Hiroyuki Matsunobu

JENDL-4.0 : A New Library for Nuclear Science and Engineering

The fourth version of the Japanese Evaluated Nuclear Data Library has been produced in cooperation with the Japanese Nuclear Data Committee. In the new library, much emphasis is placed on the improvements of fission product and minor actinoid data. Two nuclear model codes were developed in order to evaluate the cross sections of fission products and minor actinoids. Coupled-channel optical model parameters, which can be applied to wide mass and energy regions, were obtained for nuclear model calculations. Thermal cross sections of actinoids were carefully examined by considering experimental data or by the systematics of neighboring nuclei. Most of the fission cross sections were derived from experimental data. A simultaneous evaluation was performed for the fission cross sections of important uranium and plutonium isotopes above 10 keV. New evaluations were performed for the thirty fissionproduct nuclides that had not been contained in the previous library JENDL-3.3. The data for light elements and structural materials were partly reevaluated. Moreover, covariances were estimated mainly for actinoids. The new library was released as JENDL-4.0, and the data can be retrieved from the Web site of the JAEA Nuclear Data Center.

JENDL-3.2 Covariance File

Covariances of neutron nuclear data have been estimated for 16 nuclides contained in JENDL-3.2. The physical quantities for which covariances are deduced are cross sections, resolved and unresolved resonance parameters, the first order Legendre-polynomial coefficients for the angular distribution of elastically scattered neutrons, and fission neutron spectra. As for 233,235,238U and 239,241Pu, covariances were obtained also for the average number of neutrons emitted in fission. Least-squares methods were applied to obtain the covariances of those cross sections which were based on experimental data. A simultaneous evaluation method yielded uncertainties in the fission cross sections of 235U, 238U, 239Pu, 240Pu and 241Pu. Covariances of nuclear model calculations were determined from uncertainties in model parameters. The covariance file thus obtained is processed by a system which has been developed, and used for the adjustment of group cross sections.

Simultaneous Evaluation of Fission Cross Sections of Uranium and Plutonium Isotopes for JENDL-3.3

A simultaneous evaluation of the fission cross sections of 233U, 235U, 238U, 239Pu, 240Pu and 241Pu was carried out for the evaluated nuclear data library JENDL-3.3. A least-squares method was applied to selected absolute and relative measurements on the fission cross sections. Covariance matrices of the experimental data were constructed from the uncertainty information reported in the references. The fission cross sections obtained were compared with the JENDL-3.2 and ENDF/B-VI evaluations. It was found from the comparison that the present results are not so different from those in JENDL-3.2, except for the fission cross sections of 233U and the cross sections above 15MeV, and give smaller X2 value than the JENDL-3.2 cross sections. The averaged fission cross sections of 233U, 238U, and 239Pu relative to that of 235U were calculated for a neutron spectrum produced by 9Be (d,xn) reaction. It was confirmed that the calculated cross-section ratios are in good agreement with the experimental data.

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.

Evaluation of the Nuclear Data on (α, n) Reaction for F, Na, Al, Cr, Fe, Ni, and Cu

Evaluation of the nuclear data on (α, n) reaction, which are very important in analyzing radiation shielding and criticality safety relating to storage, transport, and handling of spent fuel was carried out for 18 nuclides, and the results were compared with the experimental data for (α, n) reaction cross section and thick target neutron yield.

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.

New Evaluations of Heavy Nuclide Data for JENDL-3.3

New evaluations of nuclear data for Uranium, Plutonium, and Thorium isotopes which are essential for applications to nuclear technology were carried out for the Japanese Evaluated Nuclear Data Library, JENDL-3.3. The objectives of the current release of JENDL were to fix several problems which have been reported for the previous version, to improve the accuracy of the data, and to evaluate covariances for the important nuclides. Quantities in JENDL-3.2 were extensively re-evaluated or replaced by more reliable values. The heavy nuclide data in JENDL-3.3 were validated with several benchmark tests, and it was reported that the current release gave a good prediction of criticalities.

Re-evaluation of nuclear data of fission product nuclides for JENDL-3.2

Re-evaluation was performed for the neutron-induced reaction data of fission product nuclides compiled in the Japanese Evaluated Nuclear Data Library, JENDL-3.1. The nuclides whose data were re-evaluated are 63 nuclides from 75As to 155Eu. Main quantities revised are the resonance parameters, capture cross sections and inelastic scattering cross sections. As the results of re-evaluation, the thermal neutron capture cross sections, the resonance integrals and the capture cross sections in the keV region were largely improved. The evaluated data cover the neutron energy region from 10∼5 eV to 20MeV. The present results have been adopted to JENDL-3.2.

Interpretation of Integral Test Results of FP Cross Sections in JENDL-3.2 by Analyzing the STEK Experiments

The integral test of the fission product cross sections in JENDL-3.2 was made as an activity of the Subgroup 10 in the Working Party for Evaluation Cooperation in NEANSC by analyzing the sample reactivity worths measured at the STEK cores. Analyses were separately made by a JNDC team at JAERI and K.Dietze at JNC by using the 70-group cross sections. Neutron and adjoint fluxes used in the analysis by JNDC were converted from the 25-group ones which were adjusted to reproduce the reactivity worth for standard samples by Janssen et al. At ECN, while those used for analysis by K.Dietze were calculated with the reactor core design code system of JNC. The results of both analyses were very similar to each other for strongly absorbing nuclides and in good agreement with the experimental values except for nuclides heavier than A=130 for which the reactivity worth was underestimated by about 10%. However, there were some discrepancies for weakly absorbing nuclides: The analysis by K.Dietze gave a better agreement with the experimental values than those by JNDC. In order to interpret the discrepancies among both calculations and the experimental values, the neutron and adjoint fluxes used in the JNDC calculations were compared with those calculated by K.Dietze. Additionally, flux calculations were also made with diffusion and Monte Carlo codes. For weakly absorbing nuclides, a detailed comparison of sample reactivity worths as well as adjoint spectra was made.

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.