Yasuyuki Kikuchi

Japanese Evaluated Nuclear Data Library, Version-3, JENDL-3

The third version of Japanese Evaluated Nuclear Data Library (JENDL-3) has been developed aiming at really general applications such as fission, fusion and shielding calculations. The general purpose file of JENDL-3 contains neutron nuclear data for 324 nuclides in the ENDF-5 format. In the JENDL-3 evaluation, much effort has been devoted to improve reliability of high-energy data for fusion application, which were not satisfactory in JENDL-2, and to include gamma-ray production data. Some advanced nuclear theoretical models were adopted and recent experimental data of energy-angle double-differential cross sections (DDX) mainly measured in Japan were taken into account. Various benchmark tests have so far been made in order to verify the applicability of JENDL-3 to various fields. For fast reactor calculations, JENDL-3 gives satisfactory results for most of characteristics. Particularly, space dependences of reaction rates, sodium void coefficients and control rod worths, which were significant with JENDL-2, nearly disappear. This suggests that the JENDL-3 data are well balanced. Satisfactory applicability has been also proved for the other field such as thermal reactor calculations, fusion neutronics, shielding and dosimetry.

Second Version of Japanese Evaluated Nuclear Data Library (JENDL-2)

Compilation of the second version of Japanese Evaluated Nuclear Data Library (JENDL-2) has been completed. The aim of JENDL-2 is the applications to fast, thermal and fusion reactors, and shielding problems. The evaluated data for 181 nuclides were stored in the ENDF/B-IV format. Various benchmark tests have been performed, and applicability of JENDL-2 was confirmed for the fast and thermal reactor calculations and for the shielding problems. In this review, evaluation work for JENDL-2 is briefly described, and the results of the benchmark tests are summarized.

Statistical model calculations of the 232U fission cross section

A statistical model is used to interpret the available {sup 232}U(n,f) cross-section data in the neutron energy range from 0.003 to 7.4 MeV. Below an incident neutron energy of {approx}1 MeV, the non-threshold energy dependence of the {sup 232}U(n,f) cross section is interpreted in a double-humped fission barrier model, the inner barrier of fissioning nucleus being {approximately}1 MeV lower than the outer one, as anticipated with the shell correction method calculation.

Simultaneous evaluation of fission and capture cross sections and their covariances for heavy nuclei

Abstract The fission cross sections of 235U, 238U, 239Pu, 240Pu, and 241Pu, and the capture cross sections of 197Au and 238U have been simultaneously evaluated for JENDL-3 in the energy range from 50 keV to 20 MeV. A generalized least squares method has been applied to estimate their cross sections and associated covariances from experimental data of absolute and ratio measurements. The experimental covariances were estimated from the partial errors for each measurement. The partial errors were categorized to a few groups which have specific correlation factors. The correlation only in neutron energies was taken into account. The presently evaluated fission cross sections of 235U and 239Pu are a few percent lower than those of JENDL-2 below 1 MeV.

Present Status and Benchmark Tests of JENDL-2

The second version of Japanese Evaluated Nuclear Data Library (JENDL-2) consists of the evaluated data from 10¯5 eV to 20 MeV for 176 nuclides including 99 fission product nuclei. Complete reevaluation has been made to heavy actinide, fission product and main structural material nuclides. Benchmark tests have been made on JENDL-2 for fast reactor application. Various characteristics in core center have been tested with one-dimensional model for total of 27 assemblies, and more sophisticated problems have been examined for MOZART and ZPPR-3. Furthermore analyses of JUPITER project give useful information. Satisfactory results have been obtained as a whole. However, the spectrum is a little underestimated above a few hundred keV and below a few keV. The positive sodium void reactivity worth is much overestimated. As to the latter, the sensitivity analysis with the generalized perturbation method suggests that the fission cross section of 239Pu below a few keV has an important role.

Evaluation of nuclear data for fusion neutronics

Abstract Evaluation of nuclear data for five light nuclides and three structural-material nuclides has been made for fusion application. For the light nuclides, the valuation was based on the experimental data with the aid of the R-matrix theory. For the structural-material nuclides, the direct and preequilibrium processes were taken into consideration.

Reliability of Actinoid Nuclear Data Differential Data and Integral Tests.

The current status of the neutron nuclear data of major and minor actinoid nuclides is reviewed. As to the major actinoids, most of the long standing problems have been resolved, though some problems still remain in the inelastic scattering, fission spectrum etc. On the contrary, the data status is not satisfactory concerning the minor actinoids, particularly those which play the role of main fissile and fertile materials in actinoid burner reactors. To fulfill such requirement JENDL Actinoid File will be provided, which contains 89 actinoid nuclides from Th to Es. In order to validate the minor actinoid nuclear data, two types of integral tests are described. One is the analysis of post-irradiated fuels, where JENDL-3 data give better results than ORIGEN-2 for Pu and Am production. The reaction rates and central reactivity worths have been measured in FCA IX cores, and problems of JENDL-3 have been pointed out. In order to improve the minor actinoid data, much more efforts will be required both on the experimental and theoretical work.

Evaluation of FP cross sections for JENDL-2

Abstract The evaluation method of the recently completed JENDL-2 fission-product cross sections is presented. As an example of ongoing integral tests, the adjustment of the Ag-109 cross section based on integral data is discussed.

Sensitivity analysts of jendl library using fast reactor physics parameters

Abstract Final results of the benchmark test on JENDL-2 are obtained by correcting a cross section processing error using sensitively coefficients. The investigation of the overestimation of sodium void worth reveals that the sensitivities to the Pu-239 fission cross section and U-238 capture cross section are large below a few keV. Sensitivity analysis of a JENDL-3 preliminary version is also carried out.

Evaluation of Neutron Nuclear Data for Uranium-233

Neutron nuclear data of 233U have been evaluated in the energy range from 10-5 eV to 20 MeV. Evaluated quantities are the total, fission, capture, elastic and inelastic scattering, (n,2n) and (n,3n) reaction cross sections, and the average numbers of prompt and delayed neutrons emitted per fission. The thermal and resonance cross sections have been evaluated on the basis of the measured data. The resolved resonance parameters are given up to 100 eV and the unresolved resonance parameters between 100 eV and 30keV. The total and fission cross sections have been evaluated in the higher energy region on the basis of the recently measured data, while the theoretical calculation with the optical, statistical and evaporation models has been used for evaluation of the other cross sections. The presently adopted optical potential parameters have reproduced well the experimental total cross section in the entire energy range as well as the measured data of the s-wave strength function. The structure observed in the...