[en] The validation of zirconium evaluation for CENDL-3 were reported. Leakage spectra experiment for 61-cm zirconium and critical experiment HEUCOMP-INTER-001 were considered. Calculation results obtained with MCNP from evaluations of CENDL-3, CENDL-2.1, BROND-2, ENDF/B-VI.2, JENDL-3.2 and JENDL-3.3 were compared. Underestimation of discretely inelastic scattering was observed. Slight overestimate for (n, 2n) DDX and continuum inelastic cross section were also found in the results obtained with the natural zirconium evaluation for CENDL-3. (authors)

[en] The validation of n + Cu evaluation for CENDL-3.1 is performed with the shielding benchmark OKTAVIAN and criticality benchmarks from ICSBEP. For shielding benchmark, generally good results are obtained based on the present evaluations, except for some discrepancy caused by inelastic scatting cross sections found in the analysis of the neutron and gamma leakage current. For fast criticality benchmarks, the predictions of the kerf-value results are quite poor. Overestimation the C/E values of the kerf for about 2% have been found in the calculations of benchmark HMF072, HMF073 and PMF013. Sensitivity analysis performed with the case HMF072 shows the bias of k_eff is mainly caused by the total cross section around 0.1-1.3 MeV. After small modifications done to the total cross section, much better prediction of the kerf values has been achieved. (authors)

[en] Nuclear data covariances are important input data for quantificational assessment of nuclear facility design uncertainty and nuclear data adjustment (NDA), which give direct impact on estimated uncertainties and results of NDA. To test the rationality of the newly evaluated nuclear reaction cross sections covariance in smooth region for ²³⁵U which is generated based on the analysis of source of experiments uncertainties and linear least-square method, the data were tested with the NDA benchmark exercises recommended by the OECD/NEA WPEC/SG33. The input data of the NDA generated from JENDL-4.0 library were updated with the cross sections and covariances of ²³⁵U fission, capture and inelastic scattering reactions from ²³⁵U cov, and used in NDA calculation. The new results were compared with the original JENDL-4.0 ones. The test results show that the covariance data from ²³⁵U cov are unreasonable. Too large uncertainties around threshold energy of inelastic scattering reaction were found. The uncertainties of fission and capture cross sections in smooth region are too small to be supported by NDA results. (author)

[en] The neutron reaction data of ¹³⁹La isotope at neutron energy below 20 MeV were calculated with the distorted wave Bonn approximation (DWBA), as well as the united Hauser-Feshbach and the angular momentum and parity dependent exciton model, and the existing optical model potential parameters were used. The data include cross sections, elastic and inelastic scattering angular distributions, and energy spectra and double-differential cross sections for neutron and charged particles. Moreover, resonance parameters were also added in the final data file, and unified adjustments were done to keep the consistency among different reactions. Evaluated data were compared with experimental data and existing evaluations. Finally a new data file was given in the ENDF-6 format. (authors)

[en] In order to verify the reliability of the multiple-temperature continuous energy cross section library CENACE-1.0 when used for calculating nuclear heating in reactor core, NJOY99/HEATR module and auxiliary code chkACEheat developed locally were used to perform energy-balance check for all materials in the library. The test results show that the pass rate of KERMA factors and heat production cross sections of the CENACE-1.0 library is better than that of the other ACE libraries used as comparison. However, unreasonable KERMA factors still exist in various evaluation libraries, and methods to directly revise the calculation results of KERMA factors need to be developed. (authors)

[en] On the base of the generalized perturbation theory and SN method, the theoretical model for calculating the sensitivity and uncertainty of reaction rates ratio (RRR) versus nuclear data is developed. Taking Jezebel-Pu239 critical benchmarking facility as an example, the sensitivity of F49/F25 ratio versus important nuclear data is calculated, as well as the uncertainty arising from the nuclear data. Basing on the sensitivity and uncertainty analysis as well as the covariance information of input evaluated data, refined evaluation of the inelastic data of Pu-239 is considered as a feasible measure to reduce the uncertainty of reaction rates ratios. (authors)

[en] To meet the demands of the development of the AMPX formatted multi-group neutron-photon coupled constant library for the next generation nuclear power system, thorium molten salt reactor (TMSR), based on the evaluated nuclear data library, CENDL-TMSR recommended for the TMSR, and thermal scattering data sub-library and photon-atom interaction data sub-library of ENDF/B-Ⅶ.1, an AMPX formatted 238-neutron and 48-photon coupled multi-group constant library was developed using neutron-photon coupled group constant processing system NPLC-2. The library contains 400 nuclides and 12 thermal scattering materials, with five temperatures ranging from 300 K to 2500 K. The library utilizes the narrow resonance approximation for the resonance self-shielding calculations. For some important actinium series nuclides, the Nordheim method was provided to perform the resonance self-shielding calculations. The preliminary critical and shielding benchmark verifications confirm the reliability of this new library. (authors)

[en] To validate and improve the quality of the complete set of evaluated nuclear reaction data for ²³³U, criticality benchmarks with fast, epithermal and thermal spectra from ICSBEP handbook were selected to test ²³³U evaluations from CENDL-3.1, ENDF/B-Ⅶ.0, JENDL-3.3 and JENDL-4.0. The effective multiplication factors k_eff of selected benchmarks were calculated with the Monte Carlo code MCNP5 and compared with the benchmark values. The results were analyzed with trend against energy spectrum index and sensitivity analysis. In present validation, the underestimation of k_eff for benchmarks with thermal, epithermal or some of fast spectra is the main problem existed in the tested evaluations. From the view of thermal reactors design, the ²³³U evaluation from ENDF/B-Ⅶ.0 shows better performance than other file tested, but still overestimates the contribution of capture reaction in resonance region. (authors)

[en] The under prediction of k_eff depending on ²³⁵U enrichment in low enriched uranium fueled systems, which had been a long-standing puzzle especially for slightly enriched ones, was studied in this report. Benchmark testing was carried out with several evaluated nuclear data files, including the new uranium evaluations from preliminary ENDF/B-VII and CENDL-3.1. Another problem reviewed here was k_eff underestimation vs. temperature increase, which was observed in the sightly enriched system with recent JENDL and ENDF/B uranium evaluations. Through the substitute analysis of nuclear data of ²³⁵U and ²³⁸U, we propose a new evaluation of ²³⁵U data to solve both of the problems. The new evaluation was tested for various uranium fueled systems including low or highly enriched metal and solution benchmarks in the ICSBEP handbook. As a result, it was found that the combination of the new evaluation of ²³⁵U and the ²³⁸U data from the preliminary ENDF/B-VII gives quite good results for most of benchmark problems. (author)

[en] To develop the evaluation methods for some nuclides whose measured data are scant, the systematics formulae of thermal neutron fission cross sections were established based on the fission theory. On the basis of the recommended thermal neutron fission cross sections and the height of the fission barrier, the correlations between thermal neutron fission cross sections and excitation energy of saddle point were studied. The parameters of every nuclide with different Z were obtained by fitting. Then, the systematic behaviors of these parameters with Z were studied, and the parity relations were found. The overall systematic formulae of thermal neutron fission cross sections were established by fitting. The results indicate that there are correlations between thermal neutron fission cross sections and Z. (authors)