguished from the fairly

considerable stochastic scatter. However, since the 54Fe

(n,p)54~n reaction is insensitive to neutrons below 2.7 MeV,

which account for 2/3 of the radiation damage, further

verification through comparison with measured values for

reactions with a lower threshold is desirable.

REPVICS has also been applied to calculations of fluxes and

displacement rates in the Loviisa reactors. The results agree

within 15% with the best previous estimates.

The calculation scheme is shown in figure 1.

5. DYNAMICS CALCULATIONS

Various submodels of the BWR dynamics code TRAB have been

further developed and some new options included in the

program which has been extensively utili.zed in various

transient analyses.

To avoid time-consuming manual tuning of the steady state

axial power distribution a procedure was developed to achieve

the given power distribution. Group constant data are

modified by solving the artificial control rod parameters,

which result in the desired power distribution. The power

fitting option can be used e.g. to obtain a definite initial

state for a transient calculation or in order to get the

power distribution given by a stationary calculation in spite

of the differences caused by different models in stationary

and dynamic programs. An example of the fitting is given in

figures 2 and 3.

The steam line model of TRAB has been further developed to

enable improved analysis of fast overpressurization

transients (important reactivity transients). In the original

model steam is supposed to be saturated everywhere in the

pressure vessel, but the present versi.on takes into account

the possibility of wet or superheated steam both in the steam

lines and the steam 6ome. a

A new acceleration method for coupled neutronic and

thermohydraulic iterations has been implemented in the

program to improve the convergence of the steady state

solution.

TRAB has also been modified to enable analysis of the SECURE

district heating plant. Modelling of SECURE required a

special application of the code because of the unconventional

features of the plant. The TRAB-SECURE model includes many

assumptions that are related to the description of the

consequences of an uncontrolled dilution of the borated water

in the primary system but it might be possible to extend the

model to some other cases as well.

s e c t i o n s

l---+-l c r o s s s e c t i o n s

t a d i a l ( r , e ) I lux and ! i s t r . ( r , z )

f l u x d i s t r .

Azimuthal and a x i a l f l u x d i s t r i - b u r i o n s a t r e l e v a n t

b r a d i i

FISAL R E S G L T S

( r . 0 ) f l u x d i s t r . ( d i f f u s i o n t h e o r y )

F i g u r e 1 . P r e s s u r e v e s s e l i r r a d i a t i o n

c a l c u l a t i o n scheme.

RXIRL HEIGHT

.....-..-... - ......._.......... -....-

F i q u r e 2 . Power d i s t r i b u t i o n o b t a i n e d by u s i n g t h e o r i g i n a l c r o s s s e c t i o n d a t a s e t a n d t h e d e s i r e d I R A 6 d i s t r i b u t i o n .

3.60

3 -00

2.s@

PL W 2.00 3 0

I I 1 I I I 1 I I I I I I I 5 1 I 1 1

i-, ........... ...

I : ! -.. . .ti -..: ............................. ..............................................................................................

. . . .\" :

.............................

* : . , : I j j ;

8 . - .............;..... \.; ..................................................................................................... "- j ; ; \

....... ........ ....... ....... ....... ...'.'.~....... ....... ..... ~ r i g i i a l : c = l k u l r t e d p'ower' \

.

. I : r j i :I : . . .,..... i i...,...; j i j i ..:....... !.. ............................................. r.......)......- . .

....... ....... ....... ....... .......

I :

...... ............... ....... ....... ................................................ .......

i \ i \ ?

.........................

a , &@ -. ....... ;..\..; ; ; i i ; ; ; i i ......-

......

': des i r e d power i i.....- ., : \ ......_

....... . . . . . . . ....... ....

. .

3.; -.... i i i j ' ? - - - - + - . - i - . j -

- - - - i . : - < - -

0.0 1.8 2.0 3.0 4.0 6.0 6.0 7.0 8.0 9.1, 10.811.812.0 13.814.8 15.816.8 17.0 18.819.020-0

RXIRL HEIGHT

F i g u r e 3 . C o n t r o l rod d e n s i t y needed t o modify t h e o r i g i n a l c r o s s s e c t i o n d a t a s e t

i n o r d e r t o o b t a i n t h e d e s i r e d power d i s t r i b u t i o n w i t h T R A B .

- 35 - FRANCE

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

O c t o b e r 1932 - S e p t e m b e r 1983

C . G O L I N E L L I - M. SALVATORES C o m m i s s a r i a t .?i l lEnerg ie A t o m i q u e

I - GENERAL - The e f f e c t of the economic c r i s i s has been to reduce the e l e c t r i c i t y requi-

rement. B u t , a f t e r a national debate, the French government has decided to pursue

the nuclear plant implementation with a slowing down. Now, two 1400 NU r e a c t o r s are . ordered both in 1983 and 1984 a n d one in 1985 with one o p t i o n f o r another one.

The s i tua t ion i s then the following :

New orders : 5 PWR 1400

Operational

Start-up (84)

Under construction

In 1990, the 53 uni ts wil l correspond t o 54 Cd ins ta l led .

T h i s very important domestic program leads t o some requirements in the

f i e l d of the reactor physics, mainly re la ted t o the need :

- t o lenghten the .duration of the batches (12 to 18 months)

- t o increase the fuel exposit ion (33 WD/t t o 45 G W D / t ) .

PWR 900 / PYR 1300

Secondly, s ince the French p o l i t i c s aims to r e p r o c c s spent f u e l , large

quan t i t i e s of recovered mater ia ls w i 11 be ava i lab le and recyclings of plutonium and

uranium a r e studied. ,

The third aspect of the national program which has bearings on the physics

2 3

5

6

program is connected t o the transformation of the PWRs in order to increase the conversion factor . The goal i s t o reduce the natural uranium consumption, in the

perspective of the .LMFBRis implementation a f t e r the year 2000.

0

1

14

For what concerns FaR's, the s t a r t -up of SUPEXPHENIX i s scheduled in the

second s m e s t e r of 1984. The order f o r the subssquent 1500 NWe plant i s foreseen in

1986. Final ly , a j o in t u t i l i ty-industry-national laboratory task force (EC2A) has

. . . , . . . . ?been . :?stab1 ished to study improved F8R designs, in pa r t i cu l a r from the economic poin . , . . . of view. 91100038

2 - FAST REACTOR PHYSICS -

2-1 : In t roduct ion ------------

The major a c t i v i t i e s of the F a s t Reactor Physics - Program dur ing t h e period , October 1982 - September 1983 'were devot21 t o t h e fol lowing top ic s :

- SUPEXPHENIX s t a r t - u p , r e l a t e d both t o t h e a c t u a l experimental program d ? i { n i t i o n ,

and t o t h e mock-up i n NASURCA of some c o n f i g u r a t i o n s of the approach-to-cri:ica;-:y

sequence.

- Control rod s t u d i e s i n t h e heteregeneous c o n f i g u r a t i o n s of the ZCilUE prooram.

- P H E X I X experiment a n a l y s i s t o v a l i d a t e SUPERPHEXIX design c a l c u l a t i o n s . 0

The experiments aimed t o s tudy t h e Boron cap tu re r a t e in t h e 2bssr5er j a b -

asszmbly have been analyzed , and some r e s u l t s a r e repor ted a t t h i s meeting j l j . The

agreement of c a l c u l a t i o n s (us ing s tandard des ign t o o l s ) and experiments a r e sa:isizc-

t o r y , even i f f a i r l y s i g n i f i c a n t u n c e r t a i n t i e s have t o be a s s o c i a i e d t o t h e expe r i -

mental r e s u l t s .

t Prel iminary r e s u l t s i n d i c a t e an agreeme~?t of the o rde r of - 30 5 f o r t h e

ca lcula t ion/exper iment compa;issn on t h e r e a c t i o n r a r e va lues of rhe D I X O S A t i R E 3.u.

r iment , devoted t o t h e neutron dosimetry i n subassemblies locared Far frcm t h e cor?

c e n t e r ( i n r e r n a l fue l s to rage l o c a t i o n ) .

As f o r t h e f u e l i r r a d i a t i o n exper iments , t h e TRAPU e x p e r i ~ e n t a l r e s i l l t s

r e l a t e d t o p ins w i t h , a high Pu240 and Pu241 con ten t (45 % and 15 $ r e s p e c t i v e l y ) ,

a r e now a v a i l a b l e and t h e a n a l y s i s i s underway. The PROFiL-I1 exper iments , r e l a t e d

t o small samples of s i n g l e m a t e r i a l - i r r a d i a t i o n , a r e becoming a v a i l a b l e and the expe-

rimental r e s u l t s r e l a t e d t o the major a c t i n i d e s a r e being analyzed. F i n a l l y , the

DOUBLON experiment f o r t h e i r r a d i a t i o n of s tandard b l anke t subassemblies p ins in

t h e f i r s t and second b lanke t raws nave been completed and they a r e a l s o being

analyzed.

For what concerns the secondary sodium activation, the standard desi'gn cal-

cularion scheme has been validated against the experimental value, and a good agree-

ment (less than a factor of two) has been obtained for the integral value, with a

good prediction of the spatial flux shape, experimentally observed on the heat-

exchanger axis 121.

2-2-2 : SUPERPHENIX

A first criticality configuration has been defined (chess-board type confi-

guration, vith a homogeneous distribution in the core of dummy subassemolies). The

different steps of the approach to criticality have also been defined, with a study

of tne appropriate calculation methods, which should be applicable on a large range

f reactivity values and flux levels on the in-core detectors.

The definition of the physics experiments planned for that phase, has been

continued, with special emphasis on control-rod configurations. In connection with

that program, experimental studies have been performed to define the experimental

methods to be used in SUPERPHENIX. In particular, experimental studies have been

performed in WSURCA to validate an absolute reactivity worth measurement technique

(CARPENTIER method) and a relative reactivity measurement technique (modified source

mu1 tip1 ication) , taking into account the peculiarities of the application of these techniques to a large LMF3R.

2-2-3 : RACINE program

a The RACINE 1 E configuration devoted to multiple control-rod interaction

experiments has been completed 131. Calculation/experiment comparisons show a good

coherence among the values related to different control rod patterns. Some of the

experiments, performed in subcritical configurations, will be repeated successively . in critical configuration, to allow the measurement of the perturbed reaction rate

distributions and to verify the relation between control rod reactivity worth and ,

flux distribution calculation/experiment discrepancy.

The RACINE 1 s phase has been set up to provide a validation of the data and the calculational methods used for the start-up configuration assessment of

SUPEXPHENIX. A series of configurations has been studied starting from a criticai . \ configuration in which 40 dummy steel subassemblies have been simulated (see Fig. I ) . -

[he successive configurations were nade subcritical by the insertion of further

dummy subassemaiies. Flux-to-reactivity relation of control rod ? i f?c t j 3na neurrsn flux transmission to out-of-carp detectors have been studied.

The f i r s t phase of the JASSN program has been c m p l e t e d . Differenx 3oron- ,

s t e e l l aye r s arrangements have been s tud ied and a l s o the us? of borated s t e e l . Cal-

culat ion/experiment comparison i n d i c a t e a worse agrzenent on i n t e g r a l f l u x responses

i n comparison t o s tandard s t ee l iNa mixtures /4/.

2-2-5 : E?MINE e x ~ e r i m e n t s

Heterogeneity experiments have been perforxed zo compare pin and ? l a t e conf i - . gura t ions . For each geometry, ?ito d i f f e r e n t , and ccmparable, degrees of he terogenei -

t y have been s tud ied . These experfments should be completed by the end 3f 1P83.

2-3 : Theoret ica! s t u d i e s -------------------

2-3-1 : Control rod s t u d i e s

Contrg1 rod s t u d i e s have been ad res sed , besides the a n a l y s i s of t h e i n t e -

gral exper inents of the

energy dependent and of

should be used to shape

RACiNi 1E conf igu ra t ion , t o s e n s i t i v i t y a n a l y s i s Sotn of

space dependent e f f e c t s / 5 / . The r e s u l t s of t hese s t u d i e s

f u t a r e i n t e g r a l experiments on MASURCA.

2-3-2 : Hetercgenei t : ~

The CADENZ.A e x ~ e r i m e n t , ?reposed by the Y K t o the Y E W P / 6 / a s a nerero-

genei t? c a l c ~ l a t i o n method b e n c h a r k , has been analyzed 20 c o l l i s i o n p r o b a i ? i t i e s

rout ines have jeen used t; anzlyzed the pin geometry, and a s i n u l a t i o n has jeen

attempted of the 3C s i f e c t s of the ? l a t e geometry. A discrepancy (-2 0 . 6 : i n X / K )

has been 3aserS/ed ahen compar!np the C / E f o r t h e two geometr ies . Fu r the r s x d i e s to

explain the discrepancy a r e planned, i n t h e framework of the NEACXP e x e r c i s e . .

In o rde r t o imprcve tne e f f i c i e n c y of s tandard ?do dimensiondl t r i n s m r t

c a l c u l a t i o n s , jassd on t h e d i s c r e t e o rd ina te method, i t has Seen develoned i i3

t r a n s g o r t code, r h i c h aakes use of a d i f f u s i o n i n i t i a l i s a t i o n , t h e s y n t 5 e t i c 3 i f -

fusion a c c e l e r a t i o n method For inner i t e r a t i o n s and the Tchebychev ne thca f - r 2 u t 2 ~

i t e r 3 ~ 5 o n s . ;\ gain :'n computer time s f a f a c t o r '2 d has been a b t a i n e d , < n x m a r : -

Son * i th the j ~ a n d a r d ANISM m d e . i t i s planned t o zxtend the study to a 29 c x e .

2-3-4 : Multidimensional design calculat ion codes

The new code system f o r design calculat ion / 7 / i s now operational in the

3D hexagonal option. Substantial computer time gains a re expected with respect t o

older versions.

For what concerns the inhomogeneous calculat ion option in 2D or 3 D , s ign i -

f i c an t improvements have been obtained using appropriate i n i t i a l i s a t i o n s from the

associated homogeneous problem. Moreover parametrical s tudies have been performed

to define optimal u t i l i s a t i o n options (number of groups, mesh s i ze e f f e c t s , axial

buckling, e t c . . .).

'2-3-5 : Cross-section Evaluation

There has been ac t ive p a r t i c i ~ a t i o n t o the JEF pro jec t , both in terms of

the def in i t ion of data needs (e .g . covariance matrices) and in terms of evaluation

of data (e.g. Pu239 in the resonance reg ion) . Future work concerns the benchmark

phase of the pro jec t , and in par t i cu la r the associated s ens i t i v i t y and uncertainty

s tudies .

Work has been performed in the framework of the task force on U238 reso-

nance parameter discrepancies and on a c r i t i c a l study of Ni, U235 and Pu239 da ta ,

i n the framework of I N D C .

F inal ly , a methodology i s being developed f o r . t h e simultaneous evaluation

@of isotopes of atomic masses A , A - l and A-2, with application t o the U and Pu i so-

topes.

In connection with the JEF f i l e development, i t i s foreseen to improve the

present point cross section processing s r r a t eg i e s . A system, based on the NJOY code

i s developed a t SACLAY (THEYIS system). I t i s foreseen t o improve the present unre-

solved resonance data treatment and to nake the system compatible with the sub-group

method, used in the CARNAVAL system.

2-3-6 : Fission ?yoduct 9ata

I n view of the r e su l t s of the YE.AC2P benchmark on a LXF3R burn-up caicuia-

t ion / 8 / , i t nas,been considered of i n t e r e s t to assess data unc?r ta int i?s 3ssscia tec

w i ~ h pseudo f i s s ion product cross-section cs lcx la t ion . A f i r s t study, re la ted to

f i s s ion ?r?duct y i e ld s , has been comple?sd / 9 / . Further s tudies :ii;; ;3ncern the v o l a ~ i f i s s ion ?roauct nigrat ion, and the global data adjustement, basea on oower reactor

, ; ' , ' ;

' fuel pin ir-adiation exoerinents. 97 100042

3 - LIGHT WATER REACTOR PHYSICS -

The Light Water Reactor Physics program has found a new i n t e r e s t due t o the

extension of the French nuclear program. The three main areas for research and deve-

lopment a re :

- improvement of the reactors in operstion

- modification of the fuel cycle in the next future

- s tudies of new types of reactors fo r mid-term.

3-1 : .Theoretical s tud ies -------------------

Transoort theory

A new method has been developed for f i ne s t ruc ture burn-up calculat ions of a very heterogeneous large s i z e media. I t i s the generalization of the uell-known sur-

face-source nethod, allowing the coupling of actual two-dimensional hetsre5eneous

assem01 i e s , known as ' substructures" . The method has been applied t o a rectangular

medium, divided in to sub-s t ructures , containing rectangular a n d / o r cyl indr ical f u e l ,

moderztcr and s t ruc tu re elements. A zone-wise f lux exgansion i s used to formulate a

d i r ec t mil i s ion probabil i ty problem ,dithin i t . The czupl ing of the sub-structures

i s perforxed by making extra assmptions on the currents entering a n d leaving the

in te r facss .

Xeu cha rac t e r j s t i c netncds fo r the solution of the x , y geometry d i sc re te

ordinates neutron t ranspor t equation have recently been introduced : f i ve polyncmials,

without any c3nt inui ty requirevent are used on each aesn c e i l , the f i r s t one appro

ximates the angular f lux ins ice the cel l a n d the others are vaiid along the ce l l -.

edges.

Diffusion theory

An unified formulation of non conforning f i n i t e elements with quadrature

formula and simple nodal scherne has been developed. The theoretical conver;ence i s

obtained fo r ;he ~ r e v i o u s schene ,,then the nesn i s re i lned . Xumerical t e s x are

provided in order t s support the theoretical r e s u l t s .

3-2-1 : EOLE - The (2QEEHexper imenta l program announced a t t h e 24 th meeting i s be ing p e r f o r -

med a t t h e EOLE f a c i l i t y . I t i s devoted i n p a r t i c u l a r t o the q u a l i f i c a t i o n of the

absorber e f f e c t c a l c u l a t i o n s (Program s t a r t e d i n October 1982).

The f i r s t s tud ies are devoted t o the gadol in ium e f f e c t . The program cons i s t s

i n r e a c t i v i t y e f f e c t measurement and power o r f l u x d i s t r i b u t i o n f o r :

- parametr ic s tud ies (Gd content , type o f support ) f o r one c e n t r a l r o d

- i n t e r a c t i o n between absorber rods ( 1 t o 12 Gd rods, r o d d is tance, e t c . . . )

- i n t e r a c t i o n between absorber and ,water holes

- i n t e r a c t i o n among gadol in ium and s i l v e r , indium, cadmium

- s i m u l a t i o n o f a PWR assembly.

The second p a r t i s devoted t o the o the r absorbers and m a t e r i a l s : s i l v e r - indium - cadmium - hafnium - B4C - borated g lass , s t e e l , z i rconium and "grey rods" .

The t h i r d p a r t w i l l be performed i n 1984 w i t h the 3eff measurement, the

r e f l e c t o r e f f e c t , (edge subassemblies, b a f f l e , ...).

A new experiment i s prepared f o r the end o f 1984. I t i s a t i g h t l a t t i c e assem-

b l y w i t h mixed oxyde f u e l . The neutron parameters f o r t h i s k i n d of l a t t i c e must be

q u a l i f i e d . For t h i s experiment, i t i s necessary t o design a very l a r g e zone t o o b t a i n

the asymptot ic spectrum.

3-2-2 : MINERVE

The N N E R V E f a c i l i t y i s shared SeVdeen the FBR program an0 the PWR one. The

name o f the experiment f o r the second case i s MELODIE. The goal of MELODIE i s main ly

based on the o s c i l l a t i o n technique. Small samples are o s c i l l a t e d i n the se lec ied

spectra. The microscopic cross sec t i ons can then be tes ted . Three o b j e c t i v e s charac-

t e r i z e d the experimental program :

- thor ium cyc le

- recovered uranium r e c y c l i n g

- g a d o l i n i a d i t h uranium oxyde o r a lumin ia .

The MELUSINE r e a c t o r i s loca ted a t t h e G R E X O B L E Center . A 13x13 U02 rods assem-

bly i s loaded a t t h e Center of tne !4T? s o r e . The goal of the i r r a d i a t i o n i s to fol low

t h e gadolinium dep le t ion but a l s o the power d i s t r i b u t i c n as a func t ion of time and

t h e i s o t o p i c composition of the various types of rods . The experiment '+ES 5:Erted i n

Novernber 1282. The f i r s t poisoned rod .was unloaded i n March 1983. The rx3er ;men~al

program c o n s i s t s i n two measurements :

- nap of f i u x d i s t r i b u t i o n by +< scanning over the rods

- d e s t r u c t j v e i s o t o p i c a n a l y s i s .

The r e s u l t s of the isot .opic a n a l y s i s 'will be obtained i n Septenber 1983.

end 3 i the i r r a d i a t i o n i s s iheduied i n Zanuary 1984. Tw

3-2-J : j ~ e n t f u e i Analysss

- ihe present French program i n the f i e l d of the Spent Fuel Analyses i s a long

term one, ,di thout new s p e c i f i c developments.

-. Ine r e s u l t s from TIHANGE r e i c t o r a r e analyzed and ;he s2arch of tendencies

shows m a ? ? d i f f e r e n c e s betxeen tne p red ic t ions and the measurements.

-. ine FESSEXHEiM program has been s t a r t e d by xeasuring the i s o t o p i c composition

of t h e second cycle f u e l . The four-h and the f i f t h cyc les da ta a r e expected t o be

a v a i l a b l e i n :ne near f ~ ~ t s r e .

In the frame of t h e impr3vement of s u r v e i i l a n c e technique, we can indicaxe

t h e 3a in 3c t tons during the l a s t year : . -

- implementztion of a new algori thm t o us2 the theraoc3uole xeasur%nent,

- follow-uo 3 i :he a a i f l ? j e t t i n g and, 2180, mechanical v i b r a t i o n s jy ne l~ t ron noise

3 n a l y s i s .

The indus t r ia l applications a re d i r ec t l y connected w i t h the reprocessing plant

construction and the follow-up of the process. The isotopic cor re la t ion tecnnique

i s used with sucess t o measure the burn-up of the supplied spent fue l assemoly a t the

entrance of La Hague plants . An indus t r ia l use i s envisaged a t the time of jsbassem-

b l y departure from the storage ponds.

The avai lable recovered materials introduce a requirement f o r the recyclings.

Two types a re studied :

- recovered uranium recycling by re-use in ?WRs w i t h re-enrichment

avai lable plutonium recycling.

Compatibility w i t h the ocher options (hign burn-up, very long batches; a re

examined to give the necessary I n p u t t o a t the econcmics s p e c i a l i s t s . In t h i s frame-

work, various s t r a t eg i e s will be compared.

. . 2 .

- R E F E R E N C E S -

G . HUMBERT e t a1

Paper a t t h i s meeting

J.C. CABRILLAT e t a1

Paper presented a t 5th ICRS

TOKYO May 1983

G . HUMBE?i e t a l

Paper presented a t the ANS Topical Meeting

KIAMESHA L A K E September 1982

2.P. TRAP? - A . O E CARL1

Paper presented a t the 6 th ICRS

TOKYO May 1983

G . PALMIOTT! - M . SALVAiORES

Paper presented a t t h i s meeting

J .M. STEYEXSON

NEACXP-A-445

C . GIACCMETTI e t al

NE,4CRP-A-531

G. PALYIOTTI - 1. SALVATORES

NE.ACXP-.4-504

G . RIMPAULT - L.. MARTIN-DEiDIEX

Paper t o be presented t o S p e c i a i i s t fleeting ou Yields and decay da ta of

F i s s ion Product Nuclides

BROOKHAVEX - 2 4 . 2 7 October 1983

M. ARIES - d . 3GUC:iARD e t ai

Quatre ans d 'exp6r iences 3 ' u t i l i s a t i o n des c o r r e l a t i o n s i so t cp iques c 3 ~ c u l f e s

dans l ' S t a b i i s s m e n t d u b i i a n d '?nr rGe 3e !a Hague.

4 - Co1:cque Int- . rnat ional j u r ? e s ? r?ar4s r e c e n t s e n .na t ie . re de i . a r?nc !ss

appl icuees 3ux n a t i e r e s nuc:$a i res .

WIEX - 8,:2 Uovemoer 1983

X. DARRCUZET e t 31

E t a o l i s s e n e n t d ' u n b i l a n d ' e n t r i e de c o m b u s t i b l e s de r e a c t e u r s 3 eau

p a r mesures non d e s t r u c t i v e s gamma e r n e u t r o n .

AIEA - NIEY 8 . 1 2 Yovember I983

P. 3ERNXD e r a1

Exper ience w i t h u t f l i z a t i o n o f n e u t r o n n o i s e i n PWRs f o r m o n i t o r i n g t h e

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

AIEA - PRAGUES 21.25 dune !?62

8 . PAPIX s t a;

U t i l i s a t i o n des thermocouples p o u r l a s u r v e i l l a n c e de l a d i s t r i b u t i o n

r a d i a l e de p u i s j a n c ? des r + a c t e ~ r s 3 eau p r e s s u r i s & .

AIEA - Co l loque i n t e r n a t i o n a l s u r l a commande e t l ' i n s t r u m e n t a t i o n des

C e n t r a l e j n u c ? e i i r e s .

MUNCYEY - l i . l S O c t s ~ e r 1 x 2

J. 30UCSARD e t a1

B e s o i n en dcnnees n u c i e a i r e s pour l e s r e a c t e u r s 3 n e u t r o n s t h e r x i q u e s

I n t e r n a t i o n a l t 3 n i 6 r e n c s on N u c j e a r d a t a f o r Sc ience and Technology.

ANTWERPEY - September 1982

8 . DUCHEYII e t 31

Decay h e a t c a : c s l a t i o n s any t h e CEI r a d i o a c t i v i t y d a t a bank and :he

PE?IN cooe.

ANTXERPEY - Septemcer 1982

M. DARROUZET e t a i

Uranium r e c y c l e i n PWR's

ANS Topic21 Mee t i ng.

KIAMESaA LAKE - Sepremcer i 9 8 2

T. VE2GNAUD

TRIPOLI 2 - Energy dependen1 t h r e e d i q e n s i o n a l MONTE CARLO Code.

I S P W C3urse on MGNTE CARi9 Methods and T h e i r . A p p l i c a ' t i o n t o z a d i a r i o n

S h i e i d i n g

IS?!?A - 25.29 k t o b e r i 9 8 2

C. NESSATNGUIHAL-YRUYNODGiiE

!C:ranspor: calcuiation 2srimarion of ihe nrurronicai effect induce0 3n

in-core dexcczors siyals by fuei is~emoly vibrations.

XFK Reacror ?hysics Oepr. - 16th iniormai fleeting ~n deector doise. 1 8 . 2 0 May 1 9 8 3 .

P . HENNART - J.J. LAUTARD - A. KAVENOKY

On the relationship between some modal schemes and the finite element .

method in static diffusion calculations.

Mathematics and Computational Meeting SALT LAKE C I T Y - 2 8 . 3 1 Mars 1 9 8 3

J . LAUTARD - A. KAVENOKY

State of the art in using finite element method for neutron diffusion

calculac' ion.

Mathematics and computational meeting SALT CAKE CITY - 2 8 . 3 1 Mars i 9 8 3

Z. STANKOVSKI - A. u\VENOKy

A Sub-cri tical method for integrai transport calcuiations

Mathematics and computational meeting SALT LAKE CITY - 2 8 . 3 1 Mars 1983

M.F. ROBEAU - J.J. LAUTARD - A. KAVENOKY

ARIANE-8 : Un systPme d'aide a la programation scientiiique. I N R I A - PARIS, 1 7 . 1 9 Mai 1 9 8 3

P. RIBON . Extraction de 1 'information sur les densites de niveau nucleaire dans la 0 region ies energies de resonance.

A I E A - BROOKHAVEN (USA) , 1 1 . 1 5 Avril 1 9 8 3

X. KAUEBOKY - J .J. LAUTARD The neutron kinetics and thermal-hvdraulic transient comoutational module . of the NEPTllNE svs?.em - TRONOS.

M I S - K I P M S 3 A LAKE. 2 2 - 2 4 Se~tembre 1.982 - CEA CONF 6470

GERMANY, F.R.

REACTOR PHYSICS ACTIVITIES IN THE

FEDERAL REPUBLIC OF GERMANY

Compiled by

H. Kiisters

Kernforschungszentrum Karlsruhe

GENERAL

On recommendation of the Minister for Research and Technology,

the Federal Government decided on 26 April, 1983 to complete the

construction of the fast reactor prototype SNR-300 and of the high

temperature prototype reactor THTR-300. The necessary additional

financing for SNR-300 amounts to about 1.15 billion DM and will

be shared by the government (about 2 / 3 ) , the utilities and the

manufacturers (about 1/3). For THTR-300, the Federal Government

and the State of North-Rhine-Westfalia will finance about 40 % of

the additional costs of about 1 billion DM, utilities and manu-

facturers take care of about 10 %, about SO % of the additional

costs will be covered by a collateral loan.

The fast test reactor KNK-I1 has become critical with the second

core loading. The pin diameter of this core is 7.6 mm (the first e

core had pins with 6 mm diameter). The problem of gas bubbles in

the sodium which often led to reactor scram in the first core,

was removed by installing cyclons in the primary circuit: the gas

then no longer passes through the core. Up to now, KNK-I1 could

be operated without any perturbation.

For the construction of a reprocessing plant of 350 t throughput

per year two sites are under discussion at present.

I. REACTOR PHYSICS ACTIVITIES AT THE

NUCLEAR RESEARCH CENTER KARLSRUHE

1. Fast Reactor Physics

Experiments on reactivity effects of material displacements in

fast reactor accident situations were completed in the single-zone

uranium fuelled critical assembly SNEAK-12A. Calculations used

current KfK methods and data and partially also the corresponding

moduls of the SIMMER-I1 accident analysis code system.

For all cases investigated satisfactory agreement between theory

and experiment was reached when two-dimensional transport eigen-

value calculations were used. The application of perturbation

theory or diffusion theory in a number of cases led to larger dis-

crepancies, particularly when the experiments involved fuel com-

paction /I ,2/.

SNEAK-12B contains a central Pu-zone with fuel rods instead of

fuel plates which is a better simulation of a fast power reactor;

the results obtained so far confirm those found in SNEAK-12A / 3 / .

Within the frame of the DeBeNe-British BIZET program measurements

of the worths of simulated control rods for fast power reactors

have been made in ZEBRA and SNEAK by the modified subcritical

monitoring method (MSM) . The assemblies used were the conventional and unconventional core arrangements from the BIZET programme and

a compacted version of a conventional core. The control rods were

mainly natural B4C, with some study of 40 % enriched B4C and Eu203.

Correction factors for the MSM were obtained from eigen-value and

source-mode diffusion-theory calculations in XY geometry.

The measured rod worths and interactions are compared with calcu-

lated values from methods and data similar to those used by the

different participants in the BIZET programme to predict the corre-

sponding parameters in fast power reactors. In general, acceptable

agreement is found / 4 / .

A review of the experiments and their interpretation in the

DeBeNe-French RACINE-program covers especially the determination

of critical mass, reaction rate distributions and spectral indices,

reactivity of sodium voiding and control rod worth in heterogeneous

assemblies of MASURCA.

The analysis is made independently by CEA and DeBeNe using their

own calculational techniques and cross sections / 5 / .

Another review article on the physics of unirradiated LMFBR cores

describes how problems have been identified and widely solved

over the past ten years /6/. The following table summarizes the

status of the prediction capability of present theoretical methods.

A brief outline of remaining tasks is also given in the paper.

Conventional collapsing for group cross sections used in multigroup

nuclear reactor calculations is usually performed using normal

(real; direct) flux weighting. The application of more advanced

collapsing procedures using in an appropriate manner real, adjoint

and bilinear weighting was in tine past restricted in general to

fundamental mode problems. Although the principles have been pub-

lished for more than ten years, there seems to exist little recent

experience on the merits and possible difficulties of these im-

proved procedures for multidimensional diffl~sir-. problems for

practical purposes, e.9. in the nuclear design and analysis of

large - Liquid - Metal - Fast - Breeder - Reactors (LMFBRs). A recent pub- lication explains certain somewhat unusual features of the collapsed

group constants obtained by adjoint and bilinear weighting and

describes the experience gained in representative I-dim. and 2-dim.

test cases / 7 / . It could be shown for criticality and perturbation

calculations that in general it is advantageous to apply these

improved collapsing methods if the necessary precautions are taken.

The possible disadvantages seem to be only minor and the associated

complications are considered to be tolerable. Compared to the con-

ventional collapsing procedures these improved procedures are

especially useful for multidimensional problems. It could be proven

that they are favorable with respect to computer time and storage

needed due to the fact that the necessary number of coarse groups

can be kept fairly small without deteriorating too much the accuracy

Table Survey of the P red ic t ion C a p a b i l i t i e s of Present Group S e t s i n Fast c r i t i c a l Assemblies. Figures: (C-E)/C i n %; exc 5 exception. (See Ref. 6)

Group Set

CAkNAVAL-IV (France)

JFS-2 (Japan)

Kf K I N R (DeBeNe)

OSCAR-76 (USSR)

BNAB-78 (USSR)

JENDL-2 (Japan)

Adjusted Nuclear Data

-0.5 and I f8 / f5 : = 4 b e t t e r I I

-0.5 and b e t t e r I only values from

ad jus t ed r a t i o s a v a i l a b l e

0.3 t o 1 exc: RACINE

(1.3)

?0.6

(2.9)

f8 / f5 : -0.5 t o 8 - + 2 (8 f o r BFS-35-2)

Ron-Ad justed IJuclear Data

f8 / f5 : -3 t o +3 -2 t o 3.5 exc: RTS-31-4

( -7)

< 1 exc: ZPR6/6A

(-1)

f8 / f5 : -4 t o 8 exc: JEZEBEL + 4

(-8)

Power P r o f i l e : All s e t s p red ic t the power i n most p a r t s of the core within 2%, i n regions of s t rong f l u x g rad ien t s t o about 3% (conventional c o r e s ) , increas ing up t o 5% i n heterogeneous cores. In using ENL)F/B-IV data and KfKINR, a r a d i a l t i l t i n C/E i s observed: overes t imat ion of the power i n the outer core regions r e l a t i v e to near cen te r regions: t h e t i l t is enlarged i n heterogeneous cores up t o 5%.

and reliability of the coarse group results compared to reference

results of corresponding fine group calculations with uncollapsed

group constants. This is for instance very advantageous in the

calculation of the Na-void-effect with 2- or 3-dimensional reac-

tor models.

To improve the calculation of resonance self-shielding on the

basis of multilevel cross sections, a critical comparison was made

between D. Cullen's SIGMA1 code for Doppler broadening of tabu-

lated, linearly interpolable cross sections / 8 / and the DOBRO code

developed at KfK for calculation of Doppler broadened multilevel

cross sections directly from resonance parameters / 9 / . The test

problem involved calculation of 3-channel Reich-Moore cross sec- 0

tions for Pu-241 broadened to 900 K. The CPU time needed was

shorter and the precision higher with DOBRO, where interpolation

errors do not occur, than with SIGMAI.

In the area of fast reactor safety investigations much effort was

spent to determine an upper limit for the mechanical energy re-

lease in HCDAs for the prototype reactor SNR-300 together with

corresponding risk analyses. These studies were performed at the

request of the Committee on Future Energy Politics, set up by the

Parliament of the Federal Republic.

The experimental validation of complex safety code systems like

SIMMER is pursued. As an example, the experimental modelling of

the movement of molten clad material under the drag forces of Na-

vapour could be described theoretically satisfactorily with respect , to pressure lossses and material distribution.

b

2. Fuel Cycle Analysis for PWRs

The methods and data validation for the analysis of the PWR fuel

cycle was documented /lo/. At present, the activation of the end- pieces of PWR fuel elements is under investigation; results from

calculations will be compared to experiments. Studies on pluto-

nium and uranium recycling in thermal reactors are in progress.

As a special aspect, the neutron source density in a block of

vitrified high active waste (HAW) in boron-silicate glass was

determined. The knowledge of the neutron source is important for

shielding during vitrification and during transport of HAW in

the glass product. Boron-silicate glass was selected at KfK be-

cause of its high durability and leach resistance. The neutron

emission originates from spontaneous fission in heavy nuclides,

and from (cl,n) processes on light elements as boron, oxygen and

others. The investigation was performed for vitrified PWR-HAW.

The glass contains about 57 % Si02, 14 % Na20 and 12 % B203 /11/.

The determination of the (a,n)-neutron source was performed on the

basis of recently measured neutron yields from (a,n) processes in

thick targets /12/. The result is that the (a,n)-neutron source

density in the glass is determined to about 80 % by (a,n) pro-

cesses on B 0 It should be noted that the (a,n) neutrons have 2 3' energies mainly between 5 and 5.5 MeV! In addition, the contribu-

tion from (ct,n) processes dominates the spontaneous fission neu-

trons. When Cm-242 has decayed after about 5 years, the (a,n)

high energetic neutron source is still about 50 % of the total

neutron source, originating from alphas of Am-241 and Cm-244.

A reduction of the neutron emission can be obtained, if the B203

concentration is reduced, but then the long-term quality of the

glass product might be influenced unfavorably. This topic will be

pursued.

3. Studies on Advanced PWRs (Tight Lattice Cores)

The intercomparison of homogeneous and heterogeneous tight lattice

PWR-configurations have been pursued. As the nuclear data basis,

KEDAK-4 has been adopted. A variety of about 60 critical assemblies

(fast and epithermal systems) have been calculated. The agreement

in the criticality prediction is satisfactory for the purpose of a

consistent analysis of the feasibility studies envisaged. An impor-

tant constraint of the investigations is to guarantee a sufficient-

ly negative coolant void reactivity feedback for a tight lattice

PWR, so that normal PWR licensing procedures can be applied. To

study this effect more deeply, operational transients without scram * - r.

91100056

(ATWS) were investigated with a varying coolant density reactivity

feedback in a homogeneous tight lattice core. It was found that

only in a widened lattice, i.e. enlarging the moderator to fuel

volume ratio from about 0.5 to 0.6 - 0,7, a sufficiently negative

reactivity feedback can be guaranteed. This has the penalty

of decreasing the conversion ratio, but to a still acceptable

value of 0.9. A consistent comparison of all interesting concepts

is in progress; thermal hydraulic investigations concentrate on

the improvement of the theoretical basis for tight lattice cores

as well as for LOCA and reflood conditions.

A collection of papers on the tight lattice PWR is presented

separately to this meeting / 1 3 / ,

/I/ F.Helm, G.Henneges, W.Maschek,

Measurements and Computation of the Neutron Physics

Effects of Accident-Caused Core Distortions in LMFBRs.

Submitted for publication in Nucl. Sci. Eng. (1983);

NEACRP-A-Report, 26th Meeting (1983).

/2/ H.Kiisters,

Validation of Neutronic Calculations for Distorted Core

Configurations Arising in Accident Situations of LMFBRs.

Summary Report from the 25th Meeting of NEACRP (1982)

to CSNI; NEACRP-A-Report, 26th Meeting (1983)

Reactivity Effects of Fuel Rearrangement in Fast Reactor

Rod Bundles, ANS-Winter Meeting (1983);

NEACRP-A-Report, 26th Meeting (1983)

/4/ H.Giese, S.Pilate, J.M.Stevenson,

Control Rod Worths and Interactions in Fast Reactors.

Submitted for publication in Nucl. Sci. Eng. (1983);

NEACRP-A-Report. 26th Meeting (1983)

/5/ G.Humbert, F.Kappler, M.Mortini, G.Norvez, G.Rimpault,

B.Ruelle, W.Scholtyssek, A.Stanculescu,

e* Parametric Studies for the Heterogeneous Core Concept

in the Framework of the PRERACINE and RACINE Programs

Submitted for publication in Nucl. Sci. Eng. (1983);

NEACRP-A-Report, 26th Meeting (1983)

/6/ H.Kiisters, S.Pilate,

The Present Accuracy of Physics Characteristics of

Unirradiated Fast Reactors.

To be published in Annals of Nucl. Energy;

NEACRP-A-Report, 26th Meeting (1983)

/7/ E.Kiefhaber,

Application of Real, Adjoint and Bilinear Weighting

for Collopsing Group Constants Used in Space Dependent

Neutron Diffusion Problems;

KfK-Report 3430 (1982)

/8/ D.E.Cullen and C.R.Weisbin,

Nucl. Sci. Eng. - 60 (1976) 199

/9/ F.H.FrEhner,

Proc. Conf. on Nucl. Data Eval. Meth. and Procedures,

BNL 1980, ~NL-NCS-51363.J?981)1 ~ 0 1 . 1, p- 375, also

available as KfK 2388, Karlsruhe (1980)

/lo/ U.Fischer, H.W.Wiese,

Verbesserte konsistente Berechnung des nuklearen Inventars

abgebrannter DWR-Brennstoffe auf der Basis von Zell-Abbrand-

Verfahren mit KORIGEN,

KfK-Report 301 4 (1 983)

/11/ H.W. Wiese,

private Communication (1983)

/12/ G.J.H. Jacobs,

Neutron Energy Spectra Produced by a-Bornbardement of

Light Elements in Thick Targets.

Ph.D. Thesis. University of Eindhover. (1982)

/13/ Investigations on a Tight Lattice PWR in the Federal Republic

of Germany,

NEACRP-Report, 26th Meeting (1983)

- 57 -

11. REACTOR PHYSICS ACTIVITIES AT KRAFTWERK UNION

1. Computational Methods for LWR Analysis

Recent Developments in Nodal Reactor Analysis

The progress that has been made in the past few years in

the rapidly expanding field of nodal reactor analysis

methods is discussed in some detail in a review paper 1-1 7 - - which was presented at the 1983 ANS Topical Meeting in

0 Salt Lake City.

Although the transverse nodal methods appear now to be

well established a close examination reveals a number of

weak points which require special attention. The extremely

accurate results that have been obtained - 1-2,3,4 - 7 with transverse nodal methods on an assembly size mesh for

several two- and three-dimensional benchmark problems, un-

fortunately, are not fully representative for practical

LWR applications. This high degree of accuracy is obtained

for reactors with a fresh initial core, but not necessarily

for depleted cores, that is, for situations in which the

assumption of nodewise constant cross sections is not fuL-

0 filled. As a consequence, systematic errors can accumulate

during burnup which for PWRs may become as large a 10 % in the nodewise power distribution at the beginning of a new

reload cycle. It is shown that a nonlinear extension of the

nodal expansion method - 1-5 - 7 allows to nearly eliminate

these errors in a relatively straight-forward way with only . a small increase in computing time.

Another basic problem is the fact that the existence of node-

wise constant homogenized group diffusion theory parameters,

which usually is taken for granted in nodal methods, is

questionable. Furthermore, the fact that important local in-

formation is lost when dealing only with node averaged quanti-

ties is a severe weakness of the nodal method. It is not

surprising, therefore, that these questions currently receive

increased attention. It was the purpose of the paper 1-1 7 - - to review the more recent attempts to overcome these diffi-

culties. In particular, the work on nodal methods has led

to new ideas on consistent procedures of homogenization and

dehomogenization - 1-6-8 - 7. The "equivalence theory" is des- cribed as an exact method of homogenization and as the theo-

retical basis for the development of new and powerful approxi-

mate methods of homogenization - 1-7-9 - 7. Closely related to this is the inverse problem of reconstructing detailed flux

and pin power distributions for the heterogeneous assemblies

throughout the core ("dehomogenizationn). In a number of

recent papers 1-1,8,10 7 it has been shown that this can be - - 0

done with remarkable accuracy.

Although many refinements are still needed it is very likely

that the advanced transverse nodal methods, combined with

consistent methods of homogenization and dehomogenization,

will allow order of magnitude savings without the need for

making concessions in accuracy and spatial detail compared

to conventional full-core fine mesh finite-difference proce-

dures. These new methods are theoretically well-founded and

have the potential for including transport theory approxi-

mations in a consistent manner.

Considerable progress has als been made in the development

of nodal schemes for the solution of the two-dimensional

discrete ordinate equations - 1-11-14 - 7. All these nodal trans- port metnods hive in common that a spatial integration over

the transverse dimension of the node is performed which con-

verts the 2-D transport equation to a coupled set of one-

dimensional discrete ordinate equations. Suitable approxi-

mations are then made to represent the one-dimensional spatial dependence of the node-interior sources and the angular fluxes

a l o n g t h e e d g e s o f t h e node . The h i g h e r o r d e r n o d a l schemes ,

w h i l e c o m p u t a t i o n a l l y more c o s t l y p e r mesh c e l l , a l l o w t h e

u s e of a much c o a r s e r mesh t o o b t a i n r e s u l t s o f comparable

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

and s t o r a g e r e q u i r e m e n t s . However, improved methods o f con-

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

b e n e f i t o f t h e s e new'nodal t r a n s p o r t schemes. I t may b e o f

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

method d e s c r i b e d i n r e f e r e n c e 1 was s u c c e s s f u l l y a p p l i e d t o

speed up t h e n o d a l d i s c r e t e o r d i n a t e c a l c u l a t i o n s w i t h t h e

program MULTIMEDIUM f o r a number o f PWR a s sembly problems 1 - 1 7 . - -

The Statistical Analysis for the

Minimum critical Power Ratio

A critical power ratio of CPR > 1 protects the rods

in a fuel bundle against the effect of boiling transitions.

The uncertanties in the burnout-correlation (XL-correlation),

in fuel characteristics and in the reactor conditions however

yield a small probability for a rod with a nominal CPR > 1

to have really CPRcl. The statistical analysis of the thermo

hydraulics of a BWR determines the mean number of rods being

subject to boiling transition in dependence of the minimum

critical power ratio (MCPR) in the core.

An analytical code, VASKA, has been developed to carry out

these calculations[l5~. - Fig. 1 shows the results of a

typical statistical ana1ysis.b~ the analytical model in

comparison to corresponding results yielded by a Monte Carlo

code.

Basic input for such an analysis is a "pessimistic" power

distribution in the BWR core. The analysis yields a ulalue

MCPR99 .9 , a minimum critical power ratio of the core at which a mean of 99.9 % of the rods in the core is protected against burnout. The operational MCPR has to be adjusted

in such a way, that even during the worst transient the

value of MCPR 99.9

cannot be reached. It has been shown,

that, starting from a realistic power distribution, even

durihg the worst transient none of the allowed 0.1 % rods will be subject to boiling transition.

Fig. I Mean number of rods (%) being subject to boiling transitions in dependence of the minimum critical

Knmmr(rUtim power ratio MCPR

2 . I m p r o v e d D e t e r m i n a t i o n o f t h e Power D e n s i t y D i s t r i b u t i o n

by Means o f Gamma-Sens i t ive T r a v e r s i n g I n - c o r e P r o b e s

i n B o i l i n g Wate r R e a c t o r s

Power d i s t r i b u t i o n s i n b o i l i n g w a t e r r e a c t o r s (BWR's) a re

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

f r o m t r a v e r s i n g i n - c o r e p r o b e s ( T I P ' S ) . If g a m m a - s e n s i t i v e

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

n e u t r o n T I P ' s , d e t e r m i n a t i o n s o f power d e n s i t y d i s t r i b u t i o n s

i n a b o i l i n g water r e a c t o r a r e improved s i g n i f i c a n t l y . The

r e a s o n i s t h a t s i g n a l s o f gamma T I P ' s a r e l e s s d e p e n d e n t upon

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

t h e w a t e r g a p

The r e l i a b i l i t y o f KWU-gamma T I P ' s a n d t h e e x p e r i m e n t a l v e r i f i -

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

German BWR's (KWW C y c l e - 6 , KKB C y c l e - 2 , KKPl C y c l e - 2 ) . C o m p a r i s o n s

o f p r o c e s s c o m p u t e r power d i s t r i b u t i o n c a l . c u l a t i o n s p e r f o r m e d

p r i o r t o a n d f o l l o w i n g i n s t a l l a t i o n o f g a m m a - s e n s i t i v e T I P ' S

i n d i c a t e i m p r o v e d p l a n t o p e r a t i o n m a r g i n s . T a b l e 1 s u m m a r i z e s

t h e c h a n g e s t h a t o c c u r e d as a r e s u l t o f t h e gamma T I P i n s t a l l a t i o n . a

" . .-

Core Thermal Power

I Core ~ l o v

Radial P o w Factor

Total Peaking I Factor

6aa T I P -

41.S

35.9%

1.95

1.39 -

Rmal TIP - 95.35

112.4%

1.87

1.43 --

Gsa T I P -

93.5%

112.5%

1.76

1.34 -

Rmal TIP

6aa T I P -

99.6%

1Ol.aX

1.98

1.30 -

- 63 -

3. Advanced BWR Core Performance Calculations

The objective of the Advanced BWR Core Performance Calculations

(Fortschrittliche Nuklearrechnungen FNR) is to provide improved

capacity factors and margins of safety of BWR operations. The

additional capabilities are based upon accumulated experience

of operation and licensing of BWR plants.

The Advanced BWR Core Performance Calculations are performed

independent of the NSS process computer core performance monitoring

software on a high speed minicomputer located in the BWR plant.

For FNR-system the 32-bit CPU 3287 which is an integrated part

of Siemens 300 systems for Process Automation is used. The mini-

computer CPU 3287 is produced by Gould Inc., SEL Computer

Division.

The minicomputer receives on-line operating data from the

cess computer or redundant data acquisition system at peri

Systems

pro-

odical

time intervals permitting real-time operating data to be com-

bined with current state-of-the-art analysis and color graphic

display capabilities.

The FNR software consists of the PREDICTOR, which is based upon

the 3D-BWR simulator RS3D, and a high speed version of the pro-

cess computer periodic core performance evaluation program P I .

The main purpose of the PREDICTOR is to predict future operating

core states, the effect of changing the values of existing para-

meters, and definition of core performance and characteristics

during transient operations. The PREDICTOR system will be able

to improve the efficiency of reactor startups maneuvers and

steady state operation. It is expected that significant improve-

ments in capacity factor are possible.

The BWR simulator module of the PREDICTOR uses diffusion theory

which is valid over the range of conditions involved. Further,

a diffusion theory based model is used to adapt the simulated

TIP readings of 3D-BWR simulator calculation to the measured TIP

readings. This optimizes the accuracy of the BWR simulator

calculations of the TIP readings without changing basic system

parameters and allows the flexibility of accurately evaluating

changed core conditions (e.g. exposure, core flow or pressure,

control rod positions, etc.).

Communication capabilities will permit simple commands and out-

put results. Normal terminal input is complemented by reactor

and cycle specific restart disc files which are input from mag-

netic file. Output is normally on color graphics display terminals

or line printers and permanent disc file, and magnetic tapes are used

for large volume output.

R e f e r e n c e s : -

M. R . Wagner, K. Koebke, " P r o g r e s s i n Nodal Reac- t o r A n a l y s i s " , Proceed . o f 1983 ANS T o p i c a l Meet ing , Vol. 2 , p. 941 , March 28-31, 1983, S a l t Lake C i t y , USA

1-2 T - - H. Finnemann, F. Bennewitz , M. R . Wagner, Atomkern- - e n e r g i e , - 30, 123 (1977)

1-3 7 - - M. R . Wagner, H. Finnemann, K. Koebke, H . - J . W i n t e r , Atomkernenerg ie , - 30 , 129 (1977)

1-11 7 - - J . J. Dorning , "Modern Coarse-Mesh Methods - A Deve- lopment o f t h e ' 7 O t s " , Proceed . o f ANS T o p i c a l Meet ing on "Computa t iona l Methods i n Nuc lea r E n g i n e e r i n g " , Vol. 1 , p. 3-1, A p r i l 23-25, 1979, W i l l i a m s b u r g , V i r g i n i a

1-5 7 - - M. R . Wagner, K. Koebke, H.-3. Win te r , " A N o n l i n e a r E x t e n s i o n o f t h e Nodal Expansion Method", Proceed . o f I n t e r n a t l . ANSIENS T o p i c a l Meet ing on "Advances i n Mathemat ia l Methods f o r Nuc lea r E n g i n e e r i n g Problems", Vol. 2 , p. 43, A p r i l 27-29, 1981, Munich, FRG.

1-6 7 - - K. Koebke, " A New Approach t o Homogenizat ion and Group Condensa t ion" , IAEA-TECDOC 231 , p. 303, I A E A T e c h n i c a l Comm. Mtg. Lugano, S w i t z e r l a n d (Nov. 1978)

- 1-7 7 - - K. S. Smi th , A. F. Henry, and R . A . L o r e t z , "The D e t e r -

m i n a t i o n o f Homogenized D i f f u s i o n Theory P a r a m e t e r s f o r Coarse Mesh Nodal A n a l y s i s " , Proceed . o f ANS Topi- c a l Meet ing , Sun V a l l e y , Idaho , S e p t . 14-17, 1980, p. 294

. 1-8 7 - - K. Koebke, "Advances i n Homogenization and Dehomogeni- z a t i o n " , Proceed . o f I n t e r n a t l . ANSIENS T o p i c a l Meet ing , Munich, A p r i l 27-29, 1981, Vol. 2 , p. 59

A . Y . Cheng, C . L . H o x i e , A . F . H e n r y , "A Method f o r D e t e r m i n i n g E q u i v a l e n t Homogeneous P a r a m e t e r s " , P r o c e e d . o f I n t e r n a t l . ANSIENS T o p i c a l M e e t i n g , Munich , A p r i l 27 -29 , 1 9 8 1 , Vo l . 2 , p . 3.

H . S . K h a l i l , P. J . F i n c k , A . F. H e n r y , "Recon- s t r u c t i o n o f F u e l P i n Powers f r o m Nodal R e s u l t s N , P r o c e e d . o f 1 9 8 3 ANS T o p i c a l M e e t i n g , Vo l . 1 , p . 3 6 7 , March 28-31 , 1 9 8 3 , S a l t Lake C i t y , USA

M . R . Wagner , " A Noda l D i s c r e t e - O r d i n a t e s Method f o r t h e N u m e r i c a l S o l u t i o n of t h e M u l t i d i m e n s i o n a l a T r a n s p o r t E q u a t i o n " , P r o c e e d . o f ANS T o p i c a l M e e t i n g , W i l l i a m s b u r g , V i r g i n i a , A p r i l 23 -25 , 1 9 7 9 , Vo l . 2 , p . 4-117

R . D . Lawrence a n d J . J . D o r n i n g , " A D i s c r e t e Nodal I n t e g r a l T r a n s p o r t T h e o r y Method f o r M u l t i d i m e n s i o n a l R e a c t o r P h y s i c s a n d S h i e l d i n g C a l c u l a t i o n s " , P r o c e e d . o f ANS T o p i c a l M e e t i n g , Sun V a l l e y , I d a h o , S e p t . 14 -17 , 1 9 8 0 , p . 8 4 0 .

W . F. W a l t e r s , R . D. O ' D e l l , "Nodal Methods f o r D i s - c r e t e - O r d i n a t e s T r a n s p o r t P r o b l e m s i n (x,y) G e o m e t r y u , P r o c e e d . of I n t e r n a t l . ANSIENS T o p i c a l M e e t i n g , Munich , A p r i l 2 7 - 2 9 , 1 9 8 1 , Vo l . 1 , p . 115

R . E . Pevey a n d H . L . Dodds , J r . , T r a n s . - Am. N u c l . 3 9 , 751 ( 1 9 8 1 ) - - M . S c h r a d e r , J a h r e s t a g u n g K e r n t e c h n i k , B e r l i n , J u n e 1 4 - 1 6 , 1983 Compacts p .121

111. REACTOR PHYSICS ACTIVITIES

AT THE UNIVERSITY OF STUTTGART (IKE)

1 Three-dimensional reactor burn up calculations for PWR

(A. Warner, D. Lutz, K. Neumann, W. Bernnat)

ENDF/B-IV/V cross section data and neutron physics methods available

at IKE have been applied to calculate burn up-dependent microscopic

two-group cross sections. With these two group constants also

depending on moderator density, fuel temperature and boron density

3D-reactor burn up calculations for two cycles of an operating PWR

have been carried out. The deviation from calculated to measured

cycle length was + 2,6 % (case A) and - 0,4 % (case B) . For one of these cycles (case B) also 2D-calculations were performed with

several buckling concepts. But even for the best 2D-case with burn

up-dependent bucklings from the previous 3D-calculation the

difference was + 4,4 % between calculated and measured cycle length

in comparison to - 0,4 % for the 3D-calculation.

2 2/30 cell calculation method with the pik-concept

(K. Neumann)

To consider more dimensional effects by the spectral calculation for

preparation of few-group cross section datas a method is developed

on basis of the first collision equation. The first collision pro-

babilities, pikt used in the first collision equation, are statisi-

cally calculated with a Monte-Carlo method for more dimensional

geometries represented by the combinatorial geometry. The geometry ' allows many different subzones for example radial division of

pins, fine zones for absorber pins and real 3D arrangements. The

calculation of the collision probabilities is divided in two psrts,

the analysis of geometry and the actual pik-calculation. Because

of this division it is possible to use the same geometry infor-

mation for different material-zone arrangements, for example burn

up dependent problems or multigroup resonance calculations. The

calculation time istherefore, not too high. The method has been

verified by comparison with Benchmark results.

3 Radiation transport through cavities in twodimensional SN-Method -

(K. Neumann, D. Emendorfer, W. Bernnat)

The calculation of neutron- and photon transport is difficult with

the SN-method in twodimensional geometry including enlarged cavities.

For long and narrow cavities it is nearly impossible. Since other

methods, like Monte-Carlo, PN - or first collision source methods are not simple in application, an analytical transport method,

imbedded in a SN-transport code, is developed to calculate radiation

transport in cavities.

The directional distribution of neutron fluxes is approximated at

every boundary mesh of the single cavity in legendre polynoMals,

orthogonally in each quadrant. This polynomial-approximation allows e

the analytical representation of streaming between different mesh

elements. Especially,the calculation is exact in small cavities

for mesh elements far away from one another. The transfer coeffi-

cients of the approximation can be calculated before SN-calculation

and they do not depend on energy group. Couplinq flux moments into

cavity with transfer coefficients results in flux moments out of

cavity during S -calculation. N

Therefore, no higher SN-orders are needed. The method is realized for cylindrical and annular cavities in r,z-geometry. It is possible

to define several cavities.

The method is coupled with the SN-code DOT 4.2. The variable

attributes of the code are almost conserved.

4 Theoretical Analysis of the Energy Deposition Rate in the

Gamma Thermometer due to Neutron and Gamma Sources

(W. Bernnat)

A computational study was performed to investigate the total volu-

metric energy deposition rate in gamma thermometers as a function

of the power distribution in the fuel elements. The study is divi-

ded into a radiation transport analysis for the fast neutrons and

gammas and in the determination of neutron- and gamma sources in

the surrounding fuel pins of the gamma thermometer.

The determination of the energy deposition rate in the gamma ther-

mometer due to neutron and direct or (n,y)-sources in neighbouring

pins was performed by adjoint Monte-Carlo transport calculations

for fast neutrons and gammas.

Knowing the adjoint neutron- or gamma flux in a pin surrounding

the gamma-thermometer, the contributions of this pin to the energy

deposition rate is simply the product of the volume source (neu-

tron or gammas) in the pin and the adjoint flux (integrated over

the energy).

ITALY

26th NEACriP PFETING

Oak Ridge, Oct. 17-21,1983

Reactor Physics Act ivi t ies in I t a ly

Conpliled by R. W t i n e l l i , ENEA

Ln the framework of the Research and Development Agreement w i t h

the French CEA, f a s t reactor a c t i v i t i e s a t ENEA have been concen - t ra ted mainly on the interpreta t ion of in tegra l experiments

( X % X E , ITPERTITI a ~ d JASOY Propames ) and on the validation

of t he f o m d a i r e s specif ic t o core, blanket and shielding calcu - l a t i s n s (including data evaluations and adjustments) .

1.1. Core in tegra l experinients (R&CI%)

3.e ?eas~~e!?enC,s on configurations ID and 1E ( sodim void e f fec t s

a m x 5 s i x l z t e d c m t r o l rods, and reac t iv i ty and power dis t r ibu-

t lons associated t o off-center absorbers, respectvively) have been

prab;zrl ly cc@efed /1,2/.

3 e s e ex?eri%nts !-me been designed, carried out and interpreted

according t o the methodology established a t CEA for t he extrapola-

t ion of integral p a r m t e r s -and of t h e i r uncertainties- t o the de - sim of large power reactor cores 3 . New experiments a r e planned

on tha t basis , aimed at reducing the. uncer ta int ies on the react ivi-

t y losses per cycle due t o burnup and buildup of heavy isotopes 141.

Much a t ten t ion in the MCINE p r o p a m e has been placed on problem

re l a t i ng t o the s ta r tup of SWERPHENIX 1 (Source range instrumenta-

t ion response, control rod cal ibrat ion methods, e tc . ) . Tne RACINE 1-S

e x p e ~ k n t is presently being run /5/, which s i m l a t e s a chessboard

configuration of the SPX-1 core. !he experimental program inclu-

des measurements of radial power distributions, reactivity worths

of diluent subassemblies and subcriticality margins up t o about

30 dollars.

1.2. Blankets

The parametric experbental study of simulated BOC blanket regions

(Prorgramne NEr'ERTITI i n the fast source reactor TAPIR0 of ENEP., Ca-

saccia) has been completed. The analysis of th i s large se t of expe-

riments i s underway.

A supplementary configuration, a mxk-up of an Em blanket (enriched

U-235 pins simulating Pu buildup) w i l l be studied in the next months.

1.3. Shielding

Tne version. Nr.1 of the adjusted shie ldhg formulaire PROPANE has

been implemented and i t s satisfactory performance extensively test-

ed / 6 ,7 / . A new version i s already under development , supported by

JASON experimental programe i n HARMONE a t CEA, Cadarache.

1.4. Data evaluation, adjustmnt and processing

A new version of the processing code FOUR ACES has been implemented

by ENEA (Boloma). Partial re-evaluations have been made of fission

product and transactinide isotopes; most of the f i l e s evaluated i n

the framework of CEA-ENEA cooperation have been recomnded for adop

%ion in the f i r s t version of JEF.

(The I tal ian contributions t o JEF act ivi t ies include an effort of

c r i t i ca l analysis, intercomparison and selection of available f i l e s

also for&corbinq nuclides such as Gd and H f , and for structural

materials such as Zr).

Finally, an adjustment of the effective cross-sections of actinides

has been completed /8/, basing upon the results of irradiation ex-

per&nts in PHENIX.

Other a c t i v i t i e s , not connected with CEA-ENEA a g e e m n t , are i n progress

at ENEA ( b l o r g a ) .

1.5. Neutronic d e s i m of PEC core

The analysis of t he reference configuration (78 subassemblies) has 4-

been completed. I n par t icu la r , t he tra-isport, heterogeneity and

mesh-size corrections have been recalculated f o r control rod worths,

leading t o a 13% reduction with respect t o t he worths calculated

v:a standard d e s i m routes. The s l i g h t value ( l e s s than 5%) of rod

interact ion factors has been conf i m d ,

Problems rela ted t o t he start-up of t he reactor have been tackled,

and the design of in-pile measurements devices is being developed.

1.6. Transport and improved diffusion methods

A generalized Fourier tramsport (GFT) mthod has been recently pro-

posed /9/ t o t r e a t sca t te r ing anisotropy in homogeneous spherical

geo?-etry, i n vieii of spectrum and shielding problems. A Fortran f o r - nulation and a G W extension t o heterogeneous media a r e i n progress.

L ~ r o v e 3 diffusion models based on the def ini t ion of size-dependent

SifPdsive pZmeters have been proposed f o r op t ica l ly srnall f a s t

syste-5 wi th di f ferent geometries / lo / .

Work i s s t i l l in progress a t EXEL-CRTN (Milano), aiming a t develop-

ment and validation of synthesis and polynomial nadal method (PNM)

codes f o r f a s t and slow dynamics analysis of LWRs.

2.1. Codes f o r 3-D Transient Analysis

In recent years, QUANDRY-EN /11/, developed at CR?N on t h e basic

s t ructure of EPRI's coarse-mesh code QUANDRY, has played a key r o l e

in assessing the accuracy of the horn made synthesis code SYNTH-C

1 i e i n defining the e r ro r bounds of t he f i n a l solution.

Comparison were made on the basis O f node or assembly-averaged

values. The most recent development of QUANDRY-EN by CRlN, includ - ing :

- the introduction of generalized equivalence theory concepts for

the homogenization of large nodes (proving that the new scheme

converges t o the same solution as a fully heterogeneous diffb-

sion calculation) ;

- the application of X-W poynomial interpolation methods /13/ t o

the converged solution, in order t o determine detailed flux and

power distributions inside homogeneous nodes;

- the adoption of a more sophisticated thermal-hyrlraulic module

constitute a significant improvement of the performance of t h i s

coarse-mesh code and add t o its potential as a tool for SYNTM-C

valid ation.

2.2. Depletion codes: NORMA-3D

The coarse-mesh NORMA-3D /14/ has been implenented by CFTN t o si-

mulate the long-term neutronic and thermal-hydraulic behaviour of

large PhTls in a tridimensional scheme.

The code is a development of NRMS, a code based on PNM.

The polynomial nodal approach was chosen because:

- the converged solution contains the necessary information t o de-

termine directly the detailed power distribution inside each node

by the polynomial interpolation method developed by Koebke and

Wager

- the PNM is able t o deal easily with several neutron groups.

After extending WiMS t o 3 dimensions, CRTN implemented various addi - tions t o the basic structure, namely:

- a diagonal and rotational geometry

- a new treatment of the transverse leakage and rebalance schemes

- an implicit treatment of the burnup

- a simplified thermal-hydraulic model

- the determination of the pin power d i s t r ibu t ion

- the mnagement of the fue l assemblies.

A t present, work i s i n progress i n order t o enable NORMA-3D t o deal

with the "equivalent parameters" derived from the "equivalence

theory" and with a more complex thermal-hydraulic model f o r the

deternination of t he the& limits of the core. A deta i led acc-

ount of t he computational accuracy ard efficiency of PMul and of

NCIFYA-3D i n par t icu la r , is given i n /15/.

2.3. Measurements a t Caorso BhR

A s e r i e s of d y m ~ L c s t e s t s during the second fue l cycle of the 900

VKe Caorso B! power s ta t ion have been planned and licensed. Such

t e s t s , conducted by nE.4 and AJVSU, are scheduied t o start at the

end of October 1983, and w i l l include s t a b i l i t y analysys studies.

A measurement c w a i g n aimed a t c h a r a c t e r i z h the neutron environ-

ren t inside the p r i m r y containment of the plant , was s t a r t ed i n

P k x h 1933. This cmpaign is based - l ike a previous one made i n

?3?1 inside the reactor drywell- on the use of the Multiple Foi l

Activatior! tec:mique. Seven i r rad ia t ion posit ions have been

chase2 f s r the instr-mented s ta t ions , designed and in s t a l l ed by

EEL-C.?": i n cooperation with CEShTF (Milano) 6 The ar,alysis

of the r e s u l t s obtained a t t he end of cycle 2 w i l l a l so be aimed

a t the validation of CFETN's shielding calculation methods, thus

coni?leted the information drawn from the calculation-experiment

comparisons made a t the conclusion of the f i r s t campaign /17/.

3. CRITICALITY SAFETY

A considerable calculational e f f o r t has been spent at ENEA, Fuel

Cycle Dept., in producing and t e s t i ng reference l i b r a r i e s ( 219

and 123 groups, AMPX-2 ) and i n val idat ing Montecarlo codes

( mainly KENO-IV ) f o r calculations related t o c r i t i c a l i t y safe-

t y of reprocessing plants ( storage pools and final product

c e l l s ) and t o spent f i e 1 transport cask safety ver i f icat ions.

More than s ix ty c r i t i c a l and subcr i t ica l experiments have been

analyzed i n the process of l ib rary and code Validation : the most

recent r e s u l t s of t h i s work a r e reported i n 1181.

A s a contributior, t o the second par t of CSNI Benchark on fuel

transport casks / ? 9 / , ENEA (ihlogna) has acclflately investigated

t h e e f f ec t s on k values of different ways of modelling the e f f (very) complicated array and of the spa t i a l treatment of reson-

ances. Further contributions are planned i n t h i s context, i.e.

a study of the s t a b i l i t y of Montecarlo solutions f o r large

arrays and of the e f fec t of packaging material.

4 . NOISE AVALYSIS

The analysis, p e r f o m d by ENEA Casaccia, of LPRM and TIP simal

fluctuations i n Caorso a t fill power, indicate the presence of a

second t r a n s i t t i m e of the coolant flow i n the upper half of the

core. The implications i n terms of separation i n the d is t r ibu t ion

of coolant flow ve loc i t ies a r e being studied. Preliminary con-

clusions a re presented in /20/.

Acoustic and t h e m 1 noise a r e being analyzed in other in-pile

and out-of-pi'le experiments, namely :

- vibrations of instrumented tubes and of recirculat ion pumps

in Caorso ( by ENEA Casaccia )

- vibrations of ro ta t ing machinery, l i k e a SPX-1 pump shaf t i n

a sodium loop ( by ENEA B o l o g ~ ~ )

- t e q x r a t u r e f luctuat ions of conventional and i n t r in s i c , fast

response (SS-Na) themcoup les on a U F B R fue l pin i r rad ia ted

in a sodium loop at SILOE ( CEA Grenoble ), f o r ear ly detection

and follow-up of the evolution of f a i lu re s in the pin ( by

ENEA B o l o ~ ~ l a 1.

M.mtini, P r i v a t e conn iun icat ion.

/2/ U.Broccoli, P r i v a t e communicat ion.

/3/ G.Palmiotti, M.Salvatcres, P r i w t e comm~nisa:io'.

/4/ G.Palmiotti, M.Salvatores, p r j va!-e comn?kniia.cjon.

/5/ R.DeWouters, M.Martini, P r i v a t ? comrnunics.~ion.

,'6/ J.?.Trapp, A.DeCarli, 6th ICRC', Tokyo 1983

/7/'J.C. Cabrillat, V.Rado et al. , 6th ICRS, Tokyo 1983 /8/ G.Oliva, Pri va-ie comrnunica.tion.

:3/ G.Ghinassi et al., Nat .l Seminar on Reactor Physics,Bologna 1983

/lo/ F.Prernuda, RT/FI(83)4

/11/ E.Brega et dl., CRW-N5-21 (1979)

/12/ 5.S-ee;a et al., CRTN-N5-19 (1979)

/13/ K.K@bke, M.3.Wagner, Atomkernenrgie - 30 (136-142) 1977

/12/ E.3rega et al., CRm-N5/82/06

/15/ E.&ega et al., submitted to Annals of Nuclear Energy

1161 E.Sorioli et al., Ciim-N1/83/ ( i n press)

/I7/ F.8arbucci et al., 5th ICRS, Tokyo 1983

,'IS' ?.A.krdegro et al., 1nt.l Seminar on Criticality Studies, Dijon 1983

/I?.; CSZ Sestricted Report Nr. 78

/2C.' A.?eSerico et al., S.M. on In-Core Instrumentation, Halden 1983

JAPAN

Reactor Physics A c t i v i t i e s in Japan

(September 1982 - September 1983)

Compiled by

T. Asaoka ( J A E R I ) and K. Sh i raka ta (PNC)

Tkernal Reactor Physics

A core performance c a l c u l a t i o n program, COMS (Core Operation and Management System), was developed f o r p red ic t ing p resen t and fu tu r e power d i s t r i b u t i o n s a s a BWR on-l ine core management system t o meet with deaands on p l a n t a v a i l a b i l i t y and opera t iona l f l e x i b i l i t y . The program c o n s i s t s of a nodal cougling type nuclear-thermalhydraulic model and is adapted t o t h e cur ren t core s t a t e by using measured TIP ( t r ave r s ing in-core probe) da ta . The adaption algori thm has been v e r i f i e d through t 5 e s imulat ion, where roo t mean square (RMS) e r r o r s of c a l cu l a t ed TIP readings t o t h e measured ones a r e 5% in fu tu r e p r ed i c t i on and 3% in p resen t e s t imat i0n . l ) In add i t ion , t h e TAaS software package was developed a s an e f f e c t i v e on-line, on-s i t e t o o l f o r BWR core management. To obta in h ighly accura te nodal powers and t h e core c r i t i c a l e igenvalue, newly-developed methods were implemented t o automat ica l ly minimize e r r o r s included in in-core-neutron-flux-monitor readings. A s o r t of mac5ine-learning method was a l s o developed t o minimize t he e r r o r s caused by s imp l i c i t y of t h e physics models adopted i n TAFOS. Comparisons of TARMS t o experimental da ta , including those from gamma-scannings and TIP 'S , have shown RMS e r r o r s of aboct 3% in nodal powers ( t o t h e gamma scannings) f o r t he core monitoring, and of less than 4% ( t o t h e TIP 'S) i n t h e core performance p red ic t ion. )

On t h e o ther hand, a new i t e r a t i o n method fo r so lv ing one-group d i f fu s ion equation was developed f o r an e f f i c i e n t core performance p red ic t ion on t h e b a s i s of an ana ly t i c so lu t ion technique. Test c a l cu l a t i ons f o r LWRs have indica ted t h a t t he mechod can achieve a s i g n i f i c a n t r educ t ion i n computing-time and menory in comparison with t h e conventional f i n i t e d i f fe rence

A method. 3 )

The development of a s tandard computer code system fo r nuclear c a l cu l a t i ons , SRAC (Standard Reactor Analysis Code), has been completed and t h e benchmark c a l c u l a t i o n s have shown t h a t t h e system can p r ed i c t n i ce ly t h e experimental keff values fo r var ious types of c r i t i c a l a ~ s e m b l i e s . ~ ) On t h e o ther hand, t h e research of a d a p t a b i l i t y of vector processing t o large-scale nuclear codes has been proceeded t o t e s t vector ized versions of DOT-3.5, TWOTRAN and ANISM based on t h e f i n i t e d i f fe rence method, PALLAS-2DCY and BERMUDA on t h e d i r e c t in tegra t ion method, and s o on. The gain obtained from t h e vec to r iza t ion was inves t iga ted ;n r e l a t ' . n t o tl-.; . -umerical 1 -t\od, geometry and projlem typ~ .5 ! . 6 )

A s f o r t h e t e s t i n g of t h e d a t a and methods, t empera ture c o e f f i c i e n t s were ob ta ined from 1- and 2-dimensional c r i t i c a l i t y c a l c u l a t i o n s and 1-dimensional p e r t u r b a t ion c a l c u l a t i o n , and t h e r e s u l t s were compared wi th t h e measured va lues f o r 3 l oad ing p a t t e r n s i n a l i g h t - w a t e r cooled c o r e of K U C A . ~ ) I n a d d i t i o n , an e v a l u a t i o n model of f a s t neu t ron i r r a d i a t i o n dose i n JMTR was i n v e s t i g a t e d i n r e i a t ion t o t h e r a d i a t i o n damage of m a t e r i a l s by i n t r o d u c i n g exposure parameters such a s t h e displacement pe r atom and t h e damage f luence - 8 )

Concerning t h e FUGEN, a heavy-water moderated, b o i l i n g l i g h t w a t e r coo led , p r e s s u r e t u b e type r e a c t o r , an on - l i ne c o r e performance e v a l u a t i o n system ATROPOS was developed t o c a r r y ou t s a f e and e f f i c i e n t r e a c t o r o p e r a t i o n by o f f e r i n g d e t a i l e d u s e f u l informat ion on such i tems of c o r e performance a s t h e thermal power, power d i s t r i b u t i o n and thermal o p e r a t i o n l i m i t s . The system has been v e r i f i e d 5 us ing t h e s t a r t - u p t e s t d a t a from t h e FUGEN i n i t i a l c o r e . 9r On t h e o t h e r hand, a new c l u s t e r a n a l y s i s code MESSIAH was a p p l i e d t o c a l c u l a t e r e a c t o r phys i c s parameters measured i n t h e c r i t i c a l f a c i l i t y DCA f o r t h e FUGEN. The MESSI.4il code u t i l i z e s t h e c o l l i s i o n p r o b a b i l i t y method t o s o l v e t h e neutron t r a n s p o r t equa t ion . The c a l c u l a t e d r e a c t o r phys i c s parameters , e s p e c i a l l y t h e micro-sarameters, ag ree f a i r l y w e l l w i t h t h e exper iment , bu t t h e c a l c u l a t e d void r e a c t i v i t y i n d o l l e r u n i t i s s l i g h t l y smal le r than t h e exper imenta l value a s shown in F ig . 1, which is probably a t t r i b u t e d t o an over -pred ic t ion of t h e d i f f u- s i o n c o n s t a n t . lo) I n addi- t i o n , t h e reackor phys i c s behav- i o r of Gd burnable poison f u e l p i n s was s t u d i e d through meas- urements of r e a c t i v i t y change, coo lan t void r e a c t i v i t y , l o c a l power d i s t r i b u t i o n and t h e r n a l neutron f l u x d i s t r i b u t ion a t t h e D C A . l l ) Furthermore, a f e a s i - b i l i t v s tudv on t h e use of t h o r i -

0 30 70 100

Votd ;%)

Fis. 1 Comparison of void reactivity ux th experiment

urn f u e l s was performed f o r t h e plutonium n a t u r a l uranium mixed oxide f u e l l e d FUGEN. The s tudy h a s i n d i c a t e d t h a t t h e use of Th f u e l s even i n a s imple form of on ly s t r a i g h t Th on a once-through c y c l e b a s i s , r e s u l t s i n many improvements i n r e a c t o r performance. 12)

For t h e i n v e s t i g a t i o n of n u c l e a r c h a r a c t e r i s t i c s of FUGEN HWR Demonstration P l a n t , parameter s t u d i e s on r e a c t o r phys i c s have been cont inued by us ing DCA. The i tems of t h e r e c e n t s t u d i e s a r e as fo l lows ;

(1) R e a c t i v i t y worth of B-10 s o l u t i o n h a s been measured. Boron-10 was f i l l e d i n a c y l i n d e r which s i m u l a t e s t h e gu ide t u b e of c o n t r o l r o d f o r t h e n u c l e a r s i m u l a t i o n of r a p i d i n j e c t i o n system of B-10 s o l u t i o n adopted i n t h e des ign of t h e Demonstration P l a n t . ( 2 ) Temperature c o e f f i c i e n t s of r e a c t i v i t y on bo th H20 c o o l a n t and D20 moderator have been measured f o r Pu l a t t i c e s by r a i s i n g t h e whole c o r e t emgera tu re up t o 80k. ( 3 ) Rad ia l thermal neu t ron f l u x d i s t r i b u t i o n has been measured on t h e whole Pu core .

1) Fukuzaki T., Mi t su t a T. e t a l . :"Core Performance C a l c u l a t i o n Program f o r On-Line Core Management", J. Atomic Energy Soc. Japan, 21, 639 (1983) ( i n J apanese )

2 ) T s u i k i M., Uematsu H. e t al.:"TRAMS : An On-Line Bo i l i ng Water Reactor Management System Based on Core Phys ics S imula to r " , Top ica l Meeting on Advances i n Reactor Computations, S a l t Lake C i t y (1983)

3 ) I t a g a k i M.:"Analytic S o l u t i o n Technique f o r So lv ing One-Group Di f fus ion Equa t ions f o r Core S imula t ions" , J. Nucl. S c i . Thechnol., 20, 627 (1983)

4 ) Tsuchihash i K . , Takano H. e t a l . :"SRAC: J A E R I Thermal Reactor S tandard Code System f o r Reactor Design and Ana lys i s " , J A E R I 1285 (1983)

5 ) I s h i g u r o M. and T s u t s u i T. :"Vector P roces s ing of t h e Neutron Transpor t Codes", JAERI-M 52-199 (1983) ( i n Japanese )

6 ) Harada H., Higuchi K. e t a l . : "Vector i z a t i o n of Nuclear Codes on FACOM 230-75 APU Computer", JAERI-M 83-024 (1983) ( i n Japanese )

7 ) Wakanatsu S . , Hashimoto K. e t a l . : " C a l c u l a t i o n of Tenpera ture C o e f f i c i e n t s f o r t h e Light-Water-Moderated Core of Kyoto Un ive r s i t y C r i t i c a l Assembly", J. Atomic Energy Soc. Japan, 24, 963 (1982) ( i n J a p a n e s e )

8 ) Sakura i F. a z Ni ibo T. : "Eva lua t ion Method of Fas t Neutron I r r a d i a t i o n Dose i n JMTR", J. Atomic Energy Soc. Japan , 25, - 372 (1983) ( i n J apanese )

9 ) N a t o r i H., Kaneto K. e t a l . :"Development of On-Line Core Performance Eva lua t ion System f o r FUGEN", J. A t o m i c Energy Soc. Japan, 24, 792 (1982) ( i n J a p a n e s e )

1 0 ) Kadotani H. and Hachiya Y. : "Analys i s of Heavy-Water- Moderated, Cluster-Type Fue l L a t t i c e s by C l u s t e r Phys i c s Code MESSIAH", J. Nucl. S i c . Technol., 2, 689 (1982)

11) Wakabayashi T. and Minatsuki I. : " C r i t i c a l Experiments on Gadolinium Poisoned Cluster-Type Fue l Assemblies i n Heavy Water L a t t i c e s " , Nucl. S c i . Engng., 83, 50 (1983)

1 2 ) Haga T. :"A F e a s i b i l i t y Study on U s e o f Thorium Fue l i n Pu MOX Fue l l ed FUGEN-HIJR", Japan-US Seminar on Thorium Fue l Reac tor , Nara (1982)

Fusion Neutronics

A t t h e Fusion Neut ron ics Source (FNS) f a c i l i t y , t h e angula r d i s t r i b u t i o n and t h e energy spectrum of t h e source neu t rons were measured f o r a water coo led t y p e t r i t i u m t a r g e t shown i n Fig. 1. The MORSE-GG code was used t o c a l c u l a t e t h e neutron t r a n s p o r t and t h e secondary gamma-ray emiss ion and t r a n s p o r t i n t h e t a r g e t assembly. A s seen from Fig. 2 , a f a i r l y good agreement between t h e exper imenta l r e s u l t and t h e c a l c u l a t e d va lue h a s been o b t a i n e d f o r t h e neu t ron s p e c t r a i n forward d i r e c t i o n t o t h e T beam l i n e ( 6=0•‹ ) . For t h e s p e c t r a a t o t h e r ang le s , however, some d i sc repancy is observed in t h e shape of t h e source peak and of t h e t a i l below 1.7 MeV, due probably t o an i n a p p r o p r i a t e c o n s i d e r a t i o n of t h e d e t e c t o r p o s i t ion and t h e p u l s e shape i n t h e a c c e l e r a t o r ope ra t i 0 n . l )

Fig. 1 Cross-iec:ior.zl view of, :zi;er asiembly

, , , . , . . , . r . . , , , 2 4 6 8 :O 12 14 16

NEUTRON ENEaGY I MeV I

NEUTRON ENERGY I MeV 1

Fig. 2 Measured a ~ 1 calculated neutron ::.-,ma at O', 60' and 105'

Angular dependent neutron leakage spec t r a from Li2O s l a b assemblies were measured a l s o at t h e FNS by using t he t ime-of-f l ight method2) and t h e d a t a were analyzed with a new d i s c r e t e o rd ina t e s code BERMUDA-2DS based on t h e d i r e c t i n t eg ra t i on method in a multigroup modele3) A s seen from Figs. 3 and 4, t h e ca l cu l a t ed s p e c t r a agree well with t h e observed values from t h e viewpoint of t h e absolute comparison. I n add i t ion , a t t h e OKTAVIAN f a c i l i t y , measurements of leakage angular neutron spec t r a from s l a b s of t y p i c a l sh i e ld mate r ia l s were c a r r i e d out by means of t h e t ime-of-f l ight t e ~ h n i q u e . ~ )

Fig. 3 Anole-de endent leakage spectra from Fig. 4 Angle-dependent leakage spectra from Lf10 Li iO slag assembly (5.06 crn in th~ckness) slab assembly (20.24 cm i n thickness)

Deta i l ed n e u t r o n i c s a n a l y s i s was performed on a tokamak fus ion exper imental r e a c t o r (FER) of a swimming pool t y p e shown i n Fig. 5. The Monte Ca r lo code MORSE-I, a modified ve r s ion of t h e MORSE-GG t o cons ide r t o r o i d a l geometry, asymmetric t o r u s c r o s s s e c t i o n and neutron source d i s t r i b u t i o n i n plasmas. As a r e s u l t of t h e a n a l y s i s , it becomes c l e a r t h a t a mod i f i ca t ion of t h e b l a n k e t s t r u c t u r e and t h e m a t e r i a l composit ion should be made t o improve t h e t r i t i u m b reed ing performance and t o reduce t h e n u c l e a r h e a t i n g r a t e of t h e vacuum v e s s e l i n t h e d i v e r t e r zone.5)16)

Fig. 5 Vertical cross section of STTR

h s tudy was made on t h e n u c l e a r c h a r a c t e r i s t i c s of t h e b l a n k e t / s h i e l d des ign of a D-D tokamak r e a c t o r . The g r a p h i t e b l anke t of 1 m t h i c h n e s s h a s t h e c h a r a c t e r i s t i c s of much smal le r r e s i d u a l r a d i o a c t i v i t y , af t e r h e a t and b i o l o g i c a l hazard p o t e n t i a l compared t o o t h e r material^.^ ) I n addt ion , n e u t r o n i c s a n a l y s i s was c a r r i e d o u t t o a s s e s s t h e t r i t i u m breed ing c a p a b i l i t y of t h e he l io t ron-H r e a c t o r des ign by us ing t h e LVISN code and t h e MORSE-I code.8) Furthermore, n e u t r o n i c s p r o p e r t i e s of a l a s e r f u s i o n r e a c t o r were d i scussed on t h e b a s i s of one-dimensional 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 burn ing D-T plasmas and b l a n k e t s a 9 )

I n a d d i t i o n , computa t iona l models f o r s p a l l a t i o n and f i s s i o n r e a c t i o n s were e v a l u a t e d f o r deve lop ing an a c c e l e r a t o r b reed ing and t r ansmuta t ion code system N M T C / J A E R I ~ ~ ) which performs t h e Monte C a r l o s i m u l a t i o n of nuc l ea r r e a c t ions i n a heterogeneous t a r g e t . By running t h e NMTC/JAERI code f o r t h i n t a r g e t s of B i , Pb, Th and U i n t h e energy range of 50-1000 MeV, p ro ton and neut ron n o r , - s l a s t i c and f i s s i o n ;-rl>ss s e c t i o n s were

d e r i v e d from the c o u n t s o f real c o l l i s i o n s and f i s s i o n e v e n t s i n t h e t a r g e t s , t o compare the r e s u l t s w i t h t h e e x p e r i m e n t a l da ta . l1 ) Fur the rmore , t h e e n e r g y s p e c t r a o f n e u t r o n s e m i t t e d by t h i c k t a r g e t s of C, Fe , Cu and Pb t o t h e i n c i d e n t of 30- and 52-MeV p r o t o n s were o b t a i n e d by u n f o l d i n g t h e p u l s e h e i g h t d i s t r i b u t i o n measured w i t h an NE-213 s c i n t i l l a t o r .12)

1) S e k i Y., Oyama Y. e t a l . :"Monte C a r l o C a l c u l a t i o n s o f S o u r c e C h a r a c t e r i s t i c s o f FNS Water Cooled Type T r i t i u m T a r g e t " , J. Nucl. S c i . Techno l . , 20, 686 (1983)

2 ) Maekawa H., Oyama Y. e t a l . :"Measurements of Angular F lux on S u r f a c e o f L i 2 0 S l a b A s s e m b l i e s and T h e i r A n a l y s i s by a Direct I n t e g r a t i o n T r a n s p o r t Code BERMUDA", ANS F i f t h T o p i c a l Meeting on Technology o f F u s i o n Energy (1983)

3 ) S u z u k i T., Hasegawa A. e t a l . :"BERMUDA-2DN: A Two-Dimensional Neutron T r a n s p o r t Code", S i x t h I n t e r n a t i o n a l Conf . R a d i a t i o n S h i e l d i n g (ICRS ) , Paper 3b-2 (1983)

4 ) Yamamoto J., Takahash i A. e t a l . :"Measurement and A n a l y s i s o f Leakage Neutron S p e c t r a from SS-316, C o n c r e t e , Water and P o l y e t h y l e n e S l a b s w i t h D-T Neutron Source" , S i x t h ICRS, Paper 4b-9 (1983)

5 ) Mori S., S e k i Y. e t a l . : " N e u t r o n i c s Design o f T r i t i u m Breeding Blanke t f o r F u s i o n E x p e r i m e n t a l R e a c t o r " , J. Nucl. S c i . Technol . , 20, 154 ( 1 9 8 3 )

6 ) Mori S. , Mohri K. e t a l . : "Xuc lea r A n a l y s i s of Blanket and S h i e l d Design f o r Tokamak F u s i o n Exper imenta l R e a c t o r " , S i x t h ICRS, Paper 5b-9 ( 1 9 8 3 )

7 ) Nakashima H., Tsukahara K. e t a l . :"Nuclear A n a l y s i s of B l a n k e t / S h i e l d Design f o r D-D Tokamak Fus ion Reac to r" , J. Nucl. S c i . Technol . , 19, 6 6 3 (1982)

8 ) Nakashima H., Ohta M. e t al. : " T r i t i u m Breeding C a p a b i l i t y o f Hel io t ron-H F u s i o n Reactor B l a n k e t s " , J. Nucl. S c i . Technol . , 2, 762 (1982)

9 ) do s., Nakai S. e t a l . : " N e u t r o n i c s C a l c u l a t i o n s i n P e l l e t s and B l a n k e t s o f L a s e r F u s i o n R e a c t o r Concept SENRI-I", J. Nucl. S c i . Technol . , 2, 1019 (1982)

1 0 ) Nakahara Y. and T s u t s u i T. :"NMTC/JAERI, A S i m u l a t i o n Code system f o r High Energy N u c l e a r React i o n s and Nucleon-Meson T r a n p o r t P r o c e s s e s " , JAERI-M 82-198 (1982) ( i n J a p a n e s e )

11) Nakahara Y. : " E v a l u a t i o n o f Computat i o n a l Models f o r F i s s i o n and S p a l l a t i o n R e a c t i o n s Used i n Accelerator Breeding and T r a n s m u t a t i o n A n a l y s i s Code", J. Nucl. S c i . Technol . , 20, 511 (1983)

1 2 ) Nakamura T., F u j i i M. et a l . : "Neutron P r o d u c t i o n from Thick T a r s e t s o f Cstrbon, Ira;;, Copper ahc Lead by 30- and 52-MeV i r o c o n s " , K i c l . SCA. Engng., 83. 444 (1983) -

S h i e l d i n g

Concerning t h e 3-dimensional t r a n s p o r t method, d i s c r e t e o r d i n a t e s codes PALLAS-XYZ and PALLAS-RTZ have r e c e n t l y been developed f o r d e a l i n g r e s p e c t i v e l y w i t h ( x , y , z ) and ( r . 0 , ~ ) geometr ies on t h e b a s i s of a d i r e c t i n t e g r a t i o n method t o t h e i n t e g r a l t r a n s p o r t e q u a t i o n , l ) and t h e PALLAS-XYZ h a s been a p p l i e d t o ana lyze an experiment of f a s t neu t ron s t reaming through a l a r g e vo id d u c t 2 ) and t o ana lyze d e t a i l e d neu t ron f l u x e s i n a PWR p r e s s u r e v e s s e l . 3 ) I n a d d i t i o n , a double f i n i t e element method where bo th t h e space and t h e ang le f i n i t e e lements a r e employed, was a p p l i e d t o s o l v e t h e mul t ig roup neu t ron t r a n s p o r t equa t ion by us ing t h e G a l e r k i n ' s weighted r e s i d u a l method and t h e v a r i a t i o n method.*)n5)

A s f o r s t reaming problems, a s e r i e s of measurements of 14 ?lev D-T neut rons s t reaming through a s l i t and a duc t i n c o n c r e t e s h i e l d s was c a r r i e d o u t u s ing a Cockcroft-Walton-type neutron gene ra to r and t h e r e s u l t s have been shown t o a r e e w e l l wi th t h e c a l c u l a t e d va lues w i t h t h e MORSE-CG code. 6 ) , 7 3 ~t t h e PXS f a c i l i t y , 2 t ypes of s t reaming exper iments were performed t o o b t a i n benchmark d a t a f o r v e r i f y i n g t h e d a t a and zethods .*) Furthermore, ano the r FNS benchnark experiment on D-T neu t ron and secondary gamma-ray s t reaming through a conc re t e ben t duc t was analyzed wi th t h e MORSE-GG code t o g i v e a good agreement w i t h each o t h e r . 9, 3 lo)

The a lbedo Monte C a r l o code MORSE-ALB was improved t o t r e a t t h e deep p e n e t r a t ion of r a d i a t ion i n s h i e l d i n g con•’ i g u r a t ion of l a r g e s c a l e geometry. For t h i s geometry, t h e Monte Carlo-Monte Car lo , o r d i s c r e t e ordinates-?:onte C a r l o coupled c a l c u l a t i o n i s advantageous, and t h e a r b i t r a r y coup l ing s u r f a c e s were made t o be a p g l i c a b l e t o forward and a d j o i n t c a l c u l a t i o n s . Besides , newly added t o t h e MORSE-ALB code were s e v e r a l f u n c t i o n s , such a s nex t event s u r f a c e c r o s s i n g f l u x e s t i m a t o r f o r c y l i n d r i c a l and s p h e r i c a l s h e l l geometry, importance sampling method based on t h e exponen t i a l t r a n s f o r n a t i o n , and volume sou rces f o r

a va r ious c o n f i g u r a t i o n . The improved code was a p p l i e d t o t h e ana lyses of t h e s h i e l d i n g experiment on t h e f a s t neutron source r e a c t o r YAY01 and t h e s t reaming measurement on t h e primary coo lan t p ipe of j O Y O . l l )

Concerning t h e s h i e l d i n g measurement of J O Y O , s e v e r a l measurements were performed t o o b t a i n t h e s h i e l d i n g c h a r a c t e r i s t i c s d a t a . Neutron f l u x d i s t r i b u t i o n s and energy spectrum were measured by us ing a c t i v a t i o n f o i l s i n sodium i n t h e r e a c t o r v e s s e l , i n g r a p h i t e s h i e l d and i n r e a c t o r p i t room. The measured d a t a w i l l be ana lyzed by two-dimensional d i s c r e t e o r d i n a t e s code and a lbedo Monte C a r l o code.

The FFTF/JOYO S h i e l d i n g Data Exchange Meeting was h e l d i n Tokyo on May 1 2 and 13, 1983, between USDOE and PNC. Following t h e meeting, an a n a l y s i s h a s been proceeded f o r t h e s h i e l d i n g

c h a r a c t e r i s t i c s of FFTF, and t h e c a l c u l a t i o n w i l l be compared wi th t h e measurement t o p r e d i c t t h e r e l i a b i l i t y of s h i e l d i n g c a l c u l a t i o n s .

A s concerns t h e s h i e l d i n g of s p e n t f u e l t r a n s p o r t ca sks , s h i e l d i n g exper iments were performed us ing PHR spen t f u e l assembl ies t o o b t a i n t h e benchmark d a t a f o r e v a l u a t i n g a code system of s h i e l d i n g s a f e t y a n a l y s i s . The f i r s t s t e p a n a l y s i s was c a r r i e d ou t f o r examining t h e p r e s e n t s t a t u s of t h e code system composed of ORIGEN2, ANISN and DOT-^.^.^^) I n a d d i t i o n , i n t e g r a l exper iments w i t h a Cf-252 source were performed f o r a cask a s des igned and t h a t l o s t i ts r e s i n s h i e l d . The measured neu t ron and secondary gamma-ray dose r a t e s were compared w i t h t h e Monte C a r l o c a l c u l a t i o n us ing t h e next-event s u r f a c e c r o s s i n g e s t i m a t o r . 1 3 ) , 1 4 )

1) Takeuchi K. and Kanai Y. :"Development of a S e r i e s of PALLAS Discre te -Ord ina tes D i r e c t - ~ n t e g r a t i o n Codes", S i x t h I n t e r n a t i o n a l Conf. Rad ia t ion S h i e l d i n g (ICRS), Paper 3b-1 (1983)

2 ) Sasamoto N . , Takeuchi K. e t a l . : "Analys i s of Neutron Streaming Through Void D u c t w i th Three-Dimensional Transpor t Code PALLAS-XYZ", S i x t h ICRS, Paper 6a-4 (1983)

3 ) Takeuchi K. and Sasamoto N. : "Analys i s o f De ta i l ed Neutron F luxes i n a PWR P r e s s u r e Vesse l by Two- and Three- Dimensional PALLAS Transpor t Codes", Nucl . Technology, e, 207 (1983)

4 ) Fujimura T . , Nakahara Y. e t a l . : "App l i ca t ion of Space- and-Angle F i n i t e Element Method t o t h e Three-Dimensional Neutron Transpor t Problems", S i x t h ICRS, Paper 3b-8 (1983)

5 ) F u j i n u r a T. , Nakahara Y. e t a l . : " S o l u t i o n of Three- Dimensional Neutron Transpor t Equat ion by Double F i n i t e Element Method", J. Nucl. S c i . Technol. , 20, 620 (1983)

6 ) Hashikura H. , Fukumoto H. e t a l . :"Neutron Streaming Through a S l i t and Duct i n Concre te S h i e l d s and Comparison w i t h a Monte C a r l o Ana lys i s " , Nucl. S c i . Engng., - 84, 337 (1983)

7 ) Hashikura H. , Oka Y. e t a l . : "Fas t Neutron Streaming S t u d i e s Using t h e F a s t Neutron Source Reactor , YAY01 and a 14 MeV Neutron Genera tor" , S i x t h ICRS, Paper 6b-4 (1983)

8 ) Nakamura T. , Ovama Y. e t a l . : "Rad ia t ion Streamina S t u d i e s a t t h e Fusion Geu t ron ic s Source (FNS) F a c i l i t y " , S i x t h ICRS (1983 )

9 ) Tanaka S . , Oyama Y. e t a l . : " A Benchmark Experiment on D-T Neutrons and Secondary Gamma Rays Streaming Through a Concrete Bent Duct", JAERI-M 82-130 (1982)

1 0 ) Sek i Y. , Tanaka S. e t a l . : "Monte C a r l o Analys i s of a Streaming Experiment of D-T Neutrons and Gamma Rays Through a Concrete Bent Duct", S i x t h ICRS, Paper 6b-6 (1983)

11) Kawai M . , Hayashida Y. e t a l . : "Appl ica t ion of Albedo Monte Carlo :4ethod t o FBR Neutron Streaming Ana lys i s " , S i x t h ICRS, Paper 6a-2 (1983)

12) Tanaka S. , Sakamoto Y. e t a l . : "Sh ie ld ing Experiments f o r a S h i e l d i n g S a f e t y Eva lua t ion Code System of Spent Fuel Transpor t Cask", S i x t h ICRS, Paper 7-6 (1983)

1 3 ) Ueki K. , Inoue M. e t a l . : " V a l i d i t y of t h e Monte Ca r lo Method f o r S h i e l d i n g Analys i s of a Spent-Fuel Shipping Cask : Comparison w i t h Experiment", Nucl. S c i . Engnq., 84 , 271

14) Ueki K. , Yamakoshi H. e t a l . : " I n v e s t i g a t i o n of t h e NESX E s t imat ion i n t h e Monte Car l o C a l c u l a t i o n s f o r S h i e l d i n g Analys i s of a Cask", S i x t h ICRS, Paper 3a-7 (1983)

15) Xiura T . and Sasamoto N. : "Exper imental Study of Neutron Streaming Through Steel-Walled Annular Ducts i n Reactor S h i e l d s " , Nucl. S c i . Engng., 83, 333 (1983)

Fas t Reactor Phys i c s

1. Experiments a t FCA According t o t h e m o d i f i c a t i o n of t h e F a s t Experimental

Reactor "JOYO" t o M K - 1 1 , a s e r i e s of mockup exper iments were c a r r i e d o u t on FCA Assembly X from A p r i l 1982 t o February 1983. The major m o d i f i c a t i o n s cons ide red on t h e mockup exper iments were: (1) i n c r e a s e of p lutonium c o n t e n t i n f u e l m a t e r i a l and ( 2 ) replacement of t h e uranium b l a n k e t by t h e s t a i n l e s s - s t e e l r e f l e c t o r .

FCA Assembly X c o n s i s t e d of t h r e e d i f f e r e n t v e r s i o n s of Assembly X-1, X-2 and X-3. The f i r s t two assembl ies were c l e a n cores w i t h no c o n t r o l r o d p o s i t i o n s and were c o n s t r u c t e d f o r a phys i c s mockup s tudy . To make c l e a r t h e e f f e c t s of t h e s t a i n l e s s - s t e e l r e f l e c t o r , s y s t e m a t i c exper iments were made i n t h e c l e a n c o r e w i t h t h e uranium b l a n k e t (Assembly X - 1 ) a s w e l l a s t h a t w i t h t h e s imu la t ed r e f l e c t o r (Assembly X-2). Measurements were made f o r c r i t i c a l i t y , f i s s i o n r a t e and sample worth d i s t r i b u t i o n s , and BqC rod worths .

The t h i r d assembly (Assea5ly X-3) was an engineer ing mockup c o r e of JOYO M K - I 1 which inc luded s i x sodium channe ls t o s i m u l a t e t h e c o n t r o l r o d p o s i 5 i o n s . I n Assembly X-3, mainly measured were s imu la t ed BaC rod wor ths and d i s t o r t i o n of neu t ron f l u x d i s t r i b u t i o n - d u e t o i n s e r t i o n of Na channels and/or BqC c o n t r o l rods .

Experimental s t u d i e s on fundamental p h y s i c s a s p e c t s of convent iona l l a r g e f a s t r e a c t o r c o r e s a r e i n p rog res s on FCA Assembly X I . The f i r s t v e r s i o n of t h e assembly (Assembly X I - 1 ) went c r i t i c a l a t t h e end of February 1983. The assembly has a c e n t r a l t e s t r eg ion of GOcm+ x 90cm h e i g h t s i m u l a t i n g t h e c o r e composit ion of a homo eneous f a s t r e a c t o r and a d r i v e r r eg ion mainly f u e l l e d wi th 295U . i-leasurements a r e being made f o r c r i t i c a l i t y , r e a c t i o n r a t e and sample worth d i s t r i b u t i o n s , and sodium-void and Doppler r e a c t i v i t y worths .

Var ious TLDs w i t h d i f f e r e n t e f f e c t i v e mass numbers a r e be ing i r r a d i a t e d i n Assembly X I - 1 t o expe r imen ta l ly determine t h e gamma-ra h e a t i n of t h e c o r e and b l a n k e t m a t e r i a l s i n f a s t r e a c t o r s . 2 5 ; 5 ~ and 238U f o i l s a r e be ing i r r a d i a t e d i n t h e assembly t o o b t a i n t h e d e t a i l e d in format ion on breed ing performances, t o g e t h e r w i t h on neu t ron f l u x d i s t r i b u t i o n near t h e core-b lanke t boundary.

2. Two-Dimensional Benchmark Problems Two-dimensional benchmark problems were s e t up f o r t h e t e n

f a s t c r i t i c a l a s sembl i e s w i t h c l e a n c o r e s , FCA-V-1, FCA-V-2, FCA-VI-1, FCA-VI-2, ZPR-6-7(Ref), ZPR-6-7(H240), ZPR-6-6A. ZPPR-2, ZPPR-9 and SEFOR. I n o r d e r t o perform t h e benchmark c a l c u l a t i o n s , t h e d a t a bank which c o n s i s t s of t h e d a t a f o r t h e geometr ies , t h e composi t ions and t h e exper iments and of t h e c o r r e c t i o n f a c t o r s were produced. Furthermore, two-dimensional benchmark c a l c u l a t i o n code system w a s genera ted . Using t h i s system, t h e benchmark t e s t s were performed f o r t h e JFS-3-J2 set.

3 . Group Constant S e t JFS-3-J2 The group c o n s t a n t s f o r 1 8 1 ~ a , l S 1 ~ u , 1 5 3 ~ u ,

237~qp, many f i s s i o n p roduc t s 2-4 t h e lumped n u c l i d e s of 2 3 5 ~ , 2 3 8 ~ an? 2 3 9 ~ ~ were a e n e r a t e d t o adv3c:e t h e -

. . ... J&-3,-J2 s e t f o r , , ; , . . . . . . , .. , 1. ! : 1:

a p p l i c a t i o n t o exper imenta l a n a l y s i s and des ign s tudy of l a r g e f a s t r e a c t o r s . The SLAROM code was improved s o a s t o be a b l e t o use t h e JFS-3-J2 s e t .

4. Development of C a l c u l a t i o n a l Method Concerning t h e method development, t h e f i n i t e Four ie r

t r ans fo rma t ion method was a p p l i e d t o s o l v e t h e mul t ig roup neutron d i f f u s i o n equa t ion f o r 2-dimensional t r i a n g l e geometry. Sample c a l c u l a t i o n s f o r a f a s t breeder r e a c t o r have shown t h a t t h e method g i v e s good r e s u l t s w i th fewer mesh p o i n t s than t h e u sua l f i n i t e d i f f e r e n c e meth0d. l )

A new mehtod was developed f o r f i r s t - o r d e r p e r t u r b a t ion theo ry , i n o rde r t o reduce t h e computing and l abo r c o s t of a n a l y s i s of exper iments on f a s t c r i t i c a l assmbl ies . 2 )

An e f f e c t i v e homogenization nethod of c o n t r o l r o d s , which p r e s e r v e s t h e i n t e g r a t e d r e a c t i o n r a t e s i n . a heterogeneous channel by i t e r a t i v e l y changing t h e c r o s s s e c t i o n s used i n a homogeneous s u p e r - c e l l c a l c u l a t i o n , has been extended t o t r e a t o f f - cen te r c o n t r o l rod channe ls i n FBR. An a lbedo a t t h e s u p e r - c e l l s u r f a c e was combined w i t h c o l l i s i o n p r o b a b i l i t i e s t o t r e a t t h e neutron leakage. The method h a s been a p p l i e d t o t h e c e n t r a l rod worth c a l c u l a t i o n in a t y p i c a l demonstra t ion LMFBR, and t o t h e 1 - D o f f - c e n t e r rod worth c a l c u l a t i o n . 3 )

To t r e a t t h e neu t ron d r i f t i n a asymmetric c e l l of f a s t c r i t i c a l assembly drawer, a formula h a s been de r ived f o r c a l c u l a t i n g t h e d r i f t c o e f f i c i e n t based on c o l l i s i o n p r o b a b i l i t y method. I t was u t i l i z e d t o c a l c u l a t e homogenized - d i f f u s i o n parameters f o r asymmetric p l a t e c e l l s of t h e ZPPR-9 co re . A f l u x d i s t r i b u t i o n was c a l c u l a t e d in a one-dimensional c o r e nodel w i th t h e d r i f t c o e f f i c i e n t , and was compared wi th t h a t ob ta ined from a t r a n s p o r t c a l c u l a t i o n t r e a t i n g p la te -wise he t e rogene i ty . When us ing t h e d r i f t c o e f f i c i e n t t h e f l u x dep res s ion i n lower energy was w e l l r e p r ~ d u c , ' i n t h e c o r e c e n t e r , which was caused by t h e p re sence of 2 3 8 ~ p l a t e .

5. R e a c t i v i t y Analys i s of Pin and P l a t e Cores The r e a c t i v i t i e s of p l a t e and p i n ZEBRA-CADENZA c o r e s were

analyzed a t Osaka Un ive r s i t y and J A E R I , and t h e r e s u l t s were presen ted a t t h e Nc4C-Xp s p e c i a l i s t meeting he ld on 21-23 June , 1983 a t Win f r i t h .

6. Analys i s of JUPITER Experiments The gamma-dose r a t e d i s t r i b u t i o n s i n ZPPR-9 and -10D

assembl ies , measured w i t h t h e TLD d e t e c t o r s , were analyzed us ing JENDL-2 neu t ron d a t a and ENDF/B-IV gamma-ray produc t ion da t a . The gamma-ray source d i s t r i b u t i o n s were c a l u l a t e d by 7 groups XY and RZ d i f f u s i o n model, and t h e gamma-ray d i s t r i b u t i o n s were c a l c u l a t e d by 20 groups XY and RZ SgP3 method. R e s u l t s a r e a s fo l lows .

( 1 ) A n a l y t i c a l r e s u l t s a r e c o n s i s t e n t between ZPPR-9 and ZPPR-1OD.

( 2 ) Dose r a t e s i n t h e c o r e r e g i o n a r e p r e d i c t e d w i t h i n an e r r o r of 10%. However. dose r a t e s i n C R P s t e n d t o be underpred ic ted .

( 3 ) Dose r a t e s i n t h e b l a n k e t r e g i o n a r e underpred ic ted by about 10%.

Analyses of r e a c t i o n r a t e d i s t r i b u t i o n s i n ZPPR-9 and -10 were r e f i n e d by c o n s i d e r i n g b o t h t h e shim rod and t h e ce l l asymmetry e f f e c t s f o r a l l t h e measurements. A s t h e r e s u l t s , t h e s p a t i a l dependence of C/E v a l u e was r a t h e r enhanced.4)

The r e f e r e n c e a n a l y s i s of t h e JUPITER-I1 exper iments , t h e r a d i a l l y heterogeneous l a r g e LMFBR c o r e c r i t i c a l exper iments , was s t a r t e d . C e l l models t o be a p p l i e d t o heterogeneous c o r e con•’ i g u r a t ions have been s t u d i e d .

7. E f f e c t of C e l l Model on Heterogeneous Core Parameter C a l c u l a t i o n

E f f e c t of c e l l model on p r e d i c t i n g t h e nuc l ea r c h a r a c t e r i s t i c s of a he te rogeneous c r i t i c a l assembly h a s been i n v e s t i g a t e d by ana lys ing t h e p h y s i c s exper iments made a t ZPPR-7A. The i n v e s t i g a t i o n h a s been made us ing d a t a from 3 k inds of c e l l models r e p r e s e n t i n g t h e c o r e and inner b l anke t r eg ions . The f i r s t c e l l model is s i m i l a r one used f o r homogeneous LMFBR a n a l y s i s . The o t h e r c e l l models a r e composed of c o r e drawers fol lowed by b l a n k e t d rawers , t a k i n g account of t h e i n t e r a c t i o n e f f e c t s between a d j a c e n t d i f f e r e n t k inds of subassemblies . The p r i n c i p a l e f f e c t s of c e l l model f o r ZPPR-7A a r e 0.43% Ak/kkl f o r c r i t i c a l i t y , 5% f o r 2 3 8 ~ ( n , f ) r e a c t i o n r a t e d i s t r i b u t i o n and 15% f o r Na-void r e a c t i v i t y e f f e c t .

8. Double Hete rogene i t E f f e c t of Fue l P in and Subassembly in a F a s t Power Reactor 533

The double h e t e r o g e n e i t y e f f e c t due t o t h e f u e l p i n and t h e subassembly is e s t ima ted f o r n e u t r o n i c s parameters of a p ro to type f a s t power r e a c t o r . Both of t h e h e t e r o g e n e i t y e f f e c t s caused by t h e f l u x f i n e s t r u c t u r e and t h e resonance s h i e l d i n g a r e t aken i n t o account . The model of t h e hexagonal u n i t subassembly c o n s i s t s of t h e smeared f u e l , t h e wrapper t u b e , and t h e o u t e r sodium r e g i o n s , where t h e average c r o s s s e c t i o n s of t h e smeared f u e l r e g i o n a r e ob ta ined by a u n i t p i n c e l l c a l c u l a t i o n i n c y l i n d r i c a l geometry. The h e t e r o g e n e i t y e f f e c t of t h e whole r e a c t o r model is c a l c u l a t e d based on two-dimensional d i f f u s i o n t h e o r y and p e r t u r b a t i o n theory . The double h e t e r o g e n e i t y e f f e c t is found t o be 0.5%Ak f o r k e f f , t h e p o s i t i v e sodiun-void worth is reduced by 26%, and t h e n e g a t i v e Dogpler r e a c t i v i t y i n c r e a s e s by 7 % f o r a p ro to type f a s t b reeder r e a c t o r . These r e s u l t s a r e cons ide rab ly l a r g e r t han t h e e s t i m a t e s made by e a r l i e r workers. . 9. Analys i s of Hete rogene i ty E f f e c t i n FCA-VI-2 by Monte Ca r lo

Code V I M A cont inuous energy Monte C a r l o Code VIM which h a s been

developed i n ANL has been conve r t ed i n t o t h e FACOM M200 computer. Th i s code can be used f o r v e r i f y i n g t h e accuracy of nuc l ea r c h a r a c t e r i s t i c s c a l c u l a t e d w i t h t h e code based on mul t igroup d e t e r m i n i s t i c t heo ry .

The a n a l y s i s of h e t e r o g e n e i t y e f f e c t f o r p i n and p l a t e c e l l i n FCA-VI-2 was performed by u s i n g t h e V I M code. The r e s u l t s c a l c u l a t e d w i t h V I M were i n a good agreement w i th t h o s e c a l c u l a t e d wi th t h e SRAC and/or SLAROM code based on t h e c o l l i s i o n p r o b a b i l i t y nicthod.

10.Burnup Analys i s of JOYO MK-I Core The burnup c h a r a c t e r i s t i c s of JOYO M K - I c o r e were

c a l c u l a t e d based on t h e a c t u a l o p e r a t i o n a l d a t a of t h e r e a c t o r , and were compared wi th t h e p o s t - i r r a d i a t i o n d a t a . I n c a s e of a low burnup of 20,000 MWD/T, t h e C/E v a l u e s a r e summarized a s fo l lows .

E f f e c t i v e m u l t i p l i c a t i o n f a c t o r : 0 -993 Cont ro l rod worth : 1.024 Burnup r e a c t i v i t y l o s s : 0 . 9 6 ~ 1 . 0 2 Burnup : 0 . 9 3 ~ 0 . 9 8

11. Conceptual Design Study of He-Cooled Ac t in ide Burning F a s t Reactor.

A conceptua l des ign s tudy of a He-cooled a c t i n i d e burn ing f a s t r e a c t o r was made a s a t r a n s m u t a t i o n system of waste a c t i n i d e s , i n a d d i t i o n t o t h e i n t e g r a l ex2er iments t o measure f i s s i o n r a t e r a t i o s and sample r e a c t i v i t y worths i n 8 assembl ies of FCA t o e v a l u a t e t h e n u c l e a r c r o s s s e c t i o n s of t h e major a c t i n i d e s .6 )

Some o the r r e p o r t s on f a s t r e a c t o r p h y s i c s 7 ) ~ 8 ) ~ 9 ) 1 1 0 ) , publ i shed in t h i s p e r i o d , a r e a l s o given i n t h e r e f e r e n c e .

Yokoo T . and Kobsyashi K. : " S o l u t i o n of t h e Mult igroup Di f fus ion Equation f o r Two-Dimens i o n a l T r i angu la r Regions by F i n i t e Four i e r T r a n s f o r a a t i o n " , Nucl. S c i . Engng., 83, 415 (1983) Azekura K . , Kawashima K. e t a l . :"Simple Method f o r Eva lua t ing Mesh S i z e E f f e c t s i n Con t ro l Rod Worth Ana lys i s " , J. Nucl. S c i . Technol. , 20, 518 (1983) Ono S. and Takeda T . :"A Homogeneiza?ion Method of Con t ro l Rods wi th Neutron Leakage E f f e c t " , NEACRP-A- (1983) Kato Y. e t a l . :"Radial Dependence of C / E Value i n J U P I T E R - I Core Ana lys i s " , NEACRP-A- (1983 ) Nakagawa M. an5 Inoue H. :"Double Hete ronene i ty E f f e c t of Fuel Pin and Ssbassembly i n a F a s t Power l ~ e a c t o r " , Nucl. S c i . Engng., 83, 214 (1983) blurata H. and Mukaiyama T. : " F i s s i o n Reactor S t u d i e s i n View of Reactor Waste Programme", Th i rd I n t e r n a t i o n a l Conf. Emerging Nuclear Energy Systems, H e l s i n k i (1983) Takeda T. and Wachi E. : " I n t e r f e r e n c e E f f e c t of Neutron Streaming Between D i f f e r e n t F a s t C r i t i c a l Assembly C e l l s " , Nucl. S c i . Engng., 81, 551 (1982) Kamei T. and Yoshida T. : "Er ro r due t o Nuclear Data U n c e r t a i n t i e s i n t h e P r e d i c t i o n of Large Liquid-Metal F a s t Breeder Reactor Core Performance Parameters" , Nucl. S c i . Engng., 84, 83 (1983) Akiyama M., Furu ta K. e t a l . :"Measurements of Gamma-Ray Decay Heat of F i s s i o n Produce ts f o r F a s t Neutron F i s s i o n s of U-235, Pu-239 and U-233", J . Atomic Energy Soc. Japan, 24, 709 (1982) ( i n J apanese ) z i y a m a M., Furu ta K. e t a l . :"Measurements of Beta-Ray Decay Heat of F i s s i c n Produc ts f o r F a s t Neutron F i s s i o n s of U-235, Pu-239 and U-233", J. Atomic Energy Soc. Japan , 2, 003 (1982) ( i n J apanese )

Nat i o n a l Programs

1. J O Y O A t t h e Experimental F a s t Reactor "JOYO", t h e MK-I c o r e h a s

been conver ted t o t h e M K - I 1 c o r e s i n c e January , 1982, and i ts i n i t i a l c r i t i c a l i t y was ach ieved on November 22, 1982. The M K - I 1 c o r e is t o be used a s an i r r a d i a t i o n bed f o r f u e l and m a t e r i a l .

The nuc lea r and hydrodynamic exper iments of t h e Mk-I1 c o r e were done a t low power. R e s u l t s of v a r i o u s k inds of exper iments were i n good agreement w i t h p r e d i c t e d va lues . The power ascens ion t e s t s were s t a r t e d on February, 1983, and t h e Mk-I1 c o r e reached i t s r a t e d power, 1001MWt, i n March.

A s e r i e s of c o r e performance c h a r a c t e r i s t i c s exper iments , i nc lud ing r e a c t i o n r a t e d i s t r i b u t i o n and r e a c t i v i t y worths due t o s u b s t i t u t i o n s of c o r e a s sembl i e s by i r r a d i a t i o n r i g s , were c a r r i e d o u t from A p r i l t o J u l y .

The f i r s t c y c l e of normal o p e r a t i o n , 45 days long, was s t a r t e d i n August, and s o f a r t h e r e a c t o r h a s been o p e r a t e d s a t i s f a c t o r y . The r e a c t o r w i l l be o p e r a t e d i n t h e way of four c y c l e s pe r year .

2. MONJU The s a f e t y e v a l u a t i o n work f o r "MONJU" by t h e Nuclear

S a f e t y Commission of Japan was s t a r t e d i n May, 1982 and completed i n May, 1983. P re -cons t ruc t ion works i n t h e s i t e a r e now being conducted and t h e excava t ion f o r t h e base mat w i l l be s t a r t e d i n June, 1985. The n e g o t i a t i o n between PNC and manufacturers f o r t h e c o n t r a c t of t h e p l a n t is in p r o g r e s s and w i l l be concluded soon. Achievenent of t h e i n i t i a l c r i t i c a l i t y of t h e r e a c t o r is expected on March,l991.

The Codes of P r a c t i c e on des ign and c o n s t r u c t i o n of LMFBR power p l a n t a r e now be ing s t u d i e d by t h e r e g u l a t o r y body f o r MONJU . 3. Denonstrat ion FBX

Design s t u d i e s have been con t inued f o r t h e Demonstration FBR P l a n t , 1000 MWe c l a s s subsequent r e a c t o r t o MONJU. The r e a c t o r is expected t o begin c o n s t r u c t i o n i n e a r l y 1990s, t o demonstra te i t s performance, r e l i a b i l i t y and s a f e t y a s a commercial-scale r e a c t o r and t o conf i rm t h e economic DrosDect - - f o r f u t u r e commerc ia l iza t ion . Design c r i t e r i a , systems and equipments have been reviewed mainly from t h e p o i n t of view of reduc ing c a p i t a l c o s t .

4. FUGEN The f o u r t h r e f u e l l i n g was c a r r i e d o u t d u r i n g t h e scheduled

shutdown i n June 1982. The t h i r d annua l i n s p e c t i o n and t h e f i f t h r e f u e l l i n g was c a r r i e d o u t from September 1982 t o January 1983 on schedule , and FUGEN had con t inued s t a b l e f u l l power o p e r a t i o n f o r about 7 months u n t i l t h e scheduled shutdown on August 1983 f o r t h e s i x t h r e f u e l l i n g .

12 of 40 f u e l a s sembl i e s d i s c h a r g e d i n t h i s r e f u e l l i n g were t h e i n i t i a l loaded ones and a l l t h e i n i t i a l loaded f u e l assembl ies had been d i scha rged from t h e co re .

120 MOX and 160 U02 f u e l ~ z s e m b l i e s , i nc lud ing fou r s p e c i a l f u e l a s sembl i e s have bee71 discharges :or t h e s i x r e f u e l l i n g s . The maximum burnup of MOX f u e l assembly is

, ,

13,600 MWD/ t , and no l eak ing f u e l has been found f o r 1 ,014 e f f e c t i v e f u l l power days of o p e r a t i o n up t o t h e end of August 1983.

5 . FUGEN-HWR Demonstration p l a n t S ince t h e c o n s t r u c t i o n program of t h e 6OOMWe FUGEN-type

HWR demonstra t ion p l a n t was f i x e d wi th t h e d e c i s i o n given by Japan AEC i n 1982, PSC had made i t s des ign s t u d i e s of t h e demonstrat ion p l a n t w i th coope ra t ion of EPDC ( E l e c t r i c Power Development Company), t h e pr imary under taker of t h e p l a n t . These s t u d i e s were completed in June 1963. The f u r t h e r des ign a c t i v i t i e s w i l l be followed by EPDC t o g e t t h e c o n s t r u c t i o n pe rmi t , and PNC i s a c t i v e l y c o o p e r a t i n g wi th EPDC mainly f o r R & D and b10X f u e l f a b r i c a t i o n of t h e p l a n t .

I t i s expected now t h a t t h e c o n s t r u c t i o n w i l l s t a r t in 1933 and cha t t h e commercial o p e r a t i o n w i l l s t a r t i n 1994.

6 . Experimental Very High Temperature Gas-Cooled Reactor (VHTR)

n; _..e c u r r e n t u s e o f ~ U C ~ ~ E I - e3erz f i s r e s t r i c t e d i n e l e c t r l c

, - e l e r s t i o n - v:hic'il c o n t r i b u t e s aro';-?d 30'" of t o t e 1 ur igar j r e-ergy su?;ly. T k r e Z o ~ e , a z p ? l i c a t i c r _ of n u c l e z r energy t o non- e l e c t r i c s e c t o r x i l l be a r a t t e r o f sy;ecla1 i z p o r t m c e .

i s conce=s 323 cf t h e e x r ~ e r i a e n t a l ?Zi:TZ, t 5 e t e s t l oo? f o r . '-..'3'L (--- 7- l a r g e conponents , id:\ L ..elium h n g i n e s r i n g Demonstrz t ion Loo?),

i s nvs i n o p e r a t i o n . In a d s i t i o n , t h e r e c m s t r u c t i o n of t h e c r i t i c a l e r p e r i n e n t a l f e c i l i t y , SHE, i s u n d e r v:aa t o s i ~ d l e t e t h e c o r e condi- t i o x of t h e e x > e r i n e n t a l VET3.

- 93 -

Ta3 le 1 S c e c i f i c a 3 i o n of t k e CanEickte Core

Design I t e m s 1 S o e c i f i c z t i o n s

Core dimension ( D i z v e t e r / X e i g h ~ ) No. or' f u e l c o l m a s No. or^ fuel b l o c k s p e r c o l m a So. of f u e l p i n s pe r f u e l block I n s i c e d i m e t e r of r e a c t o r v e s s e l C u z l e t c o o l z ~ t gas t e n v e r a t u r e

Cross Section of Fuel pin

Fig. 1 General view o f t h e rmdidate :cpe

Fig. 2 h e 1 block f o r

the candidate core

THE NETHERLANDS

Report on the Reactor Physics ~Ctivities

in the Netherlands

in the Period July 1982 - July 1983

compiled by M. Bustraan

1. Reactor Physics.at the Energy Research Centre Netherlands at Petten

* 1.1. Support to fast breeder reactor development

Fissionr~roduct-data A review of the status of fast capture cross-section evaluations for

important fission-product nuclides is given in Refs. ( 1,2[. This work has been used as a starting point for detailed recommendations of

evaluations for the Joint Evaluated data FILE (JEF-I). Recently, a

common report of ECN and ENEA (Bologna) was completed, with re-

commendations for the 60 most important fission products to be in-

serted in JEF-I, also including suggestions for re-evaluations, to

be made for JEF-2. Furthermore, a programme has been defined to per-

form integral-data tests of the JEF-I file, in cooperation with CEA,

Cadarache.

Work is in progress to redefine pseudo fissior ?roduct cross sections

based upon JEF-I. Recently, we have initiated a new evaluation of

1 2 9 ~ , in cooperation with R.E. Schenter, HEDL, Richland, U.S.A.

Activation of the sodium coolant ................................ Recently, the project to evaluate activation cross sections of cor-

rosion products, cover-gas nuclides and other nuclides in the primary

cooling circuit of a fast power reactor has been completed 131.

Revisions have been made for 2 2 ~ a and 64~n, based upon recent meas-

urements in the resolved resonance range. We have also conpiled 26-

group constants for natural Ar, that is used as a cover-gas of the

coolant. Further updatings are planned for the Ni-isotopes in coopera-

tion with CEA, Cadarache. The work will be directed to satisfy the

needs for re-evaluations of the JEF data file.

DeBeNe cooperation on fast reactor development.

1.2. Advanced water-cooled reactors

Upon request from KfK, Karlsruhe we have calculated group constants for

fission products in the 69-group WIMS structure for application in design

calcularions of advanced water cooled reactors, based upon the ENDFfB-V

data file. An appropriate microflux weighting spectrum was used.

1.3. Nuclear data for fusion reactors

As a contribution to the European Fusion Technology Programme (blanket

technology) we have developed nuclear-model codes for the prediction

of angle-energy correlated distributions of neutron emission cross

sections 14-81. In the existing evaluations these correlations are

0 comple,tely neglected, at least for th? continuum emission. Work i> . v

in progress to improve the ENDF/B-IV evaluation of lead, that will be

used in blanket configurations of future power reactors.

Further work on fusion reactor nuclear data is reported in semi-annual

progress reports 19 1 . References

1 1 I H. Gruppelaar, Status of recent fast capture cross section eva-

luations for important fission product isotopes, Proc. NEANDC/

NEACRP Specialist's Mtg. on Fast-neutron capture cross Sections,

Argonne 1982, NEANDC(US)-2 l4/L (1983) p. 473.

(21 H. Gruppelaar et al., Report of the working group on fast-neutron

capture cross sections for the most important fission-product

nuclei, ibid, p. 570.

131 H. Gruppelaar and H.A.J. van der Kamp, Evaluation of activation

cross sections of corrosion products, cover-gas nuclides and

other nuclides in the primary cooling circuit of a fast power

reactor, Proc. Int. Conf. on Nuclear Data for Science and Tech-

nology, Antwerp, 1982, p. 643, Reidel Publ. Co., Dordrecht (1983).

141 J.M. Akkermans, A. random walk in the land of precompound decay,

ECN-121 (1982).

151 H. Gruppelaar,. C. Costa, D. Nierop and J.M. Akkermans, Calcula-

tion and processing of continuum particle-emission spectra and

angular distributions, Proc. Int. Conf. on Nuclear Data for

Science and Technology, Antwerp, 1982, p. 537, Reidel Publ. Co.,

Dordrecht (1983).

161 C. ~&ta, H. Gruppelaar, and J.M. Akkermans, Energy dependence

of preeqqilibrium angular distributions, Lett. a1 Nuovo Cim. 2

171 C. Costa, H. Gruppelaar and J.M. Akkermans, Angle-energy corr-

lated, model of preequilibrium angular distributions, Phys. Rev.

.C28 - (1983), p. 587.

(81 J.M. Akkermans, Random-walk model of precompound decay 11: Stochas-

tic uncertainties in the lifetimes and cross-sections.

Z. Physik - 313 (1983), p. 83. ?

' I 19 1 ' J. D. Elen (comp. ) , Fusion Technology Programme Semi-Annual. Report '

January-June 1982, ECN-124 (1982), and ibid., July-December 1982,

ECN- 132 (1 983.).

1.4. Reactor Neutron Metrology

The final report on this project has been published in February 1983

1101. The aim of the interlaboratory REAL-80 exercise, organized by the M A ,

was to determine the state-of-the-art in 1982 of the capabilities of

'laboratories to adjust neutron spectrum information on the basis of a

set of experimental activation rates, and to subsequently predict the

number of displacements in steel, together with its uncertainty.

The input information distributed on magnetic tapes to participating

laboratories comprised values, variances and covariances for a set of

input fluence rates, for a set of activation and damage cross-section

data, and for a set of experimentally measured reaction rates. The

exercise dealt with two clearly different spectra: the thermal ORR

spectrum with 19 reaction rates, and the fast YAYOI spectrum with I ?

reaction rates. Cross-section data were supplied both in a 620 groups

structure and in a 100 groups structure. From 30 laboratories which

were asked to participate, 13 laboratories contributed 33 solutions

for ORR, and 35 solutions for YAYOI.

The spectral shapes of the solution spectra showed considerable spread,

both forthe ORR and the YAYOI spectrum. When the series of predicted

activation,rates in nickel and the predicted dispIacement rates in

steel derived for all solutions is considered, one cannot observe sig-

nificant differences due to the adjustment algorithm used. The largest

deviations seem to be due to effects related to group structure and/or

changes ih the input data.

When comparing the predicted activation rate in nickel with its avail-

able measured value, we observe that the predicted value (averaged over

all solutions) is lower than the measured value: 1 per cent lower for

ORR and 7 per cent lower for YAYOI.

For the predicted displacement rate in steel we observed a coefficient

of variation of 2,2 per cent for the ORR spectrum and 2,8 per cent for

the YAYOI spectrum, if all of the participant responses are considered.

1101 W.L. Zijp, E.M. Zsolnay, H.J. Nolthenius, E.J. Szondi, G.C.H.M.

Verhaag, D.E. Cullen, C. Ertek, "Final report on the REAL-80

exercise", Report ECN-128 (Also INDL(NED)-7; BME-TR-RES-6/82).

(Netherlands Energy Research Foundation ECN, Petten, Febr. 1983).

0 1.5. Experience with the On-Line Power Reactor Noise Monitoring System

for the Borssele (PWR) Power Reactor

For nearly two years the Borssele nuclear power plant (PWR, 450 MWe)

is connected to the on-line noise monitoring system at Petten 1 1 1 1 by

a special telephone link (200 km). By this system it was possible to

follow the 9th fuel cycle (full power operation as well as shutdown)

and the start-up of the 10th fuel cycle.

Noise characteristics of several primaire system sensors such as

reactor safety channels, core exit thermocouples, pressure trans-

ducers and several signals of the secondary system are continuously

0 monitored and reported. Signals from the in-core neutron detectors

and core exit thermocouple are used to determine core physics para-

meters by the noise analysis, such as core coolant flow velocity, and

reactivity effects. Local vibration of the in-core instrumentation guide

tube and thermocouple response characteristics ("in-situ test") were

determined on-line. Using the signals from the ex-core neutron de-

tectors and with aid of a spectrum decomposition, the amplitudes of

the core support barrel motions and the direction of the motion are

continuously monitored.

I t In the coarse of the year, theT'partial and multiple coherence analysis

and "univariate (UAR) and multivariate (MAR) autoregression" tech-

niques were applisd to the on-line measured reactor signals. Thus

monitoring of physical core parameters and sensor characteristics

by the current on-line noise analysis 1121, will enhance the relia-

bility of the operation of the power plant. , . . . .

I : . : : . :

The final object of this work is the development of on-line methods

which permit monitoring of the mechanical status of the main components

in the primary system and of the complete neutronic and thermodynamic

behaviour of the system. It is intented to include and integrate into

this system other noise monitoring systems such as on flow acoustics

and on loose parts monitoring in the near future.

1 1 1 1 E. Tiirkcan, Review of Borssele PWR Noise Experiments Analysis and

Instrumentation. Prog. Nucl. Energy - 9 (1982) 437. 112 1 E. Tiirkcan, R. Oguma, Improved Noise Analysis Methc~ds for On-line

Determination of In-Core Instrumentations and Power Reactor

Parameters. Paper to NEA-Couunittee on Reactor Physi.cs Specialists

Meeting on In-Core Instrumentation, OECD Halden Reactor Project,

10-13 October, 1983.

2. Reactor Physics at the Interuniversity Reactor Institute (Delft)

The first part of a reactor noise analysis programme for the Dodewaard

BWR has been concluded. Results, mainly pertaining to two-phase flow

velocity measurements and the determination of the at-power reactivity

transfer function, were published in a D.Sc. theses 1 1 I. A report has

been issued on analysis of signals from fixed in-core SPND's in the

same BWR 121. A first series of in-core measurements with a Rrin Self

Powered Gamma Detector shows promising results. The signals from the

TSPGD exhibit a phase relationship characteristic for a combination

0 of a global effect and a local effect with transport time. In the near

future a direct comparison with a neutron detector is expected to give

further information on the usefulness of gamma noise measurements.

A general paper on the interpretation of incore noise measurements

in BWR's was published 131.

The 2 MW pool type research reactor HOR has been a subject for thermal

hydraulic analysis, analyzing both stationary and noise signals of

thermocouples in an instrumented fuel assembly, also in connection

with the possible future application of Low Enriched Uranium (LEU)

in this reactor 141.

An out-of-core boiling loop has been completed which will be used in

support of a measurement programa on an in-core loop in HOR. This

0 programme aims at the study of neutron, gamma and temperature noise

signals in an electrically heated dummy fuel element with two-phase

flow.

A theoretical study on a very high temperature gaseous core reactor

was completed [5,61 and an improved method for handling continuum in-

elastic neutron scattering has been published 171.

References

( 1 ( Erik Kleiss, On the determination of Boiling Water Reactor Charac- teristics by noise analysis. D.Sc. thesis, Delft, University of

Technology, Delft University Press, 1983.

121 E.B.J. Kleiss, Investigations of the signals of Co-SPND' in the

Dodewaard reactor. Report IRI-131-82-03.

131 H. van Dam, Interpretation of In-core Noise Measurements in BWR's.

Kernenergie 26 (1983) 59-63. - , ' _

141 H. van Dam and J . W . de Vries, Research activities relevant to the

application of LEU at the HOR, Report IRI-130/131-€2-01.

( 5 1 R.M. Uleman, The influence of the core temperature distribution

on the reactivity of a gas core reactor.

M.Sc. thesis, Delft, 1982.

161 H. van Dam and J.E. Hoogenboom, Physics of a gaseous core reactor.

To be publ. in Nuclear Technology.

171 J .E. Hoogenboom, Consequences of inelastic discrete-level neutron-

collision mechanics for inelastic continuum scattering.

Ann. Nucl. En. 10 (1983) 19-29. -

3. Reactor Physics at KEMA (Arnhem)

Gadolinium ---------- 17 fuel elements, each containing 5 pins with gadolinium (out of 35

pins), have been loaded in cycle 14 of the Dodewaard reactor. The re-

activity behaviour of the core indicates that the burnout rate of the

gadolinium is underpredicted by about 10%. This seems to be consistent

with gamma scanning results of gadolinium fuel elements that have been

measured pin by pin after one cycle of burnup.

Optimalization - ------------

Our optimalization program has been expanded to three dimensional

quadrant geometry. This became necessary as the reactor can be con-

trolled by only 2 control rods, that cannot be represented in octant

geometry. Running times of 8 hours on a Univac 1160H are typical to

obtain the solutions. Simple calculations show that our optimization

goal of maximum cycle length does not necessarily lead to maximum

discharge burnup. The program is now being adapted to pressurized water

reactors.

FLARE type kernels ------- --------- The kernels as used in FLARE cannot easily be interpreted in terms of

collision probabilities. At present we are studying alternative formu-

lations, which maintain in one group theory the migration area for

point sources.

3D transient calculations ......................... For a PWR we have performed a 3D deboration transient study. A front

with lower borium concentrations passed through the reactor in the shut-

down conditions. Local supercriticality was achieved and severe damage

would be the result under the assumed conditions.

Noise analysis ---------- --- As a fruit of the noise work of IRI (see par. 2) noise methods are

applied in the Dodewaard reactor to in-vessel and in-core velocity meas-

urements using thermocouple and neutron signals. This enables a better

characterization of the thermohydraulic condition under which the re-

actor operates. Attention has been paid to the measurement of thermo-

hydraulic stability for which also calculations are under w y

9 ~ 1 0 0 1 0 4

NORWAY

STATUS REPORT TO NEACRP

(1982 - 1983)

Reactor physics activities in Norway,

September 1982 - August 1983

Compiled by:

T. ~karhamar Institute for Energy Technology Box 40 N-2007 Kjeller, Norway

S. BQrresen Scandpower A/S Box 3 N-2007 Kjeller, Norway

K. Haugset OECD Halden Reactor Project Box 173 N-1751 Halden, Norway

1. FMS CODE SYSTEM

Activities in reactor physics is concentrated on applications

and further development of FMS, the modular code system for

light water reactor calculations developed originally at the

Institute for Energy Technology (the previous Xnstitute for

Atomic Energy). Present and future developments, and maintenance,

of the system is now the responsibility of the international

consulting company Scandpower A/S, Kjeller. The main users

of FMS for fuel management work and plant operation support

are power utilities and other organisations in Europe, USA

and Japan.

1.1 Status of RECORD

Conversion of the production version of RECORD to an IBM

computer has been completed, thus both CDC and IBM code

versions are available.

A major gamma physics project has been carried out with the

purpose of including a model for gamma sensitive in-core

detectors in RECORD. A generic data base has been produced

using the AMPX-DOT code package. The model may be applied

both for calculating detector response parameters and gamma

heating in various regions within the fuel asselrbly.

Extension of the fission product modeLinclusion of Hafnium

as a possible control rod materia1,and development of a

model for hybrid BWR control blades are under way.

The main features of the RECORD code for application to BWR

and PWR fuel assembly calculations, and a review of the

code qualification, are described-in a published Kjeller

Report (ref. /I/) .

1.2 Status of PRESTO

BWR Code Version ---------------- A comprehensive qualific ect of PRESTO-B has been

carried out by Carolina Power and Light Company of Raleigh

N.C. (USA) and documented in a Topical Report to the US NRC.

Comparisons with reactor operating data and gamma scan data

from Brunswick 1, Brunswick 2 and Quad Cities 1 plants were

made. Nuclear data for PRESTO were generated with RECORD.

A schenlatic of the RECORD/PRESTO code system is shown in

Fig. 1.2.1.

The cumulative average eigenvalue for the six operating

cycles analysed was

The average standard deviation between measured and calcu-

lated local (24 nodes) TIP ratings for 3 cycles of Brunswick 1

was - 8.2%. The corresponding standard deviation for axially

integrated TIP'S was - 4.5%.

The measured versus calculated La-140 assembly peak-to-average 0

average difference for the Quad Cities 1, 1976 gamma scan

was -2.26 ? 1.55%.

PWR Code Version ---------------- The PWR version of PRESTO has now obtained production status

and has been qualified and prepared for PWR operations

support.

Three operating cycles of H.B. Robinson were analysed in

detail with PRESTO-P, using RECORD nuclear data.

Comparisons were made with measured radial and axial power

distributions, boron letdown cycles, BOC bomn endpoint

data, hot, zero power critical configurations and isothermal

core temperature coefficients. Good results were obtained

throughout. An example of power distribution comparison is

shown in Figs. 1.2.2 and 1.2.3.

2. REACTOR PHYSICS ACTIVITIES AT THE OECD HALDEN

REACTOR PROJECT

A prototype of the core surveillance system SCORPIO has

been developed within the Halden Project, and is today

operating at the Halden computer laboratory. One module

in this system, the core simulator, has been validated

against power reactor (PWR) data from power/xenon transients.

Another module, the strategy generator, is capable of

9 predicting control rod positions, coolant temperature and

boron concehtration as input to the simulator, such that

the resulting axial power offset during transients is

kept close to zero. A more advanced strategy generator,

based on optimal power distribution control, will be imple-

mented during 1984.

The Halden Project has during 1983 established an experi-

mental control room including a full scope training simulator.

This laboratory will partly be used for testing and demon-

stration of computerized operator support systems.

REFERENCES

/1/ T. Skardhamar, H.K. Ncss:

"Methods of RECORD, an LhT3 Fuel Assembly Burnup Code",

Kjeller Report IFE/KR/E-11 V (1982).

- 106 - FIGURE 1.2.1

MAJOR PUOGRAM MODULES USED FUR

CP&L STEADY STATE BWR ANALYSIS

THERMOS: Trampart theof-, code for burnable absorber . cross-section generation

GADPOL: Intwfacs code which provide r s a d b n rate matchbw batwaen transwrt and dittusion

FIECORD: Two-dimsntional multi-group , cod. for gen- eratlan of LWF! latticm phys~cs mscams

@ POLRAM: lntrntace code tor tlon f, palynomial cross-section tits 2%% &!a b r use PRESTO-t3

ALBMO: A fogram mrduie within PRESTX lor gen- era%on of corla leakage parameters

4 PRESTO -8 PRESTO-B: ihme-d imtfo~ra l uxlplod m u t r ~ l r a d hy-

draulk BWR rlodol rimulotlon cod.

MD-2 MO-1

I

FIBWR: EWR Th.rmo1-hydmuik Ods for generation of retomme prsssun and flow distrlbutfons

MD- 2: Two-dlnrmsionel, mW-gmu rim tneh diftusion thaort1 cod. tor &ion ot refer- . once flux distributiuu

IUD-9: Ons-dirnensiatxil analog ot MO-2

- 107 -

FIGURE 1.2.2

H.B. ROBINSON UNIT 2 CYCLE 6 AXIAL POUER DISTRIBUTION COMPARISONS

305 MUD/MTU 100% HFP

AXIAL POSITION NODES

H.B. ROBINSON UNIT 2 CYCLE 6 AXIAL POUER DISTRIBUTION COHPARISONS

9923 nuD/nTu leer HFP

a 0-INCORE *=PRESTO

AXIAL POSITION NODES

FIGURE 1.2.3

H.B. ROBINSON UNIT 2 CYCLE '7 AXIAL POUER D I S T R I B U T I O N U l t f P A R I S O N S

BOC HZP ARO D=INCOUE W R E S T 0

AXIAL P O S I T I O N NODES

H.B. ROBINSON UNIT 2 CYCIE 7 AXIAL POUER D I S T R I B U T I O N COHPAfUSONS

884 flUD/HTU 100% H F P o=INCORE *=PRESTO

AXIAL P O S I T I O N NODES

- 109 -

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

S P A I N

1 . - PWR CORE DESIGN AND FUEL MANAGEMENT (DENIM, Department of Nuclear Energy

F.l'SII, Universidad Po l i t ecn ica de nadr id ) .

A f i r s t phase of development and va l ida t ion of PWR core design and f u e l management was completed i n 1982, through a cooperat ive e f f o r t of an Spanish u t i l i t y (IBERDUERO), the nuclear research cen te r ( J E N ) and t h e DENIM. Main r e s u l t s were presented a t the September 82 ANS Topical Meeting on "Advances i n Reactor Physics and Core Thermal Hydraulics" (Ref. 1 )

Under a research con t rac t of the In te rna t iona l Atomic Energy Agency (CRP on codes f o r in-core f u e l management) t h i s computer code system has been documented, and is being made ava i l ab le through t h e NEA Data Bank.

The CARMEN system (Ref. 2 ) f o r f i n i t e d i f f e rence d i f fus ion ca lcu la - t i o n of PWR cores with space dependent nuclear and thermal hydraul ic feed- backs, which was opera t ing i n t h e UNIVAC-1100 a t JEN, was converted t o FORTRAN-77 and implemented i n the CYBER-835 a t DENIM. Both vers ions have been d i s t r i b u t e d t o t h e NEA Data Bank (Newsletter No. 3 of August 1983).

The MARIA system use r s nianual has been published (Ref. 3). and the implementation on t h e CYBER-835 a t DENIM was done with t h e co l l abora t ion of Dra. Teresa Kulikowska from the Badan Jadrowych I n s t i t u t e (Swierk, Poland), during her s t a y a t DENIM. This system includes an extended ve r s ion of WIMS- D-4 and a u x i l i a r codes (PREWIM and POSWIM) f o r prepara t ion and postprocessing of t h e WIMS c a l c u l a t i o n s f o r PWR f u e l assemblies. Every kind of PWR f u e l assembly, inc luding those with burnable absorber c l u s t e r s and c o n t r o l rods, can be modelled by t h e MARIA system using a bundle desc r ip t ion . The extensions i n WIMS-D.4 include t h e homogeneization and group colLapsinq t o fewgroups of the macroscopic and microscopic c r o s s sec t ions of t h e d i f f e r e n t f u e l o r absorber c e l l s i n t h e bundle f o r use i n the CARMEN system o r o t h e r core models. The MARIA system i s now a v a i l a b l e , i n both UNIVAC and CYBER vers ions , from t h e NEA Data Bank. The CYBER vers ion was converted t o FORTRAN-77 (FTN 5 . 1 ) . which requi red a b i g e f f o r t , s p e c i a l l y because of t h e non-standard procedures f o r c a l l s and e n t r i e s i n the o r i g i n a l and the UNIVAC vers ion . Most of the WIMS op t ions were t e s t e d b u t a few could s t i l l requ i re some checking by f u t u r e users .

The CICLON code ( r e f . 4 ) f o r f a s t survey f u e l management c a l c u l a t i o n s was a l s o implemented on t h e CYBER-835, with t h e co l l abora t ion of D r . R.P. J a i n from Bhabba Atomic Research Center during h i s s t a y a t DENIM i n Apr i l 83. This vers ion has been made ava i l ab le too t o t h e NEA Data Bank.

Work is proceeding f o r documentation and f u r t h e r v a l i d a t i o n of our 1 group 3D PWR s imula tor (LOLA system), which inc ludes new t r a n s p o r t kernels and cor rec t ion f a c t o r s f o r t h e nodal K- and M'. Those t r a n s p o r t and s p e c t r a l f a c t o r s , a s wel l a s t h e co re - re f l ec to r albedoes i n one-group,are e x p l i c i t l y ca lcu la ted by an a u x i l i a r y code from fine-mesh two-group d i f f u s i o n f l u x e s obtained from the CARMEN o r the VENTURE code. Those f a c t o r s and albedoes a r e such defined t h a t , f o r q u a r t e r of assembly nodes, they do not change among the d i f f e r e n t nodes of t h e same kind (enrichment and number of burnable poison rods) a s well a s with the condi t ions along opera t ion (dens i ty , Boron and burnup). Rather good accuracies have been achieved with these procedures a s discussed i n r e f s . 1 and 5. A more d e t a i l e d paper w l l l be submitted f o r - - publication i n Nuclear Technology ln a shor t term.

of Other l i n e development is under way f o r the loca l ca l cu la t ion of pin powers in f u l l f u e l assemblies o r i n 4 qua r t e r s o f neighbour assemblies under given boundary f luxes o r current - to- f lux r a t i o s . These l o c a l c a l c u l a t i o n s w i l l provide a l s o the homogeneized c ross sec t ions and co r rec t ion f a c t o r s f o r the d i f f u s i o n terms t o be used i n two-group coarse-mesh f u l l core c a l c u l a t i o n s Af ter demonstration of the accuracy and convergence of the loca l c a l c u l a t i o n s f o r PWR assemblies and color s e t s ( 4 q u a r t e r s o f neighbour assembl ies) , t he loca l -g lobal code i s being developed f o r i t e r a t i o n on the i n t e r f a c e f luxes and c u r r e n t s i n 2D. The f i n a l phase w i l l be t o incorporate these modules i n a 3D 2-group coarse-mesh simulator f o r PWR ope ra t ing and t . rans ient a n a l y s i s , including t h e CARMEN modules f o r c ross sec t ion feedback.

REFERENCES

1.- ARAGONES J . M . , AHNERT C. , GOMEZ-SANTAMARIA J . , and OLAVARRIA I . R . , "Development and Validat ion of Core Physics Methods f o r In-Core Fuel Management of PWR's", NUREG-CP-0034, pp. 315-332 (1982).

2.- AHNERT C., and ARAGONES J . M . , "CARMEN System, A Code 13lock f o r Neutronic PWR Calcula t ion by Diffusion Theory with Space-Dependent Feedback Ef fec t s " , JEN-515 (1982) .

3.- ARAGONES J . M . and AHNERT C . , "MARIA System, A Code Block f o r PWR Fuel Assembly Ca lcu la t ions" , JEN-543 (1983)

4 . - ARAGONES J . M . , "CICLON, A Neutronic Fuel Management Program f o r PWR Consecutive Cycles", JEN-336 (1977).

5.- ARAGONES J . M . , AHNERT C. GOMEZ-SANTAMRRIA J . and OLAVRRRIA I . R . , "Desarrol lo y ValidaciBn de M6todos y Programas de Cdlculo d e l NQcleo de Reactores de Aqua a PresiBn", Nuclear Espaiia, No. 5. pp. 18-34 (1982).

2.- NEUTRONICS ANDSHIELDING OFFUSION BLANKETS (DENIM)

Two mayor e f f o r t s have been done a t DENIM i n t h e development of methods and i n i t i a l c a l c u l a t i o n s on the neut ronics of b l anke t i n fus ion r eac to r s .

The f i r s t one i s l inked with the a n a l y s i s of t h e neut ronic f luxes f o r d i f f e r e n t c l a s s e s of the f i r s t wall and b lanket , t o s t a b l i s h f i n a l l y the temporal response of the system t o pe r iod ic r e p e t i t i o n r a t e s i n i n e r t i a l confinement fus ion . S o l i d , gases and wetted wal ls have been i n i t i a l l y evalua t - ed in a very s i m p l i f i e d way i n order t o c o n t r a s t t he t y p i c a l conclusions about i ts performance and cons t ruc t ion magnitudes. Other goal of t h i s work i s t o be ab le t o genera te an a n a l y t i c a l model, t e s t e d with more soph i s t i ca t ed c a l c u l a t i o n s , which could desc r ibe the thermal c h a r a c t e r i s t i c s of the cooling system in the b lanket .

The calculations begun with a reference design of blanket from thc bibliography, but getting the magnitudeand spectrum of the neutron source from precedent calculations of DENIM on target analysis.

'Pdo different libraries have been used (EURLIB-IV, MACKLIB-IV) to obtain the a's, kerma factors for the material in the first wall and the blanket. The ANISN code is used to perform a shorter multigroup library (18 groups) to be used in the dynamic calculations. Finally, the CLARA code is computing the neutron flux and heating of the zones in the blanket in a temporal frame.

The other project is connected with the neutron damage in fusion reactors. Here, the line of work is some different of the former. The basic library ENDF/B-IV is used as data source for the NJOY code which generates the a's for structural materials, gases (H, He) forming in the blanket, tritium reproduction, etc, Kermas and energy available (E ) to produce atomic

a displacements. The pairs of atomic displacement is calculated following the expression Ea/2Ed, where E is the energy needed to remove an atom in the

d. lattice obtained from experiments. Different materials have been just gene- rated and added to the description of the systems in CLARA code.

Two different designs are being tested. One of them is a filled gas cavity (SOLASE ) . Neutronic fluxes have been calculated for it, but a major effort is being carried out to evaluate the INPORT design developped by the University of Wisconsin for the HIBALL project (Heavy Ions). This system supposes a wetted wall by Li Pb permitting a higher repetition rate. A large multigroup (18 groups) library, S6-P3), has been generateu including materials as Li, Pb, Si, D, T, He, Fe, Cr, incorporated to the INPORT tubes and cooling reflector, first wall and blanket. The Onedimensional and temporal calculat- ions using the CLARA code are being just performed.

Other libraries including Va, Mo, Nb, and its alloys are being generated to in order to have the capability to analyse the characteristics of these refractory materials recommended for the first wall in fusion reactors.

3.- NUCLEAR DATA PROCESSING FOR FISSION AND FUSION REACTORS (DENIM)

In order to evaluate the nuclear data, our Department chose the NJOY code. The actual version, tested and optimized for our requirements, is denominated NJOY 10/81-3, updated in September 1982.

NJOY is the successor to the MINX code. It processes ENDF/B-4 and 5 files with the exception of MF8 (decay data) and MF32 (resonance covariances) Because of its great versatility, it is very efficient for using in thermal reactors, fast reactors, fusion reactors and shielding analysis.

NJOY produces neutron cross sections and group to group scattering ,matrices, heat and damages cross sections, activation cross sections, photon

production matr ices , photon in t e rac t ion c ross sec t ions an3 group t o group matr ices , delayed neutron spec t ra , thermal s c a t t e r i n g Cross sec t ions and matr ices , and c r o s s sec t ions covar iances l .

I t is a modular system where each module is an e s s e n t i a l l y f r ee - s tanding code devoted t o one p a r t i c u l a r processing t a sk . The communication between two modules is achieved v i a d isk f i l e s in ENDF/B format. These f i l e s - a r e the o r i g i n a l ENDF/B tape , a pointwise ENDF tape (PENUF), and a groupwise ENDF tape (GENDF; .

The fol lowing t a b l e i s a b r i e f desc r ip t ion of the modules funct ions :

MODER RECONR BROADR UNRESR HEATR HERMR GROUPR GAMINR ERRORR DTFR CCCCR MATXSR

mode conversion (BCD/ i n t e r n a l blocked binary) xs pointwise recons t ruc t ion Doppler broadening unresolved XS heat ing , damage and a c t i v a t i o n pointwise da ta . thermal pointwise d a t a . neutron and ngamma groupwise da ta gamma i n t e r a c t i o n xs xs covariances DTF format output CCCC format output MATXS format output

The modules sequence optimized by D E N I M a r e now b r i e f l y d e t a i l e d . The f i r s t provides t h e nuclear d a t a f o r neutronic r e a c t i o n s , and t h e second f o r the gamma ones.

l i b r a r y

j o g l i b r a r y

Note.- In the f i r s t sequence, the UNRESR module is only employed i f t h e s p e c i f i c ma te r i a l has unresolved resonance paranieters.

A p o s t e r i o r program (ANGIE) allows f o r the formation of the nuclear da ta matr ix f o r the coupled t r a n s p o r t of both spec ies : neutrons and photons.

By means oE t h i s code, the nuclear da ta requested f o r our ICF systems t r a n s p o r t c a l c u l a t i o n s (NORCLA code1 have been generated.

The energetic structure is 18 groups for neutrons and 12 groups for photons. The weight function for the multigroup treatment is the self flux spectrum resulting from a first interaction with a weight function (1/E + +thermal mawellian + fission spectrum).

The nuclear data obtained are essentially of two types: nuclear data for ICF target materials and nuclear data for ICF reactor structural materials (first wall, blanket,reflector and shield).

In the first group are Deuterium, Tritium, Aluminium, Gold and Lead. These two last materials, Au and Pb, required a special treatment of the Kerma factor evaluations due to a lack of consistency in the ENDF/B-4 files, revealing some negative Kerma factors in several points of the energy grid.

In the gold case, the negative Kerma factors were caused by not consistent evaluations of the reactions MT16 (n,2n) and MT17 (n.311). In order to solve this problem, the corresponding Kerma factors for these reactions were analyticaly evaluatedl. Once corrected the interface tape at HEATR output, the code is rerunned at this point. The results obtained were quite satisfactory. In the case of Lead, the negative Kerma factors are resultant of the large values of the (b Eijy) term in the equation

The solution adopted was to obtain this Kerma factor via the Kinematic limits settled in the HEATR module.

Referring to the second group of materials, the treatment was dif- ferent because gamma transport was not' . considered. The reasons for this are that the damage produced by gamma rays in materials is low compared to the damage induced by neutrons, and also that the system is sufficiently large for considering no photon looses (photon local deposition). The ma- terials evaluated are: Li6, ~i', C, Si, Pb, Fe, Cr, Ni, Mo, V, Mn, W, Q , 0 and H.

Besides Kermas, gcoup and group to group cross sections matrices, for these structural materials, the following activation cross sections have been evaluated.

ED - dpa = - 2Ed

- Helium production cross sections (through na reactions) - Tritium reproduction XS (only Li6 and Li7 through naT reactions)

- - Hydrogen production XS (through np reactions)

REFERENCES

( 1 ) "The NJOY Nuclear Data Processing System, Vols. I and 11". R.E. Mac- Farlane, D.S. Muir, R.M. Boicourt (LA-9303-M, ENDF-324) May 1982.

( 2 ) "NJOY: A Comrehensive ENDF/B Processing System" R . E . Mac Farlane, R . J . Barrett, D.W. Mu~r, R.M. Bolcourt (LA. NM 87545)

9 ~ 1 0 0 1 1 6

4 . - FINITE ELEMENT-DISCRETE ORDINATES METHOD FOR THE FOkKER-PLANK CHARGED

PARTICLES TRANSPORT EQUATION (DENIM)

A F i n i t e Element procedure t o f i n d t h e numerical, so lu t ion of Fokker- Planck equation has been developed i n order t o eva lua te the energy deposi t ion of charged fus ion products t o the thermonuclear plasma ( 1 ) . This top ic i s re l evan t f o r i g n i t i o n ca lcu la t ions on I n e r t i a l Confinement Systems. (2)

The b a s i s f e a t u r e of charged-part icle t r a n s p o r t is the s t rong space- energy coupling: t h e Fokker Planck represen ta t ion of t h e c o l l i s i o n s has continuous slowing-down and def lexion terms ( 3 ) . The treatment of these term v i a conventional methods (multigroup) gives poor r e s u l t s ( 2 ) . To solve t h i s problem a coupled d i s c r e t i z a t i o n scheme of t h e v a r i a b l e s with FEM has been s t ab l i shed . Its c a p a b i l i t y , f l e x i b i l i t y and accuracy has been f u l f i l l e d . The r e s u l t s a r e on exce l l en t agreement with o t h e r s obta ined a n a l i t i c a l l y ( 4 ) o r by o the r procedures ( S ) , and can be considered a second o rde r approach t o the so lu t ion .

Bas ica l ly , t h e FEM treatment of the equation i s based on the weak formulation and on t h e Galerkin procedure (6). Assuming the d i scon t inu i ty of t r i a l funct ions , t h e g lobal problem i s reduced t o a s e t of l o c a l problems with n a t u r a l boundary condi t ions . This r ep resen ta t ion has some advantages, d i r e c t so lu t ion , numerical s t a b i l i t y , good t rea tment of f l u x g rad ien t s , ..., and i ts inconvenient is the number of unknowns.

The following representa t ions of t h e angular f l u x a r e ava i l ab le :

i) Bi l inea r representa t ion r , E:

ii) Linear representa t ion r , E :

iii) Linear representa t ion r , E , p:

where 5 a r e 1 inea r . func t ions i n the f i n i t e element D.

The (iii) expansion has been developed i n o rde r t o make accura te c a l c u l a t i o n s when the incoming f lux i s highly a n i s o t r o p i c (confinement with unfocussed ion beams), i n these cases the d i s c r e t e o rd ina te needs a high quadrature order .

In ( i ) and ( i i) representa t ions the angular v a r i a b l e i s t r ea ted v i a d i s c r e t e o rd ina tes , computing t h e Fokker - Planck de f l ex ion term with extended t r anspor t approach ( 7 ) o r with "a c o e f f i c i e n t " method ( 3 ) . The

stability of solution is guaranteed for a mixed extrapolation - iteration procedure.

The method has been implemented on the NORCLA code (8) for ICF calculations. The coefficients of the equation are evaluated in to the code and it is possible to transport any particle in any medium without providing a cross section table.

Finally, a Lagrangian formulation is available for the efficient coupling of Nuclear and Thermo-hydrodynamics codes. (9).

REFERENCES

1.- Honrubia J.J., Aragonh, J.M. (to be published).

2.- Tran, T.M., Haldy, P.A. , Ligou, J., Lefvert, T. Atomkernergie 36,218 (1980)

3.- Mehlhorn, T.A., J. Comput. Phys. 138,l (1980)

4.- Haldy, P.A., Ligou, J., Nucl. Fus. 17,6 (1977)

5 . - Przybylski, k, Ligv~, J., Nucl. Sci. Eng. 81,92 (1982)

6.- Martin W.R, et al. Annuals of Nuclear Energy 8,633 (1981)

7.- Wienke, B.R., J. Quant Spetrosc. Radiat. Transfer. 28,4 (1982).

8.- Velarde, G. et al. Atomkernenergie 32,58 (1978)

9.- Wienke, B.R., Phys. Fluids. 17.6 (19741

5. CORE CONVERSION STUDIES FOR RESEARCH REACTORS . JEN (JUNTA DE ENERGIA NUCLEAR).

Rather extensive work in Reactor Physics and other areas

was devoted to the Lo-Aguirre swimming-pool reactor in Chile,

under collaboration between JEN and CCHEN (Comisidn Chilena de

Energla Nuclear). The work was mainly related to "core conver-

sion from 90 % enriched uranium (HEU) to 20 % enriched uranium

(LEU), but additional effort was neccesary because reactor

up-grading in several alternatives was considered, starting

from a power level of 13 MW.

The kind of calculations performed were very similar to

those described in past-year report, for a proposed 3-MW Ecua-

dorian Reactor. After an optiv.ization study to define important

characteristics of the fuel element, many analysis were done

for the design of the core and related systems; specific items

covered by neutronic calculations were: criticality, core fluxes,

burnup, peaking factors, fluxes in experimental facilities, reac-

tivity coefficients, kinetic parameters, reactivity defects,

etc.

Besides the Lo-Aguirre case, some preliminary calculations

have also be done for the JEN-1 swimming-pool reactor, related

to the proposed placement of graphite for berilium in a row of

reflector elements.

6. NEUTRONIC CALCULATIONS OF PWR CORES. (JEN)

The MARIA s y s t e m ( ' ) h a s been deve loped . T h i s code package h a s

been i n c l u d e d i n t h e C o o r d i n a t e d R e s e a r c h Programme (CRP) on

"Codes A d a p t a b l e t o Smal l o r Hedium S i z e Computer A v a i l a b l e i n

Deve lop ing C o u n t r i e s f o r In-Core Fue l Management" o f t h e I n t e r

n a t i o n a l Atomic Energy Agency.

I n g e n e r a t e s t h e c r o s s s e c t i o n s l i b r a r y f o r t h e d i f f u s i o n c a l - "

c u l a t i o n s w i t h burnup and feedback e f f e c t s and t h e km and M~

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

The MARIA sys t em i n c l u d e s t h r e e modules:

- PREWIM g e n e r a t e s t h e i n p u t d a t a f o r t h e f u e l assembly

c a l c u l a t i o n module f o r a l l - t h e f u e l assembly t y p e s i n

t h e c o r e .

- WIMS-TRACA g e n e r a t e s c o l a p s e d c r o s s s e c t i o n s v e r s u s burnup.

- POSWIM makes t r a n s p o r t c o r r e c t i o n s on t h e d i f f u s i o n

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

R e f e r e n c e s

(1) C. AHNERT, J.M. ARAGONES. PARIA System: P. Code Block f o r PWR F u e l Assembly C a l c u l a t i o n s . JEN-543 (1983).

7 , METHODOLOGY VALIDATION FOR THE DETERMINATION OF FLUENCE I N PRESSURE

VESSELS OF LWR'S. (JEN)

The v a l i d a t i o n of methods and a n a l y t i c a l t o o l s f o r t h e determinat ion of

neutron f luence i n p res su re v e s s e l s of comercial LWR's has been comple- (5) t e d .

Calcula t ions have been performed wi th t h e E U R L I B ' ~ ) l i b r a r y , ANISN-JEN ( 2 )

and DOT ~ . ~ I E - J E N ( ~ ) codes f o r t h e PCA "bl ind t e s t " (4)

The method i s based on a s p a t i a l f l u x s y n t e s i s , oh ta in ing r e a c t i o n r a t e s

f o r s e v e r a l ex-core d e t e c t o r s from t h e dosimetry f i l e of ENDFIB-IV and

ca l cu la t ed f luxes .

References

(1) J. PERA; EURLIB: Una l i b r e r i a acoplada de secc iones e f i c a c e s para

neutrones y gammas e specyf i ca para c5 lculos de b l inda je . MEMO JEN/TCR/A

01-81 (1.981).

(2) J . PEgA; ANISN-JEN. C6digo unidimensional de t r a n s p o r t e por e l mstodo

de ordenadas d i s c r e t a s . Manual de usuar io y da tos de en t rada .

(3) J . PERA; DOT 3.5/E-JEN: C6digo bidimensional de t r a n s p o r t e por e l m6-

todo de ordenadas d i s c r e t a s . Manual de usuar io y da tos de entrada.

MEMO J E N / T C R / A 02-83 (1983).

(4) Technical L e t t e r f o r t h e PCA "Blind Test". L. MILLER e t a l . Oak Ridge

National Laboratory. May 1979.

(5) J. BROS, J. PERA; Validaci6n de mstodos de c5 lculo por t r a n s p o r t e

pa ra l a de tern inac i6n de l a f l u e n c i a neut r6nica en v a s i j a s de r eac to res .

To be presented a t t h e V I I I ~ e u n i 6 n Anual de l a Sociedad Nuclear Espa-

iiola. Dic. 1983.

A nuclear c r i t i c a l i t y s a f e t y a n a l y s i s of a sh ipping cask f o r t r a n s p o r t

and s t o r a g e of spen t f u e l has been c a r r i e d out .

The WIMS-TRACA code'') has been used t o gene ra t e zone averaged multigroup

c ross s e c t i o n s , t h e Keff c a l c u l a t i o n was performed wi th t h e d i s c r e t e ord i - (2) n a t e s code TWOTRAN-GG .

I n o rde r t o achieve t h e requi red s u b c r i t i c a l i t y (Keff<0,95) t h e cask is

designed t o con ta in b lack absorbers .

References

(1) C. AHNERT; Programa WIMS-TRACA para e l c6 lculo de elementos combus-

t i b l e s . Manual de usuar io y da tos de en t rada . JEN-461.

(2) C. AHNERT y J . M . ARAGONES. E l c6digo de t r a n s p o r t e bidirnensional

TWOTRAN-GG. JEN-512.

9. SHIELDING (JEN)

The whole e f f o r d i n t h i s a r e a h a s been d i r e c t e d t o t h e i n p l e m e n t a t i o n

of t h e modular s y s t e m NJOY'~) t o t h e UNIVAC 1100. The aim is t o make

a new module which would p r e p a r e a l i b r a r y i n few groups f o r t h e Monte

C a r l o code TIMOC.

S e v e r a l modules have a l r e a d y been implementad: MODER, UNRESR, HEATR, BROADR,

THEbNR, RECONR, GROUPR and N J O Y .

Refe rences

( 1 ) R.E. MacFarlane, D.W. B o l c o u r t ; The N J O Y Nuc lea r Da ta P r o c e s s i n g

System, LA-9303-N (ENDF-324) May 82.

SWEDEN

Swedish Con t r ibu t ion t o t h e NEACRP a c t i v i t y

r e p o r t

POWER REACTOR LWR - Code development

(K Ekberg, S tudsv ik Energ i teknik AB)

The 2D d i f f u s i o n code MBS has been r e l e a s e d f o r

use by customers on CYBERNET. V e r i f i c a t i o n

s t u d i e s on a Swedish and on a US r e a c t o r are i n

p rog res s .

A new v e r s i o n of t h e MICBURN code f o r micro-

s cop ic burnup of g a d o l i n i a loaded f u e l p i n s ,

MICBURN-2, has been r e l e a s e d .

A f u r t h e r developed ve r s ion of t h e we1

c e l l and assembly code CASMO, denoted CASMO-2E,

h a s been i n s t a l l e d on CYBERNET and e lsewhere .

The enhancements a r e mostly i n t h e a r e a of PWR

a p p l i c a t i o n .

A l i n k i n g code MBSLINK between CASMO and MBS h a s

been developed.

Development work on CASMO, MBS, SIMULATE and

a u x i l i a r y codes con t inues .

R e a c t o r p h y s i c s c a l c u l a t i o n s on c l o s e - p a c k e d LWR

l a t t i c e s

( E r i k J o h a n s s o n , S t u d s v i k E n e r g i t e k n i k AB)

T h i s p r o j e c t d e a l s w i t h c a l c u l a t i o n s on Pu

r e c y c l i n g i n c lose -packed PWR l a t t i c e s . The

i n i t i a l work, i n c l u d i n g some p r e l i m i n a r y t e s t s

o f t h e c a l c u l a t i o n a l model , i . e . t h e c e l l code

CASMO and i t s a s s o c i a t e d l i b r a r i e s , h a s b e e n

d e s c r i b e d e a r l i e r - i n STUDSVIK Repor t

NR-82/126. S i n c e t h e n a s t u d y h a s been pe r fo rmed

o n t h e u s e o f such c lose -packed l a t t i c e s i n t h e

Swedish 1 2 - r e a c t o r s y s t e m ( 9 BWRs and 3 PWRs) - b o t h i n a p h a s e o u t v e r s i o n and i n a v e r s i o n

d e v e l o p i n g i n t o a n a s y m p t o t i c s t e a d y - s t a t e

o p e r a t i o n .

W e do n o t g i v e any d e f i n i t e f i g u r e s h e r e f o r

t h e s a v i n g s of n a t u r a l u ran ium and s e p a r a t i v e

work b e c a u s e t h a t would r e q u i r e a d e t a i l e d

d e s c r i p t i o n . J u s t a s a n i n d i c a t i o n w e can s a y

t h a t t h e c l o s e - p a c k e d l a t t i c e i n t h e s t e a d y -

s t a t e c a s e would r e q u i r e a b o u t 50 % less n a t u r a l

u ran ium and a b o u t 4 0 % less s e p a r a t i v e work t h a n

would once- through o p e r a t i o n i n t h e normal

l a t t i c e . These v a l u e s were o b t a i n e d unde r t h e

a s s u m p t i o n o f a 3 % l o s s o f Pu on e a c h re-

processing/refabrication o c c a s i o n .

A t p r e s e n t ( s e p t 1983) some t e s t c a l c u l a t i o n s

a r e g o i n g on u s i n g measured r e s u l t s f rom

Wfirenlingen on Pu-loaded c l o s e - p a c k e d l a t t i c e s

C a l c u l a t i o n s on t h e change from 9 3 % t o 20 %

enr iched uranium i n t h e R2 r e s e a r c h r e a c t o r

( E r i k Johansson, S tudsv ik Ene rg i t ekn ik AB)

The convers ion of t h e R2 r e s e a r c h r e a c t o r a t

STUDSVIK t o t h e use of 2 0 % enr iched uranium has

been s t u d i e d f o r some time. The i n i t i a l r e a c t o r

phys i c s work, desc r ibed i n STUDSVIK Report

NR-82/136 has been fol lowed by f u r t h e r s t u d i e s ,

i n c l u d i n g beam tube i n v e s t i g a t i o n s , a s accounted

f o r i n STUDSVIK Report NR-83/261. The c a l c u l a -

t i o n a l work i s now cont inued by t h e RERTR group

a t Argonne Nat iona l Laboratory.

MBS - a program for two-dimensional nuclear analysis of PWR cores

(Olov Norinder, Swedish State Power Board)

The computer program MBS has been designed

specially for the calculation of the power

distribution anong the fuel pins of PWRs.

Further MBS can determine the detector signals

that should be given among the input to the

computer program INCORE. This code calculates

the power distribution of the reactor core from

the signals of detectors that are moved through

the core.

MBS has been tested against measurements and

frestanding calculations for cycle 4 and 6 of

Ringhals 2. MBS has shown good results in the

testing.

The aim of the current phase of the project has

been to further develop and test MBS so that the

program can be used for ICFM work saLely and

rationally. In this context the program has been

extensively documented. In the work the demands

of the quality assurance has been particularly

considered.

- 125 -

I n Core Fuel Management a t Sydkra f t

(J 0 Gustafsson, Sydkraf t AB)

Together w i th ASEA ATOM SYDKRAFT has t h e ICFM

r e s p o n s i b i l i t y f o r Barsebeck Uni t 1 and 2. There

a r e a number of a c t i v i t i e s of which t h e most im-

p o r t a n t a r e :

- Core fo l low c a l c u l a t i o n s

- Refuel ing a n a l y s i s

- Fuel des ign s t u d i e s

- Economic a n a l y s i s

The b a s i c ins t rument t o c a r r y o u t t h o s e

a c t i v i t i e s i s t h e POLCA/CASMO system. POLCA i s a

3D nodal c o r e c imula t ion code.

CASMO i s used t o g e n e r a t e nodal 2D c r o s s

s e c t i o n s which a r e used a s i n p u t t o POLCA. We

a l s o use CASMO i n our s t u d i e s t o op t imize f u e l

des igns . The code has a l a r g e f l e x i b i l i t y , and

i t i s f r e q u e n t l y used t o s tudy t h e i n f l u e n c e on

k - i n f i n i t y and i n t e r n a l power peaking f a c t o r s

when a p e r t u b a t i o n i s in t roduced ( eg channel-

bowing, h igh tempera ture i n f u e l p i n s , e x t r a

Zr-pin o r wate r -ho le ) .

Addi t iona l t o t h e POLCA system SYDKRAFT has

developed some s imple codes used f o r adminis t ra -

t i v e purpose, eg code f o r upda t ing f u e l r eco rds .

The fueL r eco rds c o n t a i n in format ion about

burnup and channel l o c a t i o n each c y c l e e t c .

Furthermore SYDKRAFT has developed codes for

calculations of power histories, rest power,

averaged k-infinity of the core. These codes are

based on data from POLCA calculations and are

mainly administrative routinescombined with a

simple physical model. These codes are imple-

mented on a NORD 500 computer.

Effectiveness of cylindrical control rods

(D.C. Sahni* and N.G. Sjostrand,

Department of Reactor Physics,

Chalmers University of Technology,

5-412 96 Goteborq, Sweden)

The neutron flux distribution has been calcu-

lated around a cylindrical, totally absorbing

rod immersed in moderators of various size and

degree of absorption. Isotropic scattering was

assumed as well as a uniform source distribu-

tion. From the results obtained wlth a standard

ANISN code the linear extrapolation distance was

derived. It was found necessary to introduce a

correction for the fact that ANISN does not give

an exact balance between production and loss of

neutrons. After such a correction the linear

extrapolation distance could be obtained to an

accuracy of 1-2 %. Through using two different

definitions of the extrapolation distance it was

possible to reproduce the values of Pellaud

(1968) and Isakova (1968) and to explain the

difference between their results. (Ann. Nucl.

Energy 10, 351, 1983)

Complex t ime e igenva lues of t h e one speed

neu t ron t r a n s p o r t equa t ion f o r a homogeneous

sphere

(E.B. Dahl and D.C. Sahni*)

The t ime e igenva lue spectrum of t h e one-speed,

i s o t r o p i c s c a t t e r i n g neutron t r a n s p o r t equa t ion

has been s t u d i e d f o r a homogeneous sphere w i th

vacuum boundary cond i t i ons . There i s a c l o s e

r e l a t i o n s h i p between t h e t ime e igenva lue problem

and t h e c r i t i c a l i t y problem of t h e t i m e indepen-

d e n t equa t ion f o r t h e same model. I t i s shown

t h a t t h i s r e l a t i o n h o l d s even when t h e t i m e

e igenva lues a r e complex. Using C a r l v i k ' s method

t o s o l v e t h e c r i t i c a l i t y problem, it is shown

t h a t complex t ime e igenva lues do a c t u a l l y e x i s t

f o r t h i s model problem. Thus, t h e r e a l e igen-

v a l u e s found by van Norton do n o t form t h e

complete spectrum. (To be publ i shed i n Transpor t

Theory and S t a t i s t i c a l Phys ics . )

Heterogeneous treatment of water gaps and

control rods in core calculations

Nodal models for calculation of core power

distributions are conventionally based on

assemblywise homogenized procedure may introduce

substantial errors when strong heterogenities,

like control rod blades in the BWR, are present.

By extending the cross section homogenization

over the fuel pin area only and treating the

water gaps and control rods explicitely, using

response matrix concepts, the error can be

substantially reduced. It is shown how the

explicit treatment of the gaps can be incorpo-

rated into existing nodal models. The proposed

method has the side-benefit of producing cross

sections for the fuel pin area that are fairly

insensitive to the presence of control rods. The

gap responce matrices depend only weakly on the

lattice design, void content, and fuel burnup.

* Permanent address: Theoretical Physics Division, Bhabha Atomic Research Centre, Bombay 400 085, India

* * Permanent address: ASEA-ATOM, S-72 104 VasterAs, Sweden

REACTOR PHYSICS ACTIVITIES IN SWITZERLAND

October 1982 to September 1983

P. Wydler

1. Introduction

SWITZERLAND

The reactor physics activities reviewed in this paper have been carried out

at the Federal Institute for Reactor Research (EIR), except for the pre-

parations for the fusion-fission hybrid blanket experiment LOTUS which is

undertaken at the Federal In5titute of Technology, Lausanne. Because of

the needs of the Swiss reactor operators, Light Water Reactor orientated

physics has remained an important subject, but the majority of the available

funds are now concentrated on an experimental programme in which the PROTEUS

reactor is used to investigate the physics properties of Light Water High

Converter Reactor (LWHCR) lattices. Encouraged by the progress with a

small district heatrng scheme called REFUNA, which will start to deliver

heat from the Beznau reactors before the end of this year, a study of

a new type of low power distrlct heating reactor has been initiated and

carried through its first stage. Another relatively new development is

the extension of the activities in the direction of fusion problems with

0 emphasis on blanket physics.

2. Experimental Studies on LWHCR Lattices

Phase I of the PROTEUS-LWHCR experiments was conducted during the 14-month

period August 1981 to September 1982. The planning of the experimentai . programme, as well as the nature of the early results obtained, were discussed

in previous NEACRP review papers.

Nearly 80% of the time available for the Phase I measurements was devoted

to obtaining integral evidence on reactivity and reaction rate ratio varia-

tions with moderator voidage in a reference tight-pitch LWR lattice with

an average Pu-fiss enrichment of about 6%. A NEACRP-A paper for the current

Meeting presents the detailed experimental results and shows how they

may be useful in resolving the conflicting calculational results that

have been reported in the past for the void coefficient characteristics

of LWHCR's.

Other experiments carried out in the Phase I programme include measurements

in a small central test zone simulating a higher (8% Pu-fiss) enrichment,

as well as a limited set of small-sample reactivity measurements in the

reference lattice. The final analysis of these Latter experiments is

currently in progress.

In spite of its usefulness in providing a first set of integral results

for testing LWHCR physics calculations, the PROTEUS-LWHCR Phase I programme

has had some inherent shortcomings. These have stemmed largely from the

constraints imposed by the available fuel materials and include, for example,

the 2-rod nature of the experimental lattice, the use of Pu of Magnox

reactor origin and the relatively small test zone size. Moreover, the

total time available for the experiments severely restricted the range

of configurations that could be investigated.

In view of the above shortcomings it is planned to carry out a more comprehen-

sive, second phase of PROTEUS-LWHCR experiments during 1985 - 88. New,

LWHCR-representative fuel elements have been ordered for the purpose,

and these will be manufactured in the F.R. of Germany during 1984. Apart

from strengthening the currently established experimental base for LWHCR

~ 0 i d coefficient calculations, the new programme of measurements will

also deal with other power reactor features such as control rod worths,

soluble boron effectiveness and the effects of blanket zones.

Some results for the LWHCR Phase I experiments have already been reported

externally (Refs. 1 to 3).

3. Study of a Low Power District Heating Reactor

A study for a new type of low power district heating reactor has been

performed. The reactor has a homogeneous core, rated at a thermal power

of 10 MW, and was designed to supply a community of 2'000 to 5'000 inhabitants

0 with hot water at a temperature of 110 C (Of the communities in Switzerland

this size is the most numerous). A preliminary cost analysis indicates that

the reactor can be expected to be competitive with conventional thermal

energy sources.

To improve the safety features of the reactor an integrated design was

chosen (i.e. all components of the primary circuit including the intermediate

heat exchanger are enclosed in a single vessel) and the primary circuit

was designed for natural circulation cooling at full power. The cylindrical

liquid reactor core has a volume of about 2500 1 and operates at a mean 0

temperature of 140 C. The fuel is a solution of enriched uranium sulfate

(UO SO ) in water. Although the optimum enrichment is 60%, the reactor 2 4

can operate satisfactorily with 20% enriched fuel.

The physics design methods include MICROX cross section condensations,

two-dimensional FINELM diffusion theory calculations, burnup calculations

within the modular code system RSYST, and control rod calculations using

the SURCU transport code.

Using these methods, beginning-of-life 2 3 5 ~ inventories of 48.7 and 42.5 kg

were predicted respectively for the 20% and the 60% fuel enrichment. The

conversion ratio i.s low (< 0.1), but this has the advantage of only marginal

plkonium producti.on. Over a two-year reactor cycle at a load factor of

0.5, this corresponding to a burnup of 50'000 MWd/t(HM), the reactivity

loss was calculated to be 12.4 $ for the optimum enrichment. Additional

reactivity margins are needed to compensate for the relatively large tempera-

ture effect and the (negative) void effect due to the dissociation of

water. In total, this necessitates a relatively large reactivity compensation,

which is effected using 22 control rods.

Due to the relatively high enrichment the reactor has a small (but still

negative) doppler coefficient. However, a sufficiently large prompt negative

reactivity feedback is ensured by the temperature-density coefficient.

The dynamic properties of the reactor were studied for the cases of a

(secondary circuit) loss of flow and a fast control rod withdrawal. In

both cases the behaviour of the reactor is benign.

A more detailed description of the concept is given in Refs. 4 and 5

4. Reactor Noise Analysis

The activities in the reactor noise analysis field concentrate on measure-

ments of two-phase flow parameters with emphasis on measurements of flow

velocities in BWR fuel rod bundles near an instrument tube. The aim of

the work is mainly to gain the understanding, necessary for the interpre-

tation of fluid velocities measured by noise analysis, and to investigate

whether noise analysis measurements can be used for reactor monitoring

purposes and for the verification of two-phase thermohydraulics codes.

Almost all two-phase flows have a well established radial velocity and

void distribution profile. From two-phase flow models it is possible to

obtain the fluid velocity averaged over some cross-sectional area. It

might also be expected that some average velocity could be obtained from

the peak of a cross correlation function measured with two sensors. In

practice more than one peak is found and no simple average velocity is

obtainable. A better understanding of these signals has now been obtained

from out-of-pile measurements in an air-water test loop in which both

the radial velocity profile as well as the "summation velocity" obtained

from cross correlations of light beams have been determined (The radial

measurement involved a local light reflection probe which could be moved

to different radial positions lnside the flow). These studies are described

in Refs. 6 and 7.

In earlier measurements in the upper part of a BWR core such multiple

peaks in the cross correlation function had been observed. These can now

be interpreted as being due to a pronounced radial velocity profile in

the bundles around the instrument tube. Within certain limitations, by

simulating the detector signals with respect to a radial velocity profile

as given by an advanced subchannel analysis code (like THEWIT) and comparing

the resulting noise-analytical functions obtained from this simulations

with the equivalent ones inferred from reactor measurements, one is able

to "recover" the radial velocity profile in the bundles and use the results

for the verification of advanced subchannel analysis codes. A publication

on this work is about to appear (Ref. 8). 91100135

5. Spallation Neutron Source Studies

In connection with the proposed spallation neutron source at the 590 MeV

high-current proton accelerator of the Swiss Institute for Nuclear Research

(SIN) reactor physics and shielding methods are being applied for optimizing

the source design. The shield of the source is composed of 4 m of cast 17

iron and 1 m of concrete and was designed for a source strength of 10 n/s.

The performance of this shield has recently been reanalyzed using new

cross section data and improved calculational methods.

a For this purpose two high energy cross section libraries originating from

the Oak Ridge and Los Alamos National Laboratories were modified and

validated with the help of a theoretical benchmark problem specified jointly

by EIR, SIN and KFA Jiilich. Together with an experimental source spectrum

(measured at a mock-up of the target) the data were used in S transport N

theory calculations of the neutron and gamma dose rates in the shield.

The results of the study are given in Ref. 9. From the spatial dependence

of the dose rates it could be deduced that an increase in the steel to

concrete thickness ratio would allow the outer radius of the source to

be reduced by as much as 10% (This optimization did not include cost

considerations). In the study the usefulness of adjoint calculations for

determinmg surface dose rates due to monoenergetic source neutrons was

demonstrated.

The application of the SIN type spallation neutron source for simulating

the irradiation of the first wall in a fusion reactor has been investigated

further. Neutron induced helium production and atomic displacement rates

in aluminium and stainless steel samples were calculated using the neutronic

data and methods described above together with damage response functions

generated with the NJOY and DON codes. A particular objective was to assess

reflector gains for different materials such as Fe, Ni and Pb and the

influence of these reflector materials on the helium-production to displacement

ratio (or so-called CTR parameter).

The results reported in Ref. 10 show that the reflectors produce a useful

increase in the atomic displacement rate, but do not noticeably affect

the He production rate. He production and displacements per atom were

calculated to be about five times smaller than in a fusion reactor. On

the other hand, the spallation source considered in this study compares

favourably with currently available neutron sources, offering a ten fold

increase in the He production rate. The CTR parameter was found to lie

in the range 5 to 13, depending on the reflector, che latter value correspon- .

ding to an unreflected configuration and being typical for a fusion reactor.

6. Fusion Blanket Studies

Over the past two years EIR has developed methods and tested nuclear data

needed for the physics analysis of the blankets of fusion and fusion-fission

hybrid reactors with magnetic confinements. Although for the blanket physics

the computational approaches are basically the same as those used for

fission reactors, difficulties arise due to the presence of additional

nuclides, new reaction types, different neutron spectra, and novel geometric

configurations. A particular problem is the adequate prediction of tritium

breeding, which is not only affected by the lithium cross sections but

also by the n,xn cross sections of various multiplying materials.

In cooperation with General Atomic cross sectson sensitivity and uncertainty

analysis studies were made for the European INTOR and the U.S. FED design

of a fusion reactor (Refs. 11 and 12). An extension of this work inclu-

ded a comparison of the performance of the data libraries DLC-37,

VITAMIN-C/DLC-41, VITAMIN-C/MACKLIB-IV and the Los Alamos NJOY fusion

library (Ref. 13). Furthermore, three of the aforementioned libraries, together

with various transport theory approximations for the blanket calculation,

were tested for the hybrid design of a Tandem Mirror Reactor originating

from the Lawrence Livermore Laboratory and General Atomic (Ref. 14).

The methods have been applied for the neutronic analysis of a fusion-

fission hybrid reactor based on the UK design of the Reversed Field

Pinch Reactor. The EIR concept of this reactor is characterized by

either an aluminium or copper first wall/shell, which acts as a neutron

multiplier, and a He cooled hybrid blanket consisting of a Th metal

multiplier and breeder, a LiO tritium breeder and a stainless steel 2

reflector zone. Helium is used as a coolant to minimize the non-fertile

neutron captures. The study (reported in Ref. 15) showed that from the ,

neutronics point of view the concept is feasible. In addition to being

self-sufficient in the tritium fuel, the reactor breeds 0.7 kg of 233"

0

per MW(th)-year. However, an assessment of the radiation damage indicated

that the lifetime of the first wall is not adequate. The possibility

of using a thinner first wall is therefore being investigated. a In view of possible future design studies at EIR and planned hybrid

blanket experiments at the Swiss Federal Institute of Technology at'

Lausanne, the EIR computational methods are being further refined. With

the support of the 110s Alamos National Laboratory and General Atomic

a cross sectlon generation and calculational scheme including NJOY/TRANSX-

EIR/MICROX, the three-dimensional Monte Carlo codes MCNP and NMTC and

the two-dimensional finite element discrete ordinates code TRIDENT for

toroldal blanket geometry is being developed and tested.

7 . LOTUS Fusion-Fission Hybrid Blanket Experiment

-

The preparations for the fusion-fission hybrid blanket experiment LOTUS

at the "Institut de GQnie Atomique" of the Federal Institute of Technology,

Lausanne, are progressing well. The Haefely neutron generator with a <

source strength of 5 . 1 0 ~ ~ n/s is bang installed in the test cavity

and the various components of the blanket, which has the form of a 85 cm

thick, 100 cm high and 100 cm wide parallelepiped, are currently being

fabricated. The first experiments are expected in early 1984.

Starting at the source end, the reference blanket is composed of a 1

to 3 cm thick stainless steel sheet to simulate the first wall of a

fusion reactor, a 100 nun thick neutron multiplier zone made of lead

plate, a 35 nun thick spectrum adjustment zone of lithium carbonate,

a 277 nun thick 2 3 3 ~ breeding zone of thorium oxide, a 150 mm thick

tritium breeding zone of lithium carbonate, a 250 mm thick graphite

reflector and, finally, a 35 mm thick lithium carbonate absorber zone.

The purpose of the spectrum adjustment zone is to harden the neutron

spectrum in the Tho breeding zone and thereby maximize the net rate 2

233 of U production. The absorber zone behind the reflector captures

thermal leakage neutrcns which would otherwise be lost.

The Tho in the form of rods is obtained on loan basis from the Bhabha 2

Atomic Research Centre, India. The lithium carbonate is encased in rectangu-

lar boxes made from extruded aluminium channel. Although lithium carbonate

is not a viable material for fusion power reactor applications, it is

an adequate substitute for the purpose of the LOTUS experiments, which

are primarily intended as a neutron physics benchmark. Compared to

lithium oxide it has the advantage that high purity material can be

purchased at an acceptable price.

Parti-cular emphasis is being put on adequate diagnostic techniques.

These include neutron spectroscopy (NE-213 scintillators and proton

recoil telescopes) and integral reaction rate measurement techniques,

the latter being developed and adapted in collaboration with EIR. For

Th(n.7) , Th(n,f) and Th(n,Zn) measurements, methods will be applied

which had been developed earlier in connection with a zero-power reactor

physics programme on thorium-bearing fast reactor lattices at the PROTEUS

reactor. To measure tritium production rates two independent techniques

utilizing a liquid scintillator method and the self-irradiation of TLD's

are being tested.

The present status of the LOTUS experiments is described in Ref. 16.

A detailed account of the diagnostic techniques and other complementary

information is presented in a NEACRP-A paper for the current Meeting.

1. K. Q'iz. R. Chavla and R. Seiler "A LUHCR Void Simulation Expriment Using m u t h e m " Trans. ANS. 4Q . 558 (19831

2. E. Wettergott, R. Chavla and K. Gmik "Analysis of Test lattice mperiments in che Light-uater- High-Conversion-Ream PRmEIIS" Report EPRI-UP-3190 119831

3. R. Chawlai. X. Gmur, n. Hager. E. Hettergott and R. Seller "Measurement of the Mean k. Vold OxfErcient for an L m R I a t t l ~ e between 0-100 + Void" Jahrestagung Kernteotnlk. Berlin (1983)

I . W. Selfrltz et al. "Space Heatlng for Small Cwmtmltles vlth a Homogeneous Heat Prcduclng Reactor (HHRI" Paper to be presented at the I m Technical C-lttee Ueetmg and Workshop on Nuclear Heat App1,catlon. Krakov, Poland. 5 - 9 December, 1983

5 . W. Seifritz ec al. -Ein hooogener Helzreaktor (HHRI kleiner Kistung (10 WWthl fiir dle nukleare NamSmeversorgung lnterner EIR-Berich= (19831

6. 6 . Th. Analytis and D. LiiblKnmeyer " A novel cross-cor~elation technique for the detemi- nation of radial velocity profiles in tnrphase flows" EIR-Bericht Nr. 483 (19831

7 . G. Th. Analytis and D. Liibbesmyer .A study of annular flows with buMles in the liquid ring and entrained dmplets by means of stochastic analysls techniques" EIR-Bericht NT. 489 11983)

8. C. Th. Analytis and 0. Liibbesmeyer 'hrrPhase Flow Velocity Measurements in she Upper Part of a BUR; The Importance of mlti-Dimensioml Effects •’01 their Interpretation'' To be published in Proc. of the Thernohydrwlic Division of the ANS Winter Metlng, San Francisco I19831

9. V. Hezrnberqer and P. Stiller "SIUQ Bulk Shield Analysis Using the S,-method" ICANS-VII-Heeting for the International Cooperation on Advanced Neutron Sources, Chalk River. Canada. 13 - 16 September, 1983

10. V. Herrnheeger, P. Stiller and M. Victoria "Some Estimtes of me Fusion Radiation Danage sinulacion by Spallation Neutrons" Sixth International Conference on Radiation Shielding. Tokyo, 16 - 20 m y . 1983

11. S. Pellonl and E.T. Cheng "cross Sectlo" Sensktlvlty Srudles for Fusmn Blankets Inoorpxatmg Lead Neutron Multapl~er" PrOE. Int. Conf. on Nuclear Data for Sclence and Technolqy, Antveep. p. 331 (19821

12. S. Pelloni and E.T. Cheng "Cross Section Sensitivity Study for U.S. Fusion mgineering Device" To be published in Proc. Pifth Topical meting on the Technolcqy of miion Energy. LVloxville. Temeeree, 24 - 28 April. 1983

13. S. Pelloni, J. Stepirnek and D. Dudriak "Interoompar~wn of Yuclear Data L~brary Sources. Group Structures a d collaming Spectra for IWR-DC" 'le be published in Proc, Fifth Topical keetin? on the Technology of rusion Energy. Knoxville. Tennessee, 26 - 26 April, 1983

14. D. J. mdziak, J. Stepanek. U.T. Urban and G. Friedrich "Comparison of Los Alalas WTXS, VITAMIN-C and DLC-37 Wultigroup Libraries for a Reference Fusion Nybrid Blanket" PIX. Int. Conf. on b l e a r Data for Science and Technology. Antwerp. p. 339 119821

15. J.F. Jaeger et al. "A Hybrid Blanket for a Reversed Field Pinch Reactor" To be published in Proc. Course and uorkshop on nlrrar-sased and Tield-Reversed AQproashes to Mgnetic Fusion, Varenna, September 1983

16. P.A. Haldy et al. '"Present Status of the EPPL ISwissI Nsion-Fission experiment m S " Third International Conference on merging Nuclear Systems. Helsinki. 6 - 9 June. 1983

UNITED KINGDOM - 138 -

R e a c t o r P h y s i c s i n t h e U n i t e d Kingdom

3 R Askew J M S t e v e n s o n

1. GENERAL PROGRAMME

The p u b l i c e n q u i r y i n t o t h e p r o p o s e d PWR a t S i z e w e l l r e o p e n e d d u r i n g S e p t e m b e r . T h i s f o l l o w e d a summer r e c e s s a f t e r a s i t t i n g o f 104 d a y s . S u b j e c t t o a p p r o v a l o f t h e p r o p o s a l , c o n s t r u c t i o n would now b e e x p e c t e d t o s t a r t i n 1 9 8 5 , a y e a r l a t e r t h a n o r i g i n a l l y p l a n n e d .

1 4 % o f t h e e l e c t r i c i t y g e n e r a t e d by t h e CEGB i n 1982-3 was f rom n u c l e a r s t a t i o n s . The i n c r e a s e compared w i t h t h e p r e v i o u s y e a r a r o s e l a r g e l y b e c a u s e o f t h e r e t u r n t o s e r v i c e o f f i v e o f t h e s i x Magnox s t a t i o n s which had b e e n s h u t down f o r i n s p e c t i o n a n d r e m e d i a l r e p a i r o f t h e r e a c t o r b e l l o w s . The s i x t h u n i t is due t o c o m p l e t e i t s i n s p e c t i o n and r e p a i r s t h i s y e a r . I n a d d i t i o n t h e AGR a t H i n k l e y P o i n t B a c h i e v e d a 9% i n c r e a s e on i t s p r e v i o u s b e s t p e r f o r m a n c e a s a r e s u l t o f m o d i f i c a t i o n s t o e n a b l e o n - l o a d r e f u e l l i n g t o t a k e p l a c e . F u r t h e r deve lopment work w i l l make i t p o s s i b l e t o r e f u e l a t h i g h e r l o a d s .

I n November 1982 t h e Government c o m p l e t e d i t s r e v i e w o f t h e F a s t R e a c t o r , a n d g a v e t h e go-ahead t o t h e deve lopment programme b u t a t a r a t h e r s l o w e r ra te . The o r d e r i n g o f t h e f i r s t c o m m e r c i a l f a s t r e a c t o r s w i l l now n o t t a k e p l a c e u n t i l e a r l y i n t h e n e x t c e n t u r y . I n Sep tember 1 9 8 3 , t h e Government announced t h a t i t was o p e n i n g f o r m a l n e g o t i a t i o n s t o s e e k a g r e e m e n t on t h e j o i n t deve lopment o f f a s t r e a c t o r s w i t h F r a n c e , Germany, I t a l y , Be lg ium and t h e N e t h e r l a n d s .

2 . THERMAL REACTORS

F u r t h e r v a l i d a t i o n a s s e s s m e n t o f RETRAN a g a i n s t LOFT e x p e r i m e n t s h a s been c a r r i e d o u t . S i n c e t h e l a s t r e p o r t s i m u l a t i o n o f L6-5 and t h e more s e v e r e t r a n s i e n t s o f t h e L9 s e r i e s h a v e e m p h a s i s e d t h e i m p o r t - a n c e of s t e a m g e n e r a t o r m o d e l l i n g .

R e a c t o r t r a n s i e n t s t u d i e s h a v e c o n c e n t r a t e d on t h e l o s s o f o f f - s i t e power a n t i c i p a t e d t r a n s i e n t w i t h o u t s c r a m (ATWS) w h e r e a l l power t o t h e main c i r c u l a t o r s i s l o s t . The i s s u e s c e n t r e on t h e e f f e c t o f p r i m a r y v o i d a g e on n a t u r a l c i r c u l a t i o n and r e a c t i v i t y . Work on t h e l o s s of f e e d w a t e r w i t h o u t t u r b i n e t r i p ATWS was u s e d as a b a s i s f o r s e n s i t i v i t y and u n c e r t a i n t y s t u d i e s . T h i s work h a s b e e n r e p o r t e d t o t h e Sep tember 1983 ANS T o p i c a l M e e t i n g on normal a n d abnormal t r a n s i e n t s .

LWR-WIMSIJOSHUA s t u d i e s o f D o p p l e r c o e f f i c i e n t h a v e shown s e n s i t i v i t y t o t h e way i n which t h e f u e l t e m p e r a t u r e c h a n g e is o b t a i n e d . I t is i m p o r t a n t t h a t t h e c o o l a n t e n t h a l p y d i s t r i b u t i o n c h a n g e s are p r o p e r l y m o d e l l e d .

A b r i e f s t u d y h a s b e e n made o f t h e p a r a m e t e r s o f i m p o r t a n c e i n c a l c u l a t i n g t h e w o r t h o f g a d o l i n i u m b u r n a b l e p o i s o n p i n s ( 1 ) . T h e s e p a r a m e t e r s i n c l u d e d

( a ) number o f e n e r g y g r o u p s i n t h e l a t t i c e c a l c u l a t i o n (b) i n t e r v a l be tween l a t t i c e c a l c u l a t i o n s ( c ) i n t e r v a l b e t w e e n r e - e v a l u a t i o n o f p o i s o n p i n f i n e s t r u c t u r e ( d ) p o i s o n p i n r a d i a l s u b d i v i s i o n ( e ) e f f e c t o f s m e a r i n g t h e p o i s o n p i n p r i o r t o d i f f u s i o n c a l c u l a t i o n s ( f) m i n o r g a d o l i n i u m i s o t o p e s .

The c o n c l u s i o n was t h a t i t is d i f f i c u l t t o do b e t t e r t h a n 0 .5% i n r e a c t i v i t y a t a l l i r r a d i a t i o n s w i t h o u t u s i n g an i n o r d i n a t e amount o f comput ing time and t h a t t h e r e is l i t t l e p o i n t i n , f o r example , a v e r y f i n e r a d i a l p i n s u b d i v i s i o n u n l e s s a l l o t h e r a s p e c t s o f t h e c a l c u l a t i o n s are s i m i l a r l y d e t a i l e d .

With t h e u l t i m a t e o b j e c t i v e o f d e v e l o p i n g a r a p i d 3D c o r e t r a n s i e n t c o d e , e x p l o r a t o r y s t u d i e s h a v e been made w i t h a s p a r s e m a t r i x d i r e c t i n v e r s i o n t e c h n i q u e f o r s o l v i n g t h e t r a n s i e n t n e u t r o n i c s e q u a t i o n s . A t y p i c a l comput ing s p e e d a c h i e v e d is 4 s e c o n d s IBM 3081 cpu time f o r a 100 second r o d e j e c t i o n t r a n s i e n t w i t h o u t f e e d - back i n a 3D problem w i t h 1000 mesh p o i n t s . The comput ing time f inc reases r a p i d l y w i t h number o f meshes , b u t more s e r i o u s l y , t h e Crank-Nicholson time i n t e g r a t i o n which p e r m i t s l o n g time s t e p s is l e s s a c c u r a t e when f e e d b a c k e f f e c t s are i n t r o d u c e d . Thus t h e i m p l e m e n t a t i o n o f a s i m p l e Dopp le r f e e d b a c k model i n c r e a s e d t h e cpu time f o r t h e r o d e j e c t i o n t r a n s i e n t by an o r d e r o f magn i tude .

The WIMS 8 1 d a t a l i b r a r y c o n t i n u e s t o g i v e s a t i s f a c t o r y r e s u l t s . A whole c o r e s i m u l a t i o n o f o n e o f t h e commiss ioning e x p e r i m e n t s on Dungeness B AGR u s i n g t h e MONK5W Monte C a r l o c o d e w i t h WIMS d a t a gave an e i g e n v a l u e of 1 .0025 f 0.3%. T h i s was p a r t i c u l a r l y re- a s s u r i n g a s earl ier d e t e r m i n i s t i c c a l c u l a t i o n s u s i n g t h e SNAP d i f f u s i o n code had g i v e n k a 1 . 0 2 , t h u s c a s t i n g doub t on t h e WIMS 8 1 l i b r a r y . I t is now b e l i e v e d t h a t b l a c k r o d and s t r e a m i n g e f f e c t s are t h e c a u s e s o f t h e d i s c r e p a n c y , though t h i s h a s n o t y e t been d e m o n s t r a t e d i n d e t a i l . > .

The n e x t s t e p i n t h e a p p l i c a t i o n o f d i r e c t Monte C a r l o c a l c u l a t i o n s t o whole r e a c t o r c o r e s would l o g i c a l l y b e t h e t r e a t m e n t o f d e p l e t i o n . I n p r i n c i p l e i t would b e e x p e c t e d t h a t c y c l e l e n g t h c o u l d b e p r e - d i c t e d t o e q u i v a l e n t a c c u r a c y a t no more c o s t t h a n t h e c a l c u l a t i o n o f r e a c t i v i t y a t s ta r t of c y c l e . T h i s h a s been d e m o n s t r a t e d f o r g r e a t l y s i m p l i f i e d tes t p rob lems which have s t u d i e d t h e b e h a v i o u r w i t h b o t h n e g a t i v e and p o s i t i v e r e a c t i v i t y f e e d b a c k e f f e c t s , t h e l a t t e r a r i s i n g f o r example , when b u r n a b l e p o i s o n s a r e d e p l e t e d more r a p i d l y t h a n t h e a s s o c i a t e d f u e l .

E x p e r i m e n t a l s t u d i e s have been c a r r i e d o u t d u r i n g t h e l a s t y e a r i n c o l l a b o r a t i o n w i t h t h e B e r k e l e y N u c l e a r L a b o r a t o r i e s o f CEGB u t i l i s i n g t h e ze ro -ene rgy c r i t i c a l f a c i l i t y , BAGR, and t h e NESSUS i r r a d i a t i o n t h i m b l e l o c a t e d w i t h i n t h e c e n t r a l r e f l e c t o r o f t h e NESTOR s h i e l d i n g r e a c t o r a t W i n f r i t h . T h i s work h a s improved t h e a c c u r a c y w i t h which n e u t r o n a n d gamma e n e r g y - d e p o s i t i o n i n m o d e r a t o r g r a p h i t e c a n b e d e t e r m i n e d , a n d c o n s i s t e n c y w i t h * lo% h a s e v e n t u a l l y been a c h i e v e d . A n o v e l f e a t u r e o f t h i s work w a s t h e u s e o f micro- calorimeter which h a s been d e v e l o p e d i n c o l l a b o r a t i o n w i t h I m p e r i a l C o l l e g e t o measure e n e r g y - d e p o s i t i o n rates down t o a b o u t 1 0 p1Wat ts p e r gm. The programme is c o n t i n u i n g w i t h t h e a i m o f i n c r e a s i n g t h e s e n s i t i v i t y o f t h e d e v i c e t o c o v e r t h e r a n g e f rom 1 t o 1 0 p / W a t t s p e r gm. I t c o u l d t h e n b e u s e d i n p e n e t r a t i o n benchmark e x p e r i m e n t s which are c o n d u c t e d i n t h e ASPIS b u l k s h i e l d f a c i l i t y a n d also i n ze ro -ene rgy .crit ical e x p e r i m e n t s s u c h as t h o s e mounted i n t h e BAGR and DIMPLE.

3 . FAST REACTORS

3 . 1 ZEBRA and the CADENZA Programme

The ZEBRA reactor has now been shut down for about a year and the experimental team has moved to the DIMPLE reactor as discussed later in this paper. Maintenance of the ZEBRA plant is continuing to allow a return to fast reactor work when required.

A meeting was held to consider the calculations for the CADENZA benchmarks carried out in JAPAN ( 2 ) , USA ( 2 ) , Germany, France and the UK. As discussed in a separate paper, the discrepancies between the calculated and experimental k-values for the all-plate and extrapolated all-pin cores varied from 0 . 0 0 1 5 dk to 0 . 0 0 8 2 dk. Differences between the calculated reactivities for the homogeneous pin and plate cell compositions were also seen and it was agreed that all participants would calculate the perturbation worths of the various isotopic composition changes using homogeneous cell models. The results of these calculations are also compared in the same paper. It

a was also evident that the detailed modelling of the pin and plate cell heterogeneities was important. Two papers by Grimstone and Rowlands present the techniques used in the UK.

In the longer term, the participants have agreed to calculate the voided pin and plate cores, the worths of exchanging pin and plate elements in large and small zones, the effects of plate-cell heterogeneity on reactivity and reaction-rate distributions and reaction-rate ratios in the standard plate cell. Details of these measurements and of the experimental results have been issued to the participants.

The worths of removing or adding sodium in small zones of 9 elements were measured in all four assemblies. These worths were calculated using first order perturbation theory with the XYZ fluxes and exact1FOP corrections to the non-leakage, radial and axial leakage terms were then estimated from RZ mode The correction factors to these terms to bring them into agree- ment with the experiment are summarised in Table 1. Comparison

1-. of the factors for the sodium flooded cores with those from previous conventional ZEBRA assemblies shows similar values within the errors for the non-leakage term. For the leakage terms, the factors tend to be higher for the CADENZA cores, possibly due to the natural uranium breeders in these assemblies compared with the more realistic breeders in the earlier systems. For the voided cores the correction factors, particularly those for non-leakage and radial leakage, tend to be lower than those from corresponding versions with sodium present, which suggest that some caution is necessary in applying the factors in Reference 2 to power reactor situations where sodium is returned to a voided region.

The large-zone voiding measurements, which were made during the conversion to voided assemblies, have been compared with calculations using exact perturbation theory and XYZ diffusion-

t heo ry models. The r e s u l t s are shown i n Table 2 . In g e n e r a l t h e agreement between c a l c u l a t i o n and experiment is good, i n l i n e w i th t h e k-value p r e d i c t i o n s f o r f u l l - s c a l e vo id ing . For t h e p l a t e s i n f a c t , t h e agreement f o r t h e sodium worths is w i t h i n t h e exper imental e r r o r s . Applying t h e f a c t o r s ob ta ined from t h e comparisons f o r t h e 9-element v o i d s has a r e l a t i v e l y sma l l i n f l u e n c e on t h e agreement w i th experiment. I t is marg ina l ly worsened f o r p l a t e geometry and s l i g h t l y improved f o r p i n s , whi le i t is t h e a d d i t i o n a l u n c e r t a i n t i e s of about +3% a s s o c i a t e d wi th t h e c o r r e c t i o n s which provide t h e agreement w i th experiment. Nonetheless , t h e o v e r a l l conc lus ion is t h a t , f o r t h e CADENZA c o r e s , t h e c a l c u l a t i o n s p r e d i c t t h e worths of l a r g e zone v o i d s i n bo th f u e l geomet r ies q u i t e w e l l , t h e d i f f e r e n c e s between c a l c u l a t i o n and experiment corresponding t o a combination of _+5% u n c e r t a i n t i e s i n t h e non- Leakage and leakage c o n t r i b u t i o n s .

The e f f e c t i v e neu t ron source o u t p u t s from spontaneous f i s s i o n and ( a , n ) r e a c t i o n s i n ZEBRA f u e l have been measured i n t h r e e of t h e CADENZA assembl ies . The method used was based on t h e s t a n d a r d s u b c r i t i c a l moni tor ing technique whereby i n a s u b c r i t i c a l system l ~ h i c h is c l o s e t o c r i t i c a l , t h e power l e v e l is p r o p o r t i o n a l t o the source s t r e n g t h and i n v e r s e l y p r o p o r t i o n a l t o r e a c t i v i t y . Ca lcu l a t ed c o r r e c t i o n s were necessary t o r e l a t e t h e power l e v e l , a s measured by t h e multi-chamber scanning system, t o t h e t o t a l neutron produc t ion r a t e i n t h e assembly and t o a l l ow f o r t h e d i f f e r e n c e i n a d j o i n t weigh t ing of t h e neutron and f i s s i o n source$. The r e s u l t s , which a r e summarised i n Table 3 , suppor ted t h e conc lus ions from r e c e n t H a r w e l l work wi th i n d i v i d u a l ZEBRA compon- e n t s and, w i th in an u n c e r t a i n t y of f 5 % , confirmed t h e v a l i d i t y of c u r r e n t neutron source p r e d i c t i o n s f o r u n i r r a d i a t e d f a s t r e a c t o r f u e l .

Table 1 - Correction Factors for Terms of Sodium Worths Calculations to give Agreement with Small-Zone Experiments

Core

CADENZA Core 22, plates with sodium

CADENZA Core 23, pins with sodium

CADENZA Core 24, plates, sodium voided

CADENZA Core 25, pins, sodium voided

Cores 12, 13, 15 (Flooded plates and pins)

NOTE -

Non Leakage

Required Factors

Radial Leakage Axial Leakage

. - . -- Errors in the factors for the CADENZA cores arise from uncertainties in the experimental results (%2% of the major contributions to the ~erturbation), least square fitting to the sets of results and the assumption that the factors are independent of position in the core zones (%?%), the exactfFOP corrections (%2%), and the error associated with reactivity scale in each core (2%). There is, of course, also a systematic error common to all the cores, arising from uncertainties in the delayed neutron data used to calibrate the reactivity scale.

The factors for Cores 12, 13 and 15 are taken from Reference 2.

Table 2 - Comparison of Calculated and Ex erimental Worths of Large-Zone E Voiding (Units 10- dk/k)

Plates

Pins

Elements Voided

25

69

129

172

Non- Leakage

0.277

0.680

1.094

1.316

Calculated Worth

Radial Leakage

-0 .O43

-0.268

-0.810

-1.323

Axial Leakage

-0.177

-0.415

-0.635

-0.732

Total

0.057 (0 .O68)

-0.003 (0.018)

-0.350 (-0.332)

-0.740 (-0.730)

Experimental ~orths C-E

-0.003

0.016

0.007

-0.007

NOTES .-

1. The values in brackets are after the application of the factors from Table 1 for the 9-element voiding experiments to the three calculated perturbation contributions.

2 . The systematic error from the experimental reactivity scale is not included in the uncertainties shown.

3. Both cores contained %220 elements.

- 144 -

Table 3 - Comparison of Measured and Calculated Neutron Source Strengths

Assembly

Core 22 (metal)

Core 23 (mixed oxide)

Core 25 (mixed oxide)

Core 23 Core 22

:ore 25 Core 22

Experiment

(8.07 .t 0.51) x l o 7

(11.12 f 0.70) l o 7

(12.16 f 0.79) x l o 7

1.378 f 0.056

1.507 f 0.056

Calculation

7.90 x 10'

11.57 x lo7

12.23 x l o 7

1.465

1.549

Harmel l and W i n f r i t h a r e p a r t i c i p a t i n g i n t h e I n t e r n a t i o n a l F i s s i o n Mass and c o u n t i n g - compar ison which was set u p f o l l o w i n g compar i sons of r e a c t i o n - r a t e r a t i o s d e t e r m i n e d by d i f f e r e n t l a b o r a t o r i e s ( 3 , 4 ) . Two U235 d e p o s i t s we igh ing %350pg were p r e p a r e d a t H a r w e l l , c o u n t e d i n low geometry d e t e c t o r s a t Harwe l l a n d W i n f r i t h and t h e n s e n t t o ANL E a s t . P r e l i m i n a r y i n d i c a t i o n s a r e t h a t good agreement is found be tween t h e a l p h a e m i s s i o n r a t e s measured a t a l l t h r e e l a b o r a t o r i e s .

3 .2 C o n t r o l Rod Worths and t h e i r E f f e c t on Power D i s t r i b u t i o n s i n a Power R e a c t o r -

Comparisons have been made be tween c a l c u l a t e d and measured w o r t h s of boron c a r b i d e a b s o r b e r r o d s i n t h e P r o t o t y p e F a s t

0 R e a c t o r . The r e a c t o r h a s a r i n g of f i v e c o n t r o l r o d s i n t h e i n n e r c o r e and a r i n g of f i v e s h u t - o f f r o d s i n t h e o u t e r c o r e . The methods d e v e l o p e d t o p r e d i c t a b s o r b e r r o d w o r t h s i n f a s t r e a c t o r s , and t h e i r v a l i d a t i o n u s i n g ZEBRA e x p e r i m e n t s and t h e o r e t i c a l s t u d i e s have been r e p o r t e d p r e v i o u s l y ( 5 ) . These methods have been u s e d t o p r e d i c t t h e w o r t h s of t h e r o d s i n PFR. CIE was 1 . 0 7 f o r t h e c o n t r o l r o d s and 1 . 1 9 f o r t h e s h u t - o f f r o d s . T h i s r a t i o shows agreement w i t h t h e e s t i m a t e d un- c e r t a i n t i e s ( 5 % f o r t h e measurements and 6 % f o r t h e c a l c u l a t i o n s ) f o r t h e c o n t r o l r o d s , b u t t h e d i s c r e p a n c y f o r t h e s h u t - o f f r o d s is b e i n g i n v e s t i g a t e d f u r t h e r .

Another p a p e r c o n s i d e r s t h e e f f e c t s of t h e r e p r e s e n t a t i o n o f t h e PFR c o n t r o l r o d s a t o n e - t h i r d i n s e r t i o n (as a t t h e b e g i n n i n g of a c y c l e ) t h r o u g h o u t a r e a c t o r r u n . I n i t i a l c a l c u l a t i o n s have shown t h a t t h e a s s o c i a t e d e r r o r s i n c a l c u l a t e d f u e l c o m p o s i t i o n s , when u s e d i n s u b s e q u e n t f l u x c a l c u l a t i o n s p roduce e r r o r s i n f l u x e s o f ~ 2 % w i t h i n t h e c o r e f o r a s i n g l e o p e r a t i o n a l p e r i o d . T h i s e r r o r is domina ted n o t by t h e d i f f e r e n c e i n f i s s i l e c o n t e n t o f t h e c o r e b u t by t h e amounts of U238 c a p t u r e p r o d u c i i n t h e u p p e r a x i a l b r e e d e r . !f~?68 I 4 7

C a l c u l a t i o n s f o r a commercial f a s t r e a c t o r des ign have shown t h a t t h e model l ing of t h e movement of o p e r a t i n g r o d s and t h e i r u s e i n minimis ing peak subassembly power a r e important i n o b t a i n i n g t h e b e s t r a d i a l form f a c t o r . Enrichments chosen t o g i v e a h i g h e r peak subassembly power i n t h e i n n e r c o r e t han i n t h e o u t e r c o r e a t t h e end of c y c l e can h e l p improve t h e form f a c t o r a t t h e worst p o i n t ( s t a r t ) of c y c l e . These c a l c u l a t i o n s a r e d i s c u s s e d i n a s e p a r a t e paper .

- 3 . 3 D i s t r i b u t e d Hea t ing E f f e c t s

The method f o r c a l c u l a t i n g d i s t r i b u t e d h e a t i n g e f f e c t s i n TRIZ * geometry u s i n g d i f fu s ion - theo ry methods t o t r e a t neu t ron and gamma-ray t r a n s p o r t h a s been s u c c e s s f u l l y used wi th an e q u i l i b r i u m c o r e model of t h e PFR i n 60•‹ s e c t o r geometry. The gamma-ray d i f f u s i o n c a l c u l a t i o n used s i x t r i a n g l e s p e r subassembly i n t h e p l an and 20 a x i a l meshes, as f o r t h e neu t ron c a l c u l a t i o n . I t was concluded t h a t t h e e x t r a compl ica t ion and computing time r e q u i r e d t o u se a f i n e r mesh i n t h e gamma-ray c a l c u l a t i o n is n o t worthwhile when compared wi th t h e o v e r a l l accuracy of t h e d i f fu s ion - theo ry method. 37 and 13 groups were used f o r t h e neu t ron and gamma c l a c u l a t i o n s r e s p e c t i v e l y .

The method h a s a l s o been used wi th t h e s t a n d a r d PFR 6-group s t r u c t u r e t o t r e a t neu t ron d i f f u s i o n and t o c a l c u l a t e t h e gamma- r ay sou rce . Provided t h e neu t ron s p e c t r a used t o condense t h e neu t ron c r o s s - s e c t i o n s a r e s a t i s f a c t o r y , t h i s u se of a s i x neutron-group s t r u c t u r e i n t r o d u c e s e r r o r s i n t h e gamma h e a t i n g d i s t r i b u t i o n s i g n i f i c a n t l y smaller t han t h o s e i n t roduced by t h e u s e of d i f f u s i o n - t h e o r y methods. The e r r o r i n t roduced i n t h e non-gamma h e a t d i s t r i b u t i o n is a l s o sma l l (tl%).

The c a l c u l a t i o n of t h e h e a t d i s t r i b u t i o n s n e g l e c t i n g gamma t r a n s p o r t and i n c l u d i n g t h e energy c a r r i e d by t h e neu t rons as be ing d e p o s i t e d a t t h e p o i n t of neu t ron b i r t h i n t r o d u c e s s i g n i f i c a n t e r r o r s i n non f i s s i l e r e g i o n s .

T h i s work is d e s c r i b e d i n d e t a i l i n a s e p a r a t e paper t o t h e NEACRP . The above methods do no t a l l ow c a l c u l a t i o n s of r a t i n g s of i n d i v i d u a l components such as samples i n m a t e r i a l s - t e s t i n g sub- a s sembl i e s o r abso rbe r p i n s i n c o n t r o l r ods . T h i s in format ion is needed f o r exper imenta l a n a l y s i s o r s a f e t y assessment . Some a t tempt h a s been made t o model such a s sembl i e s u s i n g s imple c y l i n d r i c a l models b u t t h e s e a r e of very l i m i t e d a p p l i c a t i o n . I t is planned t o u se t h e MONK6 Monte-Carlo code which h a s been i d e n t i f i e d as s u i t a b l e f o r c a l c u l a t i n g t h e s e complex geomet r ies i n d e t a i l .

3.4 Bowing Predictions and Damage Gradients

The COSMOS workshop program BOWHIST has been used to compute subassembly distortions and interactions for the early stages of PFR operation. Analysis shows non-symmetric bowing when one subassembly in a cluster, having relatively li.ttle resistance to bending is forced out by the others with greater resistance to bending. It has also been found that the presence of a lowly-rated subassembly (eg. an experimental demountable sub- assembly) can induce thermal bowing in neighbouring standard subassemblies in the reverse direction to normal.

A new version of the CRAMP subassembly distortion code has been produced. The principal new feature is the representation of control rod distortion. The distortion of the rod and the guide tube are both computed, and the interaction between them, as a function of control-rod insertion. The program is being tested in the context of PFR operation. 0 The Nb93(n,n')Nb93m reaction is of interest in fast and thermal reactor investigations as a monitor of neutron damage. Its differential cross-section approximates to that for neutron damage in steel and the Nb93m half-life of 16 years ensures good integration of neutron fluence in long irradiations. Unfortunately there is a lack of experimental information about its cross-section above about 2.5 MeV.

In 1982 a collaborative experimental programme to measure the differential cross-section was undertaken using the Dynamitron accelerator at Birmingham University. The participants are; Reactor Physics Division at Winfrith, Nuclear Physics Division at Harwell, and the Radiation Centre at Birmingham. To date six energy points have been measured, spanning the range 1 MeV to 6 MeV; further measurements are planned for November 1983. It is expected that the target accuracy of +5% will be obtained for most of the energy points. The results obtained so far are in agreement with the calculations of Strohmaier and, in the lower energy region, with the experimental results from other workers.

0

3.5 Gamma-Rag Energy Deposition

A detailed re-appraisal of the TLD technique used to measure gamma-ray energy deposition in the MOZART and BIZET programmes has been performed in support of the collaborative MASURCA RACINE programme. The experimental evaluation and the cal- culational analysis are nearing completion and comparison will be available later this year. An overall improvement of con- fidence in the measurements has been provided by the studies summarised below.

( a ) E lec t ron Migrat ion

In t h e prev ious ZEBRA work t h e PROCEED Monte Ca r lo Code was used t o r e l a t e t h e energy absorbed i n t h e TLD t o t h a t i n t h e surrounding m a t e r i a l , assuming a homogeneous medium around t h e TLD and an i s o t r o p i c photon f l u x . For t h e l m m x l m m x 6mm 7LiF TLD used, surrounded by i r o n o r zirconium, t h e c o r r e c t i o n is about 1 0 % i n c o r e r e g i o n s , i n c r e a s i n g t o about 15% i n a r eg ion where t h e r e is a g r e a t e r low energy photon component. These c o r r e c t i o n s a r e comparable wi th t h e t a r g e t a c c u r a c i e s of t h e des ign c a l c u l a t i o n s f o r bo th thermal and f a s t power r e a c t o r s . To r e p r e s e n t t h e more complicated geomet r ies achieved i n p r a c t i c e and t o t a k e account of t h e photon source d i s - t r i b u t i o n s , an e l e c t r o n t r a n s p o r t module has been developed f o r t h e s t anda rd gamma-heating Monte Ca r lo Code, McBEND. Va l ida t ion of t h i s module is n e a r i n g complet ion.

( b ) Absolute Photon C a l i b r a t i o n

0 TLD do not p rov ide an a b s o l u t e measure of energy depos i t i on and must be c a l i b r a t e d a g a i n s t a primary dosemeter: T O - r e l a t e t h e q u a n t i t y measured by t h e primary dosemeter, t o t h e energy absorbed by t h e TLD r e q u i r e s a c a l c u l a t e d f a c t o r which c o n s i s t s of a component due t o photon a t t e n u a t i o n and a component due t o e l e c t r o n mig ra t ion . The magnitude and u n c e r t a i n t y of t h e f a c t o r is minimised by p o s i t i o n i n g t h e TLD wi th in a medium immediately a f t e r i ts e l e c t r o n bui ld-up r eg ion . A s e r i e s of exper iments has been performed t o t e s t t h e q u a l i t y of c a l i b r a t i o n procedures and i n p a r t i c u l a r t h e p r e d i c t i o n of t h e c a l i b r a t i o n f a c t o r . The a b s o l u t e energy f l u e n c e from a range of photon s p e c t r a ob ta ined a t t h e UK Nat iona l Phys i ca l Laboratory (Co60, 1 MeV and 2 MeV x-ray s p e c t r a ) , I nc iden t on v a r i o u s media (LiF , Perspex, F e ) , provided t h e i npu t source t o t h e c a l c u l a t i o n s . For t h e f i v e combinations of s p e c t r a and build-up media chosen, t h e range of t h e v a r i a t i o n i n t h e c a v i t y c o r r e c t i o n ( i e . e l e c t r o n migra t ion e f f e c t ) was about 20%. The mean c o r r e c t e d a b s o l u t e energy depos i t i on p e r u n i t TLD read-out l i gh t - coun t

0 had a r o o t mean square d e v i a t i o n of 1 .5%.

A secondary s t anda rd i o n i s a t i o n chamber, a l s o c a l i b r a t e d a t NPL, was compared wi th chambers of t h e RACINE p a r t n e r s a t Cadarache, l m from Co60 sou rce . The v a l u e s ob ta ined were i n

v good agreement, ie . UK chamber 2.92 Rlhr If: 1 .6%; French chamber 2.94 R lh r ; Belgian chamber 2.87 R lh r , p rov id ing a good common

base f o r comparing t h e r e a c t o r measurements. *

( c ) Neutron Con t r ibu t ion t o t h e TLD S igna l

Two major i n v e s t i g a t i o n s of t h e f a s t neu t ron response of 7LiF TLD have been c a r r i e d ou t i n t h e USA and Japan. A d a t a s e t p repared from t h e s e r e s u l t s h a s been used i n a l l p r ev ious ZEBRA work and t y p i c a l l y p r e d i c t s a 20% neutron c o n t r i b u t i o n t o t h e TLD s i g n a l . However, r e c e n t doubts about t h i s work prompted a s e r i e s of exper iments t o check t h e f a s t neu t ron response. Neutrons were genera ted a t n i n e energy p o i n t s i n t h e range 200 keV t o 6 MeV u s i n g t h e Dynamitron machine a t t h e Birmingham Radia t ion Centre . The measured response f a c t o r s (photon e q u i v a l e n t energy d e p o s i t i o n p e r u n i t neu t ron f l u e n c e ) were, i n f a c t , i n good g e n e r a l agreement w i th t h e prev ious two s t u d i e s and confirmed t h e s u i t a b i l i t y of t h e e x i s t i n g response d a t a set.

This work also led to the investigation of a second technique for establishing the neutron component, based on the high temperature read-out count. The crystals are normally readout using a fast ramp and plateaux heating cycle of 16s at 135'C to empty the shallow traps followed by 32s at 240•‹C, which provides the main integrated counts normally used in practical dosimetry, and finally 16s at 300•‹C to empty the residual deeper traps. The ratio of the light count from the 300•‹C plateaux to the 240•‹C plateaux for TLD-700 irradiated solely by Co60 gamma-rays is 0.05. The neutron calibrations showed this ratio ranges from about 0.3 at low energies to 0.4 at the upper calibration energies. Advantage can therefore be taken of this . change in ratio to deduce directly the gamma and neutron compon- ents in a mixed-field irradiation. For the RACINE measurements comparison of the two relatively independent methods for determining the gamma component showeh 70% of the results agreed to within 3% and 90% to within 6%.

4. CRITICALITY WORK - 0

The DIhlPLE water-moderated zero-power reactor at Winfrith has been restored and recommissioned. A simple 3% enriched U02 pin lattice, identical to an earlier DIMPLE core is presently loaded. Current experimental techniques are being tested and the results will be compared with those in the first version. T h i s assembly will be followed by critical and subcritical arrangements of the 3% U02 pins in the prototype BNFL CAGR storage and transport skip which has been specially acquired.

Further validation of the X3SK6 Monte-Carlo code and data has been carried out. Substantial progress has been made towards the utlimate aim of producing a code to predict keff to within 1% for all systems. Updating the uranium and plutonium data in the UKNDL is expected to bring that aim within sight. The results are described in a separate paper which also notes the value of inter-Code comparisons, both xithin the UKAEA and through the recent inter- national CSSI coinparison exercise, and suggests some areas where goo quality benchmark criticality experiments are still required. a 5. - SHIELDING STUDIES The programme of data-testing benchmarks in the ASPIS shielding facility of NESTCR is continuing. Detailed studies have been carried out in graphite, iron and sodium, the last of these being irradiated with both the standard fission source plate and a simulated natural-uranium oxide breeder. Winfrith has collaborated with the ESIS Shielding Group at Ispra who have made complementary measurements in iron and sodium in their EURACOS I1 fission-plate facility operated at the University of Pavia.

The Winfrith programme is now turning to the materials used in transport flasks: in addition to measurements in water which have utilised a Californum-252 source to reduce the neutron background for spectrometer operation, benchmark experiments are planned in lead, concrete, polythene, boro-silicone and Jabroc. These measurements will be analysed in collaboration with BNFL using the McBEND-DUCKPOND sensitivity Monte Carlo code. The results of these programmes, which will include adjustments of the data, will be made available to the JEF evaluators for the benchmark testing of the new files. 9 I100751

6. FORTRAN-77

Fortran-77 comwilers are now available on all UK AEA mainframe computers and are in use. The CEGB have recently acquired such a compiler on their IBM machine in London. Many staff throughout the Authority have been on Fortran-77 conversion courses and are using Fortran-77.

The use of non-standard features has been considered at Winfrith and Risley. It has been agreed that in some circumstances, the additional facility obtained by the use of non-standard features outweighs possible reduced portability, particularly in large programs. The main area in which conformance to the standard would restrict programmers is in the simulation of dynamic run-time storage allocation.

It is intended to acquire a Fortran-77 Verifier specifically designed to flag the use of some non Fortran-77 features. However it is unlikely that the Verifier will be able to check interfaces between called and calling routines or to check violations which can only be detected durjng actual execution of a program.

REFERENCES

HALSALL M J. The Treatment of Burnable Poison Pins in LWR-WIMS. AEEW-M1999.

BUTLAND A T D, SIMMONS W N, STEVENSON J M. An Assessment of Methods of Calculating Sodium-Voiding Reactivity in Plutonium- Fuelled Fast Reactors. Proceeding of a Symposium on Fast Reactor Physics, Aix-en-Provence, September 1979. Volume 1, p281.

MADDISON D W, INGRAM G. ANLIAEEW Comparison of Reaction-Rate Ratio Techniques in ZEBRA. NEACRP-A-542.

BOHME R, BURBIDGE B L H. A Comparison of Central Reaction-Rate Ratio Measurement Techniques in BIZET Cores. NEACRP-A-543.

ROWLANDS J L, GRIMSTONE M J.. Fast Reactor Control Rod Calculation Methods and Validation Experiments. NEACRP-A-514.

. , UNITED STATES - 150 -

. . Reac to r Phys ics A c t i v i t i e s i n t h e U n i t e d S t a t e s A Repor t t o t h e NEACRP

October 17-21 , 1983 Po B. Hemmig and J. W. L e w e l l e n

U. S. Department o f Energy Washington, DC 20545

I n t r o d u c t i o n .. Reac to r p h y s i c s a c t i v i t i e s i n t h e U.S. have p r o v i d e d s u p p o r t necessary f o r t h e o p e r a t i o n o f FFTF and l i c e n s i n g r e v i e w s f o r CRBR. O p e r a t i n g measurements i n 4

FFTF have c o n f i r m e d t h e adequacy o f t h e methodology used f o r FFTF c o r e and s h i e l d i n g des ign. Benchmark ana lyses o f FFTF o p e r a t i n g and burnup p h y s i c s a r e c o n t i n u i n g u s i n g improved d a t a and methods.

The U.S. c r i t i c a l exper imen t program now uses o n l y t h e ZPPR f a c i l i t y a t ANL-Id . U s i n g ZPPR, a s e r i e s o f l a r g e heterogeneous co res a r e b e i n g s t u d i e d as p a r t o f t h e J u p i t e r I 1 c o o p e r a t i v e program w i t h PNC, Japan.

0 Assessments o f t h e s t a t e o f t h e a r t i n LMFBR p h y s i c s have c o n t i n u e d a t ANL. Ana lyses o f t h e sodium v o i d r e a c t i v i t y wor ths were comple ted i n s u p p o r t o f CRBR l i c e n s i n g . M a j o r assessments o f Dopp le r c o e f f i c i e n t and power d i s t r i b u - t i o n p r e d i c t i o n s a r e i n p rog ress .

Reac to r d e s i g n s t u d i e s a r e c o n t i n u i n g a t i n d u s t r i a l and n a t i o n a l l a b o r a t o r y o r g a n i z a t i o n s t o e v a l u a t e c o r e d e s i g n a l t e r n a t i v e s o f f e r i n g improved economics, r e l i a b i l i t y and l i c e n s a b i l i t y .

C r i t i c a l Exper iments

The ZPPR-13 s e r i e s o f t h e J u p i t e r I 1 program began i n June 1982 as a c o o p e r a t i v e program w i t h PNC, Japan. T h i s program was des igned t o p r o v i d e a s y s t e m a t i c s t u d y o f c o r e n e u t r o n i c s f o r t h e d e s i g n and a n a l y s i s o f l a r g e LMFBR heterogeneous cores. The ZPPR-13 s e r i e s o f benchmark c o n f i g u r a t i o n s i s shown i n F i g u r e s 1 and 2. 0 - The f i r s t co re , d e s i g n a t e d 13A, c o n t a i n e d a c e n t r a l b l a n k e t r e g i o n and t h r e e a l t e r n a t i n g a n n u l a r r i n g s o f f u e l and b l a n k e t r e g i o n s . The c o r e volume was abou t 4000 l i t e r s and t h e c r i t i c a l mass was about 2500 Kgs o f p l u t o n i u m .

The i n i t i a l measurements were o f excess r e a c t i v i t y , gap w o r t h , tempera tu re c o e f f i c i e n t , s a f e t y r o d wor th , sh im r o d wor th , source t e r m and t h e r e l a t i v e abundance o f t h e de layed n e u t r o n groups. Next , an e x t e n s i v e s e t o f U-235, Pu-239 and U-238 f o i l s was i r r a d i a t e d t o s t u d y t h e s p a t i a l d i s t r ' i b u t i o n s o f t h e U-235 ( n , f ) , Pu-239 (n , f ) , U-238 ( n , f ) and U-238 (n,y) r e a c t i o n s . O i s t r i - b u t i o n s were measured r a d i a l l y a l o n g t h e x and y axes and a t 30•‹ and 60' t o t h e axes. A x i a l d i s t r i b u t i o n s were a l s o measured w i t h i n drawers i n f u e l r i n g s 1 and 3. Smal l sample w o r t h measurements were made f o r Pu-239, U-235, U-238, 6-10, g r a p h i t e , i r o n , s t a i n l e s s s t e e l and Cf-252. F o l l o w i n g t h e smal l -sample w o r t h measurements, t h e r e a c t o r was made 0.10$ s u b c r i t i c a l and a s e t o f c o n t r o l r o d w o r t h measurements was made.

The next two cores, 13B-1 and B-2. were segmented blanket r ing configurations. The f i b s t measurements i n ZPPR 138-1 were operational measurements of the type carried out on the 13A core. Wi th the reactor i n i ts reference c r i t i c a l configurations, dis t r ibut ions of neutron reaction ra tes and gamma doses were measured using f o i l s and thermoluminescent detectors respectively. W i t h the reactor i n a 0.09$ subcr i t ical configuration, worths were measured f o r several control rod patterns. These included two-by-two drawer control rods, groups of rods i n single rings, and interactions between groups of rods i n d i f fe ren t rings. Further control worth data were obtained for rods (two-by-three drawers i n s ize) located i n the gaps of the blanket rings. Measurements i n t h i s case included worths of individual rods, individual sodium-filled rod locations. groups of rods and groups of sodium-filled rod locations. In t o t a l , t h i r t y separate control rod worth measurements were made in the ZPPR 138-1 configuration.

Configuration 138-2 provided a reference t rans i t ion to hex geometry. The conver- sion from 13B-2 to the blanket island configuration of 138-3 was accomplished by a s e r i e s of a l ternat ing negative reac t iv i ty steps ( subs t i tu t ing blanket drawers f o r fuel drawers) and positive reac t iv i ty steps ( subs t i tu t ing fuel drawers f o r blanket drawers). Since flux d is t r ibu t ions were measured a f t e r each s tep , the conversion served as blanket thickening and thinning experiments.

A s e r i e s of p i n and plate control rod worths were measured i n the ZPPR 138-4 engineering benchmark core. This configuration i s s imilar t o 13B-3 w i t h the addition of 30 mockup control rod positions. Worths were obtained for an extensive s e t of rod groups with par t icular emphasis on the interact ion between groups and the e f fec t s of a s ingle withdrawn rod. The properties of loosely coupled cores are now being studied in core 13C. These include the decoupling of the kinet ic response and the spa t ia l s ens i t i v i t y t o compositional perturba- t ions.

In the ZPPR-13 ser ies t o date, the experimental reaction ra tes were generally well calculated. The C / E radial bias of 2-6% i n f i s s ion r a t e s agrees closely w i t h the radial bias i n control rod worths. A small (1-3%) asymmetry was observed i n azimuthal reaction ra tes . This was traced to a small asymmetry i n blanket ce l l loadings and to fuel gap uncertainties (which were on the order of 1 mm) on closure of the s p l i t table assembly. The control rod position worths were not well calculated by diffusion theory. The observed variations of 10-20% indicate the need f o r a be t t e r treatment of neutron streaming i n the sodium channels.

. Computational Physics Methods

Development of an integrated design capabi l i ty adequate for optimization and . detai led design of LMFBR plants up to the l a rges t practical ratings has progressed s ignif icant ly . Much of 1983 e f f o r t was directed t o the needs for e f f i c i e n t energy-dependent solutions w i t h i n the range of the following variations:

o Reactor s i z e Large s izes are most demanding and a re receiving the most a t ten t ion .

o Computer hardware Most implementation is on CDC, IBM, and more advanced CRAY and CYBER 205 equipment.

. . o Rigor - d i f fus ion and higher order t r a n s p o r t approximations.

o Methodologies - f i n i t e d i f f e rence and advanced nodal.

o Mode - forward and a d j o i n t .

o Mesh geometry - R Z , XY(Z), H E X ( Z ) , e t c .

0 P a r t i c l e - neutron, gamma, neutron-gamma

REBUS-3 was developed a t ANL t o perform burnup c a l c u l a t i o n s f o r f a s t r eac to r design. The code now runs success fu l ly on t h e CRAY-1. There a r e d i f fus ion theory opt ions based on f i n i t e d i f f e rence , nodal, and f lux synthes is methods, and mu1 t i p l e - c y c l e , non-equilibrium c a l c u l a t i o n s with fuel enrichment i n s e l e c t e d regions a t each cycle a r e permit ted. The code has genera1 fuel s h u f f l i n g c a p a b i l i t y .

A new a n a l y t i c a l model was developed t o p red ic t t h e r e a c t i v i t y e f f e c t s o f i r rad ia t ion- induced subassembly bowing i n an LMFBK. T h i s model i s t h e f i r s t which can r ep resen t heterogeneous core geometries i n t h e necessary d e t a i l . A computer code based on t h i s work has been w r i t t e n and l inked with output f i l e s o f neut ronics , dep le t ion , and thermal hydraul ics codes.

A version of t h e TWODANT t r a n s p o r t code has been developed by LANL f o r t r i a l

!safe;y, This t r a n s p o r t code so lves eigenvalue and inhomogeneous source problems rn x and ( r , z ) geometries. I t employs t h e d i f f u s i o n s y n t h e t i c acce le ra t ion method with a mul t igr id scheme f o r so lv ing t h e d i f fus ion acce le ra t ion equat ions. Arbi t ra ry o rde r -o f - sca t t e r ing i s permit ted.

A companion t o TWODANT f o r ana lys i s o f hexagonal geometries i s TNOHEX. I t uses an e q u i l a t e r a l t r i a n g u l a r mesh with s p a t i a l d i f f e renc ing based on t h e l i n e a r c h a r a c t e r i s t i c method. The l a t t e r is s i g n i f i c a n t l y more accura te than diamond- 1 i ke d i f f e renc ing f o r t r i a n g l e s and r equ i re s no negat ive f l u x f ix-up. TWOHEX uses 60•‹ r o t a t i o n a l l y - i n v a r i a n t quadrature t o ensure 60' symmetries and c u r r e n t l y employs Chebyshev acce le ra t ion on both inner and o u t e r i t e r a t i o n s . Performance o f t h i s code i s undergoing i n i t i a l user t e s t s a t Los Alamos.

In o t h e r work, an advanced deple t ion module cons i s t ing o f t h e LANL CINDER code and t h e ANL DIF3D d i f fus ion code i s being developed. CINDER performs d e t a i l e d summation ca l cu la t ions f o r 877 nucl ides i n 40 a c t i n i d e and 102 f i s s i o n product chains with data f o r 31 ENDFIB-V y i e l d s e t s . The module wi l l be optimized f o r t h e CRAY computer.

Nuclear Data

The e f f o r t s t o improve t h e U.S. nuclear data f i l e s f o r r e a c t o r design r e su l t ed i n issuance o f several updates and changes which were designated ENDFIB-V Mod I 1 These included r ev i s ion o f t h e Fe, Th-232, U-239, L i , W , Ag, Rb, Pu-239, Kr, Ag, E u , Xe, Gd, and Zr data f i l e s . Data t e s t i n g o f t h e Mod I1 data i n d i c a t e small but s i g n i f i c a n t improvewnts over t h e ENDFIB-V data. Development o f t h e

ENDFIB-VI data f i l e i s i n progress. I n i t i a l e f fo r t s include a consistent evaluation of the standards f i l e s w i t h the key reaction cross sections important in reactor design.

ANL s c i e n t i s t s measured cross sections for several s tab le f i s s ion products near the l i g h t mass f iss ion yield maximum (A-85-125). The new resu l t s d i f f e r from previously used data by 50-100% in some cases. The new data will be used i n updates of the ENDFIB f i l e s and to improve nuclear model predictions for experimentally inaccessible radioactive f iss ion products. Measurements of higher act inide cross sections have continued a t ORELA. These include t o t a l , f i s s ion and capture measurements on Am, T h , and Pu-240 as well as the extension of U-238 resonance parameters u p to 6 Kev. Measurements of v(E) were completed for U-233, U-235, Pu-239 and Pu-241.

- Shielding

Studies of the FFTF shielding have been carried out by HEDL and ORNL. Measurements of the radiation f i e ld s i n FFTF were made i n the re f lec tor , the in-vessel storage locations, the reactor cavity, and the head compartment regions. The major shielding measurements were made during an 8 day FFTF r u n a t fu l l power; however, some measurements were made a t low power and some with 36 spent fuel elements stored in-vessel . Analyses to date indicate general 1 y good agreement between the FFTF measurements and calculations based on TSF benchmark experiments. The use of bias factors obtained from benchmark measurements was necessary t o provide good predictions of the FFTF shield performance. The ringed geometry modeling which was used to calculate the spent fuel mu1 t i pl icat ion appears adequate.

ORNL and GA have investigated the neutron streaming through the e x i t cooling passages of the lower core re f lec tor and s t ructural core support region in high temperature gas cooled reactors. Mockup measurements using the Tower Shielding Fac i l i ty indicate substantial streaming e f fec t s which are not well predicted by the design methods currently used. Results to date indicate the need to use measured bias factors and/or improved computational techniques to predict neutron streaming through the lower core region.

Shield design s tudies have been carr ied out on large LMFBR plant concepts by O R N L , G E , A1 and ANL. These s tudies have indicated tha t substantial weight and cost savings can be obtained by the use of a l te rna te shield materials and " appropriately optimized shield configurations.

. FFTF Physics

Act iv i t i es included core characterization measurements, operating physics measurements and benchmark analyses. Papers on the core characterization and capture r a t e measurements will be presented a t the November 1983 ANS meeting.

physics measurements i n cyc le 2 included:

o Individual BOC rod worths by a combination o f inverse k i n e t i c s rod drop and modified source mu1 t ip1 i c a t i o n techniques.

o Burnup r e a c t i v i t y by change i n pos i t ion of c a l i b r a t e d rods.

o S ingle assembly s u b s t i t u t i o n worth.

o Close monitoring of t h e power c o e f f i c i e n t o f r e a c t i v i t y and s t a b i l i t y phase margins.

Calcula t ions o f rod worths, burnup r e a c t i v i t y , and assembly worths agreed q u i t e well with measurements. The measured power c o e f f i c i e n t s and s t a b i l i t y phase margins were well within requirements. Thi r ty- three experiments were continued and nine new ones were loaded a t the beginning of cyc le 2 i n January 1983. A capac i ty f a c t o r o f 83% was achieved over t h e cyc le . a Cycle 3 began on schedule Ju ly 4 , 1983. A key goal f o r this cycle i s t o extend peak p e l l e t exposure; about 75-80 MWd/Kg has been achieved so f a r . Seven new experiments were added a t BOC 3.

Measured FFTF data a r e being used as benchmarks f o r t e s t i n g neutronic codes and da ta . Calcula t ions have been performed f o r cold condit ion c r i t i c a l i t y and cont ro l worths using various models and data s e t s . These wi l l be extended by near-term analyses of burnup data from t h e e a r l y FFTF cyc les .

Core Design and Assessments

Assessments o f the s t a t e - o f - t h e - a r t f o r c a l c u l a t i n g key f a s t r e a c t o r physics parameters a r e continuing a t A N L . Recent assessments address our a b i l i t y t o p red ic t c r i t i c a l i t y , r eac t ion r a t e s , Doppler and sample worths.

Core design s t u d i e s a r e c a r r i e d o u t a t t h e major i n d u s t r i a l con t r ac to r s and nat ional l a b o r a t o r i e s . The major i s sues being addressed a r e design t r adeof f s t o increase inhe ren t s a f e t y and reduce p l an t and fuel cyc le c o s t s .

- 155 -

Figure 1

a BLANKET REFLECTOR

BLANKET REFLECTOR

ZPPR- 13B/3

BLANKET REFLECTOR

CONTROL ROO POSITION

/ " - 176 - F igu re 2

BLANKET [7 REFLECTOR

ZPPR - 13 A

ZPPR - 13 C

- - 1 5 7 - JRC-ISPRA

REACTOR P H Y S I C S A C T I V I T I E S A T THE J R C I S P R A

I. CORE PHYSICS STUDIES

Planning and design of a large scale loss of coolant experiment typical

of LWRs led to a considerable intensification of the reactor physics

activities at the JRC-Ispra. The project under investigation consisted

of a light water test loop (containing 32 fuel pins) immersed in a heavy

water reactor of a high thermal f l u generated by fully enriched uranium

aluminum fuel elements.

The main objectives of the study which includes among its tasks a series

of rather unusual problems are given in the following summary:

- To satisfy the condition of a radiall-, flat power and decay heat

distribution in the test channel an enrichment scheme depending on the

fuel pin position had to be developed. The power distribution analysis

was performed by DOT 3.5 and validated by KENO-IV.

An "a posteriori" comparison with mock-up measurements fully confirmed

the theoretical results. ORIGEN calculations of the decay heat sources

for only 48h of irradiation time gave poor results.

- The axial power distribution of the test channel had to be determined for different control rod positions. This problem was solved by the

use of 3D few group diffusion codes such as SYNTH-C and VENTURE-2.

- All parameters needed for controlling the reactor with the test loop in operation had to be verified and if necessary reanalyzed.

Particular attention had to be paid to the positive void coefficient \

following a voiding of the test channel.

For this task DOT 3.5, KENO-IV and a newly developed Monte Carlo

perturbation scheme based on correlated tracking and a second order

Taylor series approach were used. With these new methods perturbation

effects as small as a few pcm could be reliably estimated (e.g.

temperature coefficient of the coolant).

- Finally the mock-up experiments performed for different test loop

configurations were analyzed to validate the theoretical and calcula-

tional methods in use.

Reports

- "Confronto fra Calcoli di Fisica e Valori Sperimentali per il MOCK-UP di SuperSara in ESSOR"

E. Caglioti, R. Ricchena - TN/1.05.01.83.65 (Agosto 1983)

- "Improved Neutronic Analysis of the SuperSara Experiment by the 3-D Few Group Diffusion Code Synth-C"

E. Salina - Contact EURATOM/ARS N.1667-81-12 ED ISP I

- "Reactor Physics Analysis for the SuperSara Test Project using Monte Carlo Method"

Tayyab Abbas, M. Aglietti-Zanon - Contract TEAM/EURATOM N. 1668-8112

ED ISPI (Feb. 83)

- "Core Physics Analysis for the SuperSara Test Project using Venture-2 Diffusion Theory Code"

M. Aglietti-Zanon, Tayyab Abbas - Contract TEAM/EURATOM - N.

1828-82-03 ED ISP I (May 83)

- "Generalized Monte Carlo Perturbation Algorithms for Correlated

Sampling and a Second Order Taylor Series Approach"

H. Rief - submitted to Annals of Nucl. Energy (Oct. 1983)

11. RADIATION SHIELDING

During the reporting period a deep penetration sodium benchmark

experiment has been carried out.The sodium column in which the meas-

urements were performed was about 400 cm long and consisted of seven

steel boxes of approximately 200 X 150 cm cross section filled with

sodium. The neutron source was a highly enriched U235 converter plate.

a Measurements were performed along the central axis of the sodium

assembly, using activation and threshold detectors for the different

energy regions and proton recoil counters. In Fig. 1 activation rates

for sulphur and gold a r e shown as functions of the source detector

distance.

Interpretation of the iron and sodium benchmarks has continued

simultaneously with the measurements. The first results of group cross

section adjustments were presented to two international Conferences

(Antwerp, Sept. 1982 and Tokyo, May 1983).

In the case of iron they indicate, for example, that the inelastic cross

section of ENDF/B4 is about 5% too high near the threshold.

Publications

- W . Matthes et al.

Adjustment of Neutron Multigroup Cross Sections with Error Covariance

Matrices to Deep Penetration Integral Experiments

Paper presented at the International Conference on "Nuclear Data for

Science and Technology" - Antwerp, Sept. 1982

- "A Parametric Representation of Gamma Ray Attenuation in Two-Layer Shields"

H. Penkuhn (EURATOM-Ispra), Schultz (U. Hannover, Germany) - 6th ICRs, Tokyo (1983)

- "Monte Carlo Shielding Analysis Using Deep Penetration Biasing Schemes Combined with Point Estimators and Algorithms for the Scoring of

Sensitivity Profiles and Finite Perturbation Effects"

H. Rief (EURATOM-Ispra), A. Fioretti (A.M.N., Genova-Italy)

6th ICRs, Tokyo (1983)

- "Adjustment of Neutron Multigroup Cross-Sections to Integral Experi- ments"

G.Hehn, R.D.Bachle, G.Pfister, M.Matthes (IRE, U.Stuttgart, Germany) ,

W.Matthes (EURATOM-Ispra) - 6th ICRs, Tokyo (1983)

- "On Unfolding Counting-Rate Spectra of Recoil Proton Neutron De-

tectors", Y. Yeivin (Hebrew Univ., Jerusalem; visiting scientist at

the JRC-Ispra) - 6th ICRs, Tokyo (1983)

- "MORSE-Z1, Source and Analysis Routines to be used with the MORSE Code"

C. Ponti and R. Van Heusden - EUR 8320 EN (19833

- ZSIS Newsletter 43 (October 1982, H. Penkuhn - "A Shielding Kernel for Fission Gammas - Part I: The Unscattered Flux from a ?oint Source"

- ESIS Newsletter 44 (January 1983)

- ESIS Newsletter 45 (April 1983), H. Penkuhn - "A Shielding Kernel for Fission Gammas - Part 11: The Scattered Energy Flux from a Point

Source

111. ACTINIDE MONITORING

In the field of nuclear safeguards and monitoring of 4-contaminated

waste streams special work has been carried out applying the passive

neutron interrogation technique.

For the existing shift register technique measuring the doublet neutron

signals of spontaneous and induced fission events corrections for

induced fission counts were carried out. Investigations are continued to

obtain from the analysis of multiplets of higher order additional

information on the physical state of the test items.

References

"Neutron Multiplication Corrections Applying the shift Register

Technique", W. Hage, L. Anselmi, K. Caruso - ESARDA, Proceedings of the

5th Annual Symposium on Safeguards and Nuclear Material Management,

(Versailles, France) JRC Ispra Publication (1983)

"Neutron Signal Multiplet Analysis for the Mass Determination of

Spontaneous Fission Isotopes", R. Dierckx, W. Hage - Nuclear Science and Eng., to be published