The revision work of JENDL-3 has been made by considering feedback information of various benchmark tests. The main revised quantities are the resonance parameters, capture and inelastic scattering cross sections, and fission spectra of main actinide nuclides, the total and inelastic scattering cross sections of structural materials, the resonance parameters the capture and inelastic scattering cross sections of fission products, and the γ-ray production data. The revised data were released as JENDL-3.2 in June 1994. The preliminary benchmark tests indicate that JENDL-3.2 predicts various reactor characteristics more successfully than the previous version of JENDL-3.1.

https://www.tandfonline.com/doi/pdf/10.1080/18811248.2002.9715303

Japanese evaluated nuclear data library version 3 revision-3: JENDL-3.3

Energy levels of light nuclei A = 18–19☆

Adopted values for the excitation energy, &(3;), of the first 3- state of even-even nuclides are tabulated. Values of the reduced electric-octupole transition probability, B(E3;O: + 3;), from the ground state to this state, as determined from Coulomb excitation, lifetime measurements, inelastic electron scattering, deformation parameters /3s obtained from angular distributions of inelastically scattered nucleons and light ions, and other miscellaneous procedures are listed in separate tables. Adopted values for B(E3;O: + 3;) are presented in the final table, together with the E3 transition strengths, in Weisskopf units, and the product EA3;) X B(E3;O: --* 3;), expressed as a percentage of the energyweighted E3 sum-rule strength. An evaluation is made of the reliability of B(E3;O: + 31) values deduced from deformation parameters &. The literature has been covered to March 1988.

/pdf/for-even-even-nuclides-throughout-the-periodic-table-tacjid065q.pdf

For even-even nuclides throughout the periodic table

Lattice Physics Computations

Absolute neutron capture cross sections for natural elements of copper, yttrium, zirconium, niobium, lanthanum, gadolinium, terbium, tantalum, tungsten, rhenium, platinum, thallium, bismuth, and separated isotopes of /sup 63/Cu, /sup 65/Cu, /sup 155/Gd, /sup 157/Gd, /sup 158/Gd, /sup 160/Gd, /sup 182/W, /sup 183/W, /sup 184/W, /sup 186/W, /sup 203/Tl, /sup 205/Tl have been measured in the 0.5- to 3.0-MeV energy range. For most of these nuclides and isotopes, available data were scarce and discrepant, especially for neutron energies above 0.7 MeV. A spectrum-fitting method was developed to deduce the radiative capture cross section from prompt gamma rays emitted by the sample. The gamma rays were recorded by a NaI scintillator surrounded by an annular detector and the capture gamma-ray spectrum was obtained by unfolding the observed pulse-height distribution with the response function of the detector. Gamma ray spectra emitted in the capture of 0.5-MeV neutrons as well as the multiplicity of the gamma-ray transmissions are presented.

Capture Cross Sections and Gamma-Ray Spectra from the Interaction of 0.5- to 3.0-MeV Neutrons with Nuclei in the Mass Range A = 45 to 238

Neutron nuclear data in the energy range between 10−5 eV and 20MeV have been evaluated for 172 nuclides from 75As to 169Tb in the fission product mass region to provide data for the JENDL-3 fission product nuclear data library. Evaluation was made on the basis of recent experimental data reported up to 1988 and the nuclear model calculations. Resonance parameters have been evaluated on the basis of measured data set and a REPSTOR system developed in JAERI. The spherical optical model and statistical theory were applied to calculation of the total, capture, elastic and inelastic scattering cross sections, and the multistep evaporation model and pre-equilibrium theory were used for threshold reaction cross section calculations. For the even-even nuclides around fission yield peaks, direct inelastic scattering cross sections were calculated with the distorted wave Born approximation. Nuclear model parameters, such as optical model parameters, level density parameters, x-ray strength functions and Kalbach con...

https://www.tandfonline.com/doi/pdf/10.1080/18811248.1992.9731515

JENDL-3 Fission Product Nuclear Data Library

Results of evaluation of neutron cross sections are presented for 27 fission product nuclides selected as being most important for fast reactor calculation The cross sections considered are total, elastic scattering, inelastic scattering and capture cross sections in the energy range from thermal to 15 MeV Thermal and resonance cross sections were calculated from resonance parameters The calculated thermal capture cross section was adjusted by the measured value by adding a background cross section of 1/v form A modified multi-level Breit-Wigner formula was developed to avoid the well-known occurrence of negative values in elastic scattering cross section Smooth cross sections above resonance region were calculated with the spherical optical model and the statistical theory, taking account of neutron width fluctuation and level interference The calculation was adjusted by capture data when available The joining between the resonance and smooth cross sections was performed with the aid of statistica

Evaluation of Neutron Cross Section of 27 Fission Product Nuclides Important for Fast Reactor

Uncertainties of decay heat summation calculations are derived for the thermal fission of 235U and 239Pu and the fast fission of 238U through sensitivity analyses using data given in the JNDC FP nuclear data library. The uncertainties analyzed are those relevant to decay energies, fission yields and decay constants among the nuclear data contained in the summation calculation. For nuclides lacking complete experimental data, the uncertainties of the decay energies are theoretically calculated. Thus analyzed, the maximum uncertainties of burst fission are 2.8 % for 235U, 3.2%” for 239Pu and 4.2% for 238U in the range of cooling time between 1 and 109s, and in the case of infinite irradiation, the corresponding level of maximum uncertainty is below 1.6% for all three fissioning nuclides.

https://www.tandfonline.com/doi/pdf/10.1080/18811248.1992.9731491

Analysis of Uncertainties in Summation Calculations of Decay Heat Using JNDC FP Nuclear Data Library

Numerical Study on Applicability of Benoist's Diffusion Coefficient to Sodium Void Reactivity Analysis

A benchmark calculation of full fission product was performed for thermal reactor application using an isotope transmutation code DCHAIN based on 185 nuclides with revised nuclear data library. The fission product model for BWR lattice calculation was studied and tested with the benchmark results, and a model containing 45 explicit nuclides and one pseudo nuclide was selected as a reasonably best model to predict the burn up reactivity with high precision for practically all types of fuel and reactor operating conditions. The evaluated thermal cross section and resonance integral for the pseudo nuclide are σ2,200 = 2.6b and.RI = 10.6b, combined with the pseudo fission yield values of 1.3898, 1.3233, 1.3675 and 1.2773 for fissions from 235U, 238U, 239Pu and 241Pu, respectively. The present results are believed as equally applicable to PWR lattice calculation.

https://www.tandfonline.com/doi/pdf/10.1080/18811248.1982.9734122

Shungo Iijima

Papers

JENDL-3 Fission Product Nuclear Data Library

Evaluation of Neutron Cross Section of 27 Fission Product Nuclides Important for Fast Reactor

Analysis of Uncertainties in Summation Calculations of Decay Heat Using JNDC FP Nuclear Data Library

Numerical Study on Applicability of Benoist's Diffusion Coefficient to Sodium Void Reactivity Analysis

Fission product model for BWR lattice calculation code.