JENDL Dosimetry File

Abstract

The JENDL Dosimetry File based on Japanese Evaluated Nuclear Data Library Version 3 (JENDL-3) was compiled, and integral tests of the dosimetry cross section data were performed by the Dosimetry Integral Test Working Group of the Japanese Nuclear Data Committee. Sixty-one reaction cross sections and their covariance data are stored in the JENDL Dosimetry File. The cross sections were mainly taken from JENDL-3 and the covariances from IRDF-85. The data are given in the neutron energy region below 20 MeV in point and group files in the ENDF/B-V format. In order to confirm reliability of data in the JENDL Dosimetry File, several integral tests were carried out: comparison with average cross sections measured in fission neutron and fast reactor spectra, DT, and Li(d,n) neutron fields. As a result, it has been found that the calculated and the measured average cross sections are in general agreement with each other. However, some of the dosimetry cross sections still showed marked discrepancies between the measured and the calculated data. These discrepant cross sections have been reevaluated and the integral tests have been repeated.

Application of a6LiF Small Neutron Detector with an Optical Fiber to Tritium Production Rate Measurement in D-T Neutron Fields

“…20) The charged particle production reactions induced by the D-T neutrons are listed in Table 1. Cross sections of reaction rate calculations were obtained from the following libraries: JENDL Dosimetry File 99, 21) JENDL-3.3, 22) and ENDF/B-VII.0. To obtain the tritium production rate accurately, JENDL Dosimetry File 99 was employed for the calculation of 6 Li(n,)T and 33 Si nat Zn Zn(n,n)Ni, Zn(n,np)Cu, Zn(n,p)Cu, ENDF/B-VII.0 Zn(n,t)Cu, Zn(n,)Ni Application of a 6 LiF Small Neutron Detector with an Optical Fiber to Tritium Production Rate Measurement 7 Li(n,n)T reaction rates because this data file has been evaluated by integral experiments in several fields, including fission neutron fields and D-T neutron fields.…”

Integral Experiment on Beryllium with DD Neutrons for Nuclear Data Benchmarking

“…In the calculation, the experimental assembly, activation foils and Li 2 CO 3 pellets were precisely modeled and their reaction rates were obtained using the track length tally (F4) function of MCNP with JENDL Dosimetry File 99 11) . The source term of the calculation is described in Ref.…”

A new unfolding code combining maximum entropy and maximum likelihood for neutron spectrum measurement

“…Details of the Joyo dosimetry experiments are described elsewhere [8]. The unfolding calculations were carried out using five kinds of reaction rate data, where the input neutron cross section as a response function prepared by the NJOY code from JENDL Dosimetry File 99 [9] is given in 23 energy group numbers. The uncertainty value for the reaction rate input was set to the measurement uncertainty as a diagonal matrix, which consists mainly of calibration uncertainties of the γ -ray detector.…”