Measurement of the U(n, f) cross section relative to the Li(n, t) and B(n, ) standards from thermal to 170 keV neutron energy range at n_TOF

Amaducci, S.*; Harada, Hideo; Kimura, Atsushi; 118 of others*

European Physical Journal A, 55(7), p.120_1 - 120_19, 2019/07

https://doi.org/10.1140/epja/i2019-12802-7

The U(n, f) cross section was measured at n_TOF relative to Li(n, t) and B(n, ), in a wide energy range (25 meV170 keV) with 1.5% systematic uncertainty, making use of a stack of six samples and six silicon detectors placed in the neutron beam. The present results indicate that the cross section in the 918 keV neutron energy range is indeed overestimated by almost 5% in the recently released evaluated data files ENDF/B-VIII.0 and JEFF3.3. Furthermore, these new high-resolution data confirm the existence of resonance-like structures in the keV neutron energy region. From the present data, a value of 249.71.4(stat)0.94(syst) b eV has been extracted for the cross section integral between 7.8 and 11 eV, confirming the value of 247.53 b eV recently established as a standard.

Comparison of U fission cross sections in JENDL-3.3 and ENDF/B-VI

Kawano, Toshihiko*; Carlson, A. D.*; Matsunobu, Hiroyuki*; Nakagawa, Tsuneo; Shibata, Keiichi; Talou, P.*; Young, P. G.*; Chadwick, M. B.*

JAERI-Research 2001-058, 28 Pages, 2002/01

JAERI-Research-2001-058.pdf:1.71MB

no abstracts in English

Experimental analyses results on the BFS 58-1-I1 critical assemblies

; Sato, Wakaei*; Iwai, Takehiko*

JNC TN9400 2000-096, 113 Pages, 2000/06

JNC-TN9400-2000-096.pdf:3.1MB

This report describes the updated analyses results on the BFS-58-1-I1 core. The experiment was conducted at BFS-2 of Russian Institute of Physics & Power Engineering (IPPE). The central region is "non-Uranium fuel zone", where only Pu can induce fission reaction. The non-U zone is surrounded by MOx fuel zone, which is surrounded by U0 fuel zone. Sodium is used for simulating the coolant material. As it was found that the lattice pitch had been incorrectly understood in the past analyses, all items have been re-calculated using the corrected number densities. Furthermore, significantly softened neutron spectrum in the central region caused problems in applying the plate-stretch model that has been established for fast reactor cores through JUPITER experimental analyses. Both keeping the pellet density and using SRAC library for the elastic cross section for lighter nuclides allow us to obtain reasonable analysis accuracy on the spectral indices that were measured at the center of the core. Application of such a cell model was justified through comparison among various cell models using continuous energy Monte-Carlo code MVP. It is confirmed that both the MOX zone and the U0 zone can be correctly evaluated by the plate-stretch model. Based on the updated cell calculation, both the effective multiplication factor (k-eff)and the spectral indexes agree well with the measured values. The transport and mesh-size correction is made for the k-eff evaluation. Those results also agree well within reasonable difference between those obtained by IPPE and CEA, which were obtained by using sub-group method or continuous-energy Monte Carlo code. Evaluation by the nuclear data library adjustment confirmed that the analyses results of the BFS-58-1-I1 core have no significant inconsistency with JUPITER experimental analyses results. Those results are quite important for starting BFS-62 cores, which will be analyzed in the framework of supporting program for Russian ...

Maxwellian-averaged cross sections calculated from JENDL-3.2

Nakagawa, Tsuneo; Chiba, Satoshi; Osaki, Toshiro*; Igashira, Masayuki*

JAERI-Research 2000-002, p.93 - 0, 2000/02

JAERI-Research-2000-002.pdf:4.41MB

no abstracts in English

Super-Phenix Benchmark used for Comparison of PNC and CEA Calculation Methods,and of JENDL-3.2 and CARNAVAL IV Nuclear Data

Hunter

PNC TN9410 98-015, 81 Pages, 1998/02

PNC-TN9410-98-015.pdf:3.15MB

The study was carried out within the framework of the PNC-CEA collaboration agreement. Data were provided, by CEA, for an experimental loading of a start-up core in Super-Phenix. This data was used at PNC to produce core flux snapshot calculations. CEA undertook a comparison of the PNC results with the equivalent calculations carried out by CEA, and also with experimental measurements from SPX. The resu1ts revealed a systematic radial flux tilt between the calculations and the reactor measurements, with the PNC tilts only 30-401 of those from CEA. CEA carried out an analysis of the component causes of the radial tilt. It was concluded that a major cause of radia1 tilt differences between the PNC and CEA calculations lay in the nuclear datasets used: JENDL-3.2 and CARNAVAL IV. For the final stage of the study, PNC undertook a sensitivity analysis, to examine the detailed differences between the two sets of nuclear data. The PNC flux calculations modelled SPX in both 2D (RZ) and 3D (hex-Z) geometries, using the diffusion programs CITATION and MOSES. The sensitivity analysis of the differences between the JENDL-3.2 and CARNAVAL IV nuclear datasets used the SAGEP calculational route. Both datasets were condensed to a single, non-standard, set of energy group boundaries. There were some incompatibilities in the cross-section formats of the two datasets. The sensitivity analysis showed that a relatively small number of nuclear data items contributed the bulk of the radial tilt difference between calculations with JENDL-3.2 and with CARNAVAL IV. A direct comparison between JENDL-3.2 and CARNAVAL IV data revealed the following. The Nu values showed little difference (<5|%). The only large fission cross-section differences were at low energy (<30% otherwise, with <10% typical). Although down-scattering reactions showed some large fractional differences, absolute differences were negligible compared with in-group scattering; for in-group scattering fractional ...

R-matrix analysis of Pu neutron transmissions and fission cross sections in energy range from 1.0 keV to 2.5 keV

H.Derrien*

Journal of Nuclear Science and Technology, 30(9), p.845 - 862, 1993/09

https://doi.org/10.3327/jnst.30.845

no abstracts in English

Evaluation of Neutron Nuclear Data for Bk and Cf

Kikuchi, Yasuyuki; Nakagawa, Tsuneo

JAERI-M 85-138, 89 Pages, 1985/09

JAERI-M-85-138.pdf:3.12MB

no abstracts in English

Evaluation of Am Neutron Data

JAERI-M 6221, 26 Pages, 1975/08

JAERI-M-6221.pdf:0.85MB

no abstracts in English

Improvement of a high-energy fission model for spallation reactions

Iwamoto, Hiroki; Meigo, Shinichiro

no journal, , 

In the research and development of spallation neutron source facilities including accelerator-driven systems, it is necessary to accurately predict the yields of fission fragments produced from the spallation reactions, which include various volatile materials such as xenon, iodine, and krypton important to their analysis of the gas treatment system for the target. To ameliorate the accuracy of the yield prediction, we have improved the high-energy fission model of Generalized Evaporation Model (GEM) implemented in the Particle and Heavy Ion Transport code System, PHITS. In this study, the description of the fission probability was modified to account for the fission cross sections for both subactinide and actinide nuclei. The fission probability was deduced from Prokofiev's phenomenological systematics of the proton-induced fission cross sections and the Lige intranuclear cascade model version 4.6, INCL4.6. Comparing with experimental data shows that the modified model can predict the proton-, neutron-, and deuteron-induced fission cross sections for a wide range of target nuclei from the threshold energies to the GeV range with markedly improved accuracy. We also provide comparisons with other spallation models embedded in MCNP6.