Study on the basic core analysis of the new STACY

Gunji, Satoshi; Yoshikawa, Tomoki; Araki, Shohei; Izawa, Kazuhiko; Suyama, Kenya

Proceedings of 12th International Conference on Nuclear Criticality Safety (ICNC2023) (Internet), 8 Pages, 2023/10

Since the compositions and properties of the fuel debris are uncertain, critical experiments are required to validate calculation codes and nuclear data used for the safety evaluation. For this purpose, JAEA has been modifying a critical assembly called "STACY". The first criticality of the new STACY is scheduled for spring 2024. This paper reports the consideration results of the core configurations of the new STACY at the first criticality. We prepared two sets of gird plates with different neutron moderation conditions (their intervals are 1.50 cm and 1.27 cm). However, there is a limitation on the number of available UO fuel rods. In addition, we would like to set the critical water heights for the first criticality at around 95 cm. This is to avoid the reactive effect of the aluminum alloy middle grid plates (Approx. 98 cm high). The core configurations for the first criticality satisfying these conditions were constructed by computational analysis. A square core configuration with the 1.50 cm grid plate that is close to the optimum moderation condition needs 261 fuel rods to reach criticality. As to the 1.27 cm grid plate, we considered two core configurations with 1.80 cm intervals by using a checkerboard arrangement. One of them has two regions core configuration with 1.27 and 1.80 cm intervals, and the other has only 1.80 cm intervals. They need 341 and 201 fuel rods for the criticality, respectively. This paper shows these three core configurations and their calculation models.

Inter-codes and nuclear data comparison under collaboration works between IRSN and JAEA

Gunji, Satoshi; Araki, Shohei; Watanabe, Tomoaki; Fernex, F.*; Leclaire, N.*; Bardelay, A.*; Suyama, Kenya

Proceedings of 12th International Conference on Nuclear Criticality Safety (ICNC2023) (Internet), 9 Pages, 2023/10

Institut de radioprotection et de sret nuclaire (IRSN) and Japan Atomic Energy Agency (JAEA) have a long-standing partnership in the field of criticality safety. In this collaboration, IRSN and JAEA are planning a joint experiment using the new STACY critical assembly, modified by JAEA. In order to compare the codes (MVP3, MORET6, etc.) and nuclear data (JENDL and JEFF) used by both institutes in the planning of the STACY experiment, benchmark calculations of the Apparatus B and TCA, which are critical assemblies once owned by both institutes, benchmarks from the ICSBEP handbook and the computational model of the new STACY were performed. Including the new STACY calculation model, the calculations include several different neutron moderation conditions and critical water heights. There were slight systematic differences in the calculation results, which may have originated from the processing and/or format of the nuclear data libraries. However, it was found that the calculated results, including the new codes and the new nuclear data, are in good agreement with the experimental values. Therefore, there are no issues to use them for the design of experiments for the new STACY. Furthermore, the impact of the new TSL data included in JENDL-5 on the effective multiplication factor was investigated. Experimental validation for them will be completed by critical experiments of the new STACY by both institutes.

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.

Fluence to effective dose conversion coefficients for electrons from 1MeV to 100GeV

Tsuda, Shuichi; Endo, Akira; Yamaguchi, Yasuhiro; Sato, Osamu*

KEK Proceedings 2000-20, p.40 - 47, 2000/12

http://rcwww.kek.jp/egsconf/proceedings/2iwoegs/tsuda.pdf

no abstracts in English

Preparation of next generation set of group cross sections; A Task report to the Japan Nuclear Cycle Development Institute)

*

JNC TJ9400 2000-005, 182 Pages, 2000/03

JNC-TJ9400-2000-005.pdf:4.74MB

The SLAROM code, performing fast reactor cell calculation based on a deterministic methodology, has been revised by adding the universal module PEACO of generating Ultra-fine group neutron spectra. The revised SLAROM, then, was utilized for evaluating reaction rate distributions in ZPPR-13A simulated by a 2-dim RZ homogeneous model, although actually ZPPR-13A composed of radial heterogereous cells. The reaction rate distributions of ZPPR-13A were also calculated by the code MVP, that is a continuous energy Monte Carlo calculation code based on a probabilistic methodology. By coparing both results, it was concluded that the module PEACO has excellent capability for evaluating highly accurate effective cross sections. Also it was proved that the use of a new fine group cross section library set (next generation set), reflecting behavior of cross sections of structural materials, such as Fe and O, in the fast neutron energy region, is indispensable for attaining a better agreement within 1% between both calculation methods. Also, for production of a next generation set of group cross sections, the code NJOY97.V107 was added to the group cross section production system and both front and end processing parts were prepared. This system was utilized to produce the new 70 group JFS-3 library using the evaluated nuclear data library JENDL-3.2. Furthermore, to confirm the capability of this new group cross section production system, the above new JFS-3 library was applied to core performance analysis of ZPPR-9 core with a 2-dim RZ homogeneous model and analysis of heterogeneous cells of ZPPR-9 core by using the deterministic method. Also the analysis using the code MVP was performed. Bycoaparison of both results the following conclusion has been derived; the deterministic method, with the PEACO module for resonance cross sections, contributes to improve accuracy of predicting reaction rate distributions and Na void reactivity in fast reactor cores. And it ...

Calculation of fluence-to-dose equivalent conversion coefficients for neutrons to be used for calibration of personal dosimeters

Yoshizawa, Michio; Yamaguchi, Yasuhiro

IRPA9: 1996 International Congress on Radiation Protection, Proceedings, 3, p.41 - 43, 1996/00

no abstracts in English

Analysis of critical experiment BFS-61 by using the continuous energy Monte Carlo code MVP and the JENDL-3.1 nuclear data

; Takano, Hideki; ; A.G.Morozov*; V.S.Smirnov*; V.V.Orlov*

Proc. of ARS94 Int. Topical Meeting on Advanced Reactors Safety,Vol. 1, 0, p.544 - 548, 1994/00

no abstracts in English

Application of KENO-IV code to calculation of kinetic parameter /l

Katakura, Junichi; *; Naito, Yoshitaka

Proc. of the 91 Int. Conf. on Nuclear Criticality Safety, p.II-19 - II-23, 1991/00

no abstracts in English

KENO-IV Code Benchmark Calculation,6; Plutonium Fuel in Various Shape

; ; *

JAERI-M 9201, 51 Pages, 1980/11

JAERI-M-9201.pdf:2.78MB

no abstracts in English

KENO-IV Code Benchmark Calculation,5; Cylinders or a Tank Containing Uranyl Nitrate Solution

; *;

JAERI-M 9108, 53 Pages, 1980/10

JAERI-M-9108.pdf:1.47MB

no abstracts in English