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Reactor Physics Activities in Nea Member Countries 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 IN NEA MEMBER COUNTRIES October 1982 - September 1983 OECD NUCLEAR ENERGY AGENCY 38 Boulevard Suchet, 75016 Paris 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. The method of Iida and Seki (1), 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, no single self-consistent set of results for a comprehensive rangebecau"e. 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 , but uses calculated instead of measured parameters. This has the(2' 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 of 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 detectors have been exposed in the first stage to spent TRIGA fuel elements and in a second stage to spent fuel elements from the Halden boiling water reactor. It could be shown that TLDs and SSTDs offer some considerable benefits compared to ionization and fission chambers such as easy trans- portation and handling, simple application, parallel exposure to many fuel assemblies at the same time and low costs. In no way these two detector types may replace chambers but they can be used as additional back-up methods in case that routine measurements show deviations from the expected burn-up value. 2.2 Reactivity Measurement Measurements of the subcritical reactivity have been carried out at the Siemens-Argonaut reactor of TU Graz using the inverse kinetics method and the source jerk method respectively. The results gained with these two methods agree satisfactorily. REACTOR PHYSICS ACTIVITIES IN BELGIUM
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