Japanese Evaluated Nuclear Data Library (JENDL)

Three versions of Japanese Evaluated Nuclear Data Library (JENDL) are briefly explained. The first version, JENDL-1, was made aiming mainly at application to fast reactors. The second version, JENDL-2, was built as a bigger general purpose evaluated nuclear data library. It stores nuclear data for 181 nuclides. It has, however, some drawbacks particularly for fusion neutronics calculation. To remove these drawbacks of JENDL-2 and further extend its applicability, the third version, JENDL-3, has now been made. JENDL-3 includes photon-production data for some nuclides, in addition to the data contained in JENDL-2.     

Benchmark Experiment on Vanadium with D-T Neutrons and Validation of Evaluated Nuclear Data Libraries by Analysis of the Experiment

A neutronics benchmark experiment on vanadium, which is a low activation fusion reactor material, was conducted by using the D-T neutron source facility of FNS/JAERI. Neutron spectra, dosimetry reaction rates, γ-ray spectra and γ-ray heating rates were measured in a vanadium experimental assembly. Benchmark tests for four evaluated nuclear data files were performed by analyzing the experiment. As a result, the following problems were pointed out in view of accuracy of fusion reactor designs. (1) JENDL-FF and JENDL-3.2: Total cross section should be reexamined especially at ～2keV. (2) ENDF/B-VI: Double differential cross sections for 14 MeV neutrons should be revised because of the isotropic angular distribution for continuum neutron emission. Gamma-ray production cross sections are too small and discrete γ-ray peaks are not represented clearly. (3) EFF-3: Gamma-ray production cross sections are too large.     

Calculation of neutron cross-sections on copper-63 and -65

Toward the development of the next version of Japanese Evaluated Nuclear Data Library (JENDL) general-purpose file, we calculate neutron cross-sections on ^{63, 65}Cu from 50 keV to 20 MeV, which is the incident energy range above the resolved resonance region in JENDL-4.0. A dispersive optical model potential is adopted with a coupled-channel method for interaction between neutron and ^{63, 65}Cu. Direct, pre-equilibrium, and compound processes are taken into account in the calculation. All cross-sections, differential and double-differential cross-sections are consistently calculated with a single set of model parameters. The calculation results reproduce the measured data very well. In addition, disagreement between the calculated and experimental values seen in an integral test for the ⁶³Cu(n, α)⁶⁰Co reaction is improved by using the cross-section data obtained from the present work.     

Neutron energy dependence of delayed neutron yields and its assessments

Incident neutron energy dependence of delayed neutron yields of uranium and plutonium isotopes is investigated. A summation calculation of decay and fission yield data is employed, and the energy dependence of the latter part is considered in a phenomenological way. Our calculation systematically reproduces the energy dependence of delayed neutron yields by introducing an energy dependence of the most probable charge and the odd–even e?ect. The calculated fission yields are assessed by comparison with JENDL/FPY-2011, delayed neutron activities, and decay heats. Although the fission yields in this work are optimized to delayed neutron yields, the calculated decay heats are in good agreement with the experimental data. Comparison of the fission yields calculated in this work and JENDL/FPY-2011 gave an important insight for the evaluation of the next Japanese evaluated nuclear data library (JENDL) .     

Benchmark Experiment for Radio-Activation of Ex-core Structural Materials using NCA and Analysis with Three-Dimensional Transport Codes

In this study, a radio-activation experiment was conducted using stainless steel outside the active fuel region (active core) in the Toshiba Nuclear Critical Assembly (NCA) in order to verify the homogenization method by simulating the NCA experimental reactor system and understand the effects of this method on the analysis accuracy. In order to validate homogenization method, we simulated the system using the continuous energy Monte Carlo code MCNP, which allows heterogeneously modeling, and examined application of the homogenization method used for modeling commercial boiling water reactors (BWRs) with the TORT code. The calculation results of activation rate obtained by using the MCNP code with either heterogeneous or homogeneous models do not affect the calculation result of activation rate outside the active core. As the homogenization method was validated, the calculation of activation rate using the TORT code was performed with the same homogeneous model as in the MCNP calculation. The results of the activation rate calculation using the TORT code gave values 20 to 30% larger than the calculation results obtained via MCNP for ⁵⁵Mn. This is considered to be caused by thermal energy group structure which is treated as one group.     

Evaluation of neutron nuclear data on platinum isotopes

Neutron nuclear data on 10 isotopes of platinum have been evaluated for the next version of Japanese Evaluated Nuclear Data Library general-purpose file in the energy region from 10^?5 eV to 20 MeV. Resolved resonance parameters of naturally occurring isotopes were taken from a compilation work, while unresolved resonance parameters were obtained by fitting to the total and capture cross sections calculated from nuclear models. A statistical model code was applied to evaluate cross sections above the resolved resonance region. Compound, pre-equilibrium and direct-reaction processes were considered for cross-section calculation. Coupled-channel optical model parameters were employed for the interaction between neutrons and nuclei. Giant-dipole and pygmy resonance parameters for E1 γ-ray transition from platinum isotopes were determined so as to reproduce measured γ-ray spectrum. The present results reproduce experimental data very well. The evaluated data are compiled into Evaluated Nuclear Data File formatted data files.     

Self-Shielding Factors for Neutron Capture Reactions of Uranium-238 and Thorium-232 in Energy Range of 1–35 keV

Self-shielding factors for the neutron capture reactions of ²³⁸U and ²³²Th were measured in the resonance energy region of 1–35 keV, using a neutron time-of-flight method with an electron linear accelerator. The self-shielding factors for arbitrary dilution cross sections were obtained from sets of neutron transmission ratios and self-indication ratios measured with several transmission samples of different thicknesses. The maximum experimental errors for ²³⁸U and ²³²Th were about 3 and about 7%, respectively.

The experimental results were compared with calculations based on JENDL-2, JENDL-3 and ENDF/B-IV. For ²³⁸U, an energy dependent structure was observed in the experimental self- shielding factors. The calculations based on JENDL-2 and ENDF/B-IV did not show this structure in the unresolved resonance region and were smaller than the experimental values from 4 to 6 keV. The calculation based on the resolved resonance parameters in JENDL-3 showed better agreement with the experiment from 4 to 6 keV, but discrepancies still remained in other energy ranges.

For ²³²Th, no remarkable discrepancy was observed in the unresolved resonance region, but JENDL-2 and JENDL-3 tended to give smaller values than the experimental self-shielding factors in the resolved resonance region.     

Determination of Gamma-Ray Exposure Rate from Short-Lived Fission Products under Criticality Accident Conditions

For the assessment of γ-ray doses from short-lived fission products (FPs) under criticality accident conditions, γ-ray exposure rates varying with time were experimentally determined in the Transient Experiment Critical Facility (TRACY). The data were obtained by reactivity insertion in the range of 1.50 to 2.93$. It was clarified from the experiments that the contribution of γ-ray from short-lived FPs to total exposure during the experiments was evaluated to be 15 to 17%. Hence, the contribution cannot be neglected for the assessment of γ-ray doses under criticality accident conditions. Computational analyses also indicated that γ-ray exposure rates from short-lived FPs calculated with the Monte Carlo code, MCNP4B, and photon sources based on the latest FP decay data, the JENDL FP Decay Data File 2000, well agreed with the experimental results. The exposure rates were, however, extremely underestimated when the photon sources were obtained by the ORIGEN2 code. The underestimation is due to lack of energy-dependent photon emission data for major short-lived FP nuclides in the photon database attached to the ORIGEN2 code. It was also confirmed that the underestimation arose in 1,000 s or less of time lapse after an initial power burst.     

Neutronics Calculations of HTTR with Several Nuclear Data Libraries

Benchmark calculations for several HTTR core states were performed with four cross-section sets which were generated from JENDL-3.3, JENDL-3.2, ENDF/B-VI.8 and JEFF-3.0 using a continuous energy Monte Carlo code MVP. The core states were a critical approach in which an annular core was formed at room temperature and solid cores at room temperature and at full power operation. Study of k_eff discrepancies caused by difference of the nuclear data libraries and identification of nuclides which have large effects on the k_eff discrepancies were carried out. Comparison of the respective k_eff from calculations and experiments was also carried out. As the results, for each of the HTTR core states, JENDL-3.3 yields a k_eff agreeing with the experiments within 1.5%Δk, JENDL-3.2 yields k_eff agreement within 1.7%Δk, and ENDF/B-VI.8 and JEFF-3.0 yield k_eff agreement within 1.8%Δk. There is little k_eff discrepancy between ENDF/B-VI.8 and JEFF-3.0. The k_eff between JENDL-3.3 and JENDL-3.2 is caused by difference of ²³⁵U data and has temperature dependency. The k_eff discrepancy between JENDL-3.3 and ENDF/B-VI.8 or JEFF-3.0 is mainly caused by difference in graphite data.     

Reconsideration of the Theoretical Supplementation of Decay Data in Fission Product Decay Heat Summation Calculations

In the early stage of the development of the fission product (FP) decay data library for decay-heat summation calculations a serious disagreement between calculations and sample-irradiation measurements was experienced world wide. This problem was essentially circumvented by introduction of the nuclear model calculations. By applying the recent results using the total absorption gamma-ray spectrometer to the decay heat calculations we found that TAGS results seem to be free from the pandemonium problem and, in this respect, it provides a solid basis of the summation calculations without the supplementation of the nuclear model calculations. Among the typical FP nuclides which deserve the future TAGS experiment there are ⁹⁸Nb, ¹⁰⁰Nb, ¹⁰⁵Mo and ¹⁰²Tc and others from the present survey.