ACE-FRENDY-CBZ; A New neutronics analysis sequence using multi-group neutron transport calculations

Chiba, Go*; Yamamoto, Akio*; Tada, Kenichi

Journal of Nuclear Science and Technology, 60(2), p.132 - 139, 2023/02

https://doi.org/10.1080/00223131.2022.2087783

A new multi-group neutronics analysis sequence ACE-FRENDY-CBZ is proposed. This sequence is free from uses of any application libraries; with the ACE files as the starting point, multi-group cross section data of media comprising a target system are calculated with the FRENDY code, and multi-group neutron transport calculations are performed with modules of the CBZ code system. The ACE-FRENDY-CBZ sequence was tested against the eight fast neutron systems, and good agreement with the reference Monte Carlo results was obtained within 30 pcm differences in the bare systems and the thorium-reflected system, and approximately 100 pcm differences in the uranium-reflected systems. The use of the current-weighted total cross sections in the multi-group neutron transport calculations had non-negligible impacts over 100 pcm on k-eff, and the calculations with the current-weighted total cross sections systematically underestimated k-eff in the uranium-reflected systems.

MVP/GMVP version 3; General purpose Monte Carlo codes for neutron and photon transport calculations based on continuous energy and multigroup methods (Translated document)

Nagaya, Yasunobu; Okumura, Keisuke; Sakurai, Takeshi; Mori, Takamasa

JAEA-Data/Code 2016-019, 450 Pages, 2017/03

JAEA-Data-Code-2016-019.pdf:4.43MB
JAEA-Data-Code-2016-019-hyperlink.zip:2.36MB

In order to realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two Monte Carlo codes MVP (continuous-energy method) and GMVP (multigroup method) have been developed at Japan Atomic Energy Agency. The codes have adopted a vectorized algorithm and have been developed for vector-type supercomputers. They also support parallel processing with a standard parallelization library MPI and thus a speed-up of Monte Carlo calculations can be achieved on general computing platforms. The first and second versions of the codes were released in 1994 and 2005, respectively. They have been extensively improved and new capabilities have been implemented. The major improvements and new capabilities are as follows: (1) perturbation calculation for effective multiplication factor, (2) exact resonant elastic scattering model, (3) calculation of reactor kinetics parameters, (4) photo-nuclear model, (5) simulation of delayed neutrons, (6) generation of group constants, etc. This report describes the physical model, geometry description method used in the codes, new capabilities and input instructions.

Guideline of Monte Carlo calculation; Neutron/ ray transport simulation by Monte Carlo method

Nuclear Code Evaluation Special Committee of Nuclear Code Research Committee

JAERI-Review 2002-004, 401 Pages, 2002/03

JAERI-Review-2002-004.pdf:16.2MB

no abstracts in English

Revise of a basic data base for shielding design

*; Takemura, Morio*

JNC TJ9440 2000-005, 157 Pages, 2000/03

JNC-TJ9440-2000-005.pdf:3.7MB

With use of the two-dimensional discrete ordinates code DORT and the standard groupwise shielding design library JSSTDL produced from the latest evaluated nuclear data library JENDL-3.2, experimental analyses for the representative configurations in the Radial Shield Attenuation Experiment of the JASPER were performed. The results were compared with those obtained with use of traditional method DOT3.5/JSDJ2 for the previous JASPER experimetal analyses. In general, the change of the cross section library gives higher results and the change of the transport code gives lower results. Finally the new analysis method gives better agreement with the experimental results and also less deviations of calculational errors between various detectors. Experimental analyses for the thick concrete configulation in the Gap Streaming Experiment of the JASPER was also performed with the new analysis method, after solving the poor agreement found in last year with the original JASPER experimental analyses. The same tendency due to the library change was confirmed with the above mentioned analyses of the Radial Shield Attenuation Experiment. Compilation of the input data necessary for future reanalyses of important configurations in JASPER experiments were continued through the above-mentioned experimental analyses and related informations were added for repletion of the database preserved in a computer disk holding previously accumulated data. Input data descriptions were made for auxiliary routines needed for the experimental analyses and their sample data were compiled and stored in the database.

Improvement of neutron source introduction method for absolute measurements of low reactor power

Yamamoto, Toshihiro; Miyoshi, Yoshinori

Journal of Nuclear Science and Technology, 36(11), p.1069 - 1075, 1999/11

https://doi.org/10.3327/jnst.36.1069

no abstracts in English

Preparation of methods to calculate pin-wise intra-subassembly power density distribution of a new in-pile experimental reactor for FBR safety research

Mizuno, Masahiro*; Uto, Nariaki

JNC TN9400 98-007, 147 Pages, 1998/11

JNC-TN9400-98-007.pdf:8.32MB

A design study of a new in-pile experimental reactor, SERAPH (Safety Engineering Reactor for Accident PHenomenology), for FBR safety research has progressed at JNC (Japan Nuclear Cycle Development Institute). SERAPH is intended for various in-pile experiments to be performed under steady state and various transient operation modes. Heavy water is selected as a coolant material for heat removal of the SERAPH driver core during the experiments. Control rods are needed to conduct the experiments, and a control rod with heavy water follower is considered as one of the promising ideas and is now under investigation. In this idea, care must be taken to avoid production of local power peaks which are caused by neutron moderation in the follower and may appear in the vicinity of the boundary between the control rod and its neighboring fuel subassembly, since deuterium has an excellently high moderation power. Therefore, preparation of some methods of evaluating power density distribution in detail is required for control rod design. This report describes preparation of a set of neutronic calculation methods to evaluate intra-subassembly power density distribution including local power peaks around a control rod. A two-dimensional S transport calculation code TWOTRAN-II is selected as a tool for evaluating neutron transport phenomena near the control rod with no cares for statistical influence. A two-dimensional rectangular super-cell model, which is a physical model composed of a control rod and its surrounding fifteen fuel sub-assemblies, and a method to construct the super-cell model based on thirteen unit cells are created, considering neutron mean free path near a control rod. Two processing tools are newly developed to generate a material region map and mesh boundaries for an efficient super-cell construction procedure and to obtain pin-wise power densities based on calculated mesh-wise neutron flux data. The results in this report are expected to be ...

Transport calculation of thermal and cold neutrons using NMTC/JAERI-MCNP4A code system

Iga, Kiminori*; Takada, Hiroshi; *

JAERI-Tech 97-068, 58 Pages, 1998/01

JAERI-Tech-97-068.pdf:1.67MB

no abstracts in English

DOG-II:Input data generator program for DOT 3.5 code

Koizumi, Koichi; *; *; *; *; Takatsu, Hideyuki; Seki, Yasushi; Sato, Satoshi

JAERI-M 92-106, 62 Pages, 1992/08

JAERI-M-92-106.pdf:1.0MB

no abstracts in English

None

*; *; *

PNC TJ2222 92-001, 145 Pages, 1992/03

PNC-TJ2222-92-001.pdf:4.74MB

None

Monte Carlo calculations on vector supercomputers using GMVP

Nakakawa, Masayuki; Mori, Takamasa; *

Prog. Nucl. Energy, 24, p.183 - 193, 1991/00

https://doi.org/10.1016/0149-1970(90)90035-4

no abstracts in English