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JAEA Reports

Neutron flux estimation and neutronics characteristics calculation in post-JMTR conceptual study

Oizumi, Akito; Akie, Hiroshi

JAEA-Technology 2023-017, 93 Pages, 2023/12

JAEA-Technology-2023-017.pdf:8.45MB

After the decision of decommissioning JMTR (Japan Materials Testing Reactor), Japan Atomic Energy Agency investigated the possibility to construct a new irradiation test reactor to succeed JMTR (post-JMTR), and the final report of the investigated result was submitted to the Ministry of Education, Culture, Sports, Science and Technology on March 30th 2021. This investigation was carried out in 4 steps of (1) selection of reactor type, (2) reactor core plans studies, (3) neutronic studies, (4) thermal studies, and was finally (5) considered and evaluated. This JAEA-Technology report summarizes the process and the results of (3) neutronic studies. Neutron fluxes were calculated at irradiation sample positions in the investigated cores, the standard core and the compact core, and the calculated fluxes satisfied the required irradiation capability. It was also evaluated the two investigated cores' continuous reactor operation time in days in one refueling cycle, and the results guaranteed an operation days equality with that of existing JMTR. In addition, neutronic characteristics of the cores were estimated, such as power distribution in the core, control rod reactivity worth, reactivity coefficients, distribution of fuel burnup rate of each fuel element, and kinetics parameters. The evaluated neutronic characteristics were used in the post-JMTR final investigation report to confirm the neutronic feasibility by comparing with the neutronic limiting values of existing JMTR, and to estimate the cooling capability to make the core thermally feasible.

JAEA Reports

Investigation of the core neutronics analysis conditions for evaluation of burn-up nuclear characteristics of the next-generation fast reactors

Takino, Kazuo; Oki, Shigeo

JAEA-Data/Code 2023-003, 26 Pages, 2023/05

JAEA-Data-Code-2023-003.pdf:1.66MB

Since next-generation fast reactors aim to achieve a higher core discharge burn-up than conventional reactors do, core neutronics design methods must be refined. Therefore, a suitable analysis condition is required for the analysis of burn-up nuclear characteristics to accomplish sufficient estimation accuracy while maintaining a low computational cost. We investigated the effect of the analysis conditions on the accuracy of estimation of the burn-up nuclear characteristics of next-generation fast reactors in terms of neutron energy groups, neutron transport theory, and spatial mesh. This study treated the following burn-up nuclear characteristics: criticality, burn-up reactivity, control rod worth, breeding ratio, assembly-wise power distribution, maximum linear heat rate, sodium void reactivity, and Doppler coefficient for the equilibrium operation cycle. As a result, it was found that the following conditions were the most suitable: 18-energy-group structure, 6 spatial meshes per assembly with diffusion approximation. Additionally, these conditions should apply to correction factors for energy group structure, spatial mesh and transport effects.

JAEA Reports

Development of the unified cross-section set ADJ2017R

Yokoyama, Kenji; Maruyama, Shuhei; Taninaka, Hiroshi; Oki, Shigeo

JAEA-Data/Code 2021-019, 115 Pages, 2022/03

JAEA-Data-Code-2021-019.pdf:6.21MB
JAEA-Data-Code-2021-019-appendix(CD-ROM).zip:435.94MB

In JAEA, several versions of unified cross-section set for fast reactors have been developed so far; we have developed a new unified cross-section set ADJ2017R, which is an improved version of the unified cross-section setADJ2017 for fast reactors. The unified cross-section set is used for reflecting information of C/E values (analysis / experiment values) obtained by integral experiment analyses in reactor core design via the cross-section adjustment methodology; the values are stored in the standard database for FBR core design. In the methodology, the cross-section set is adjusted by integrating the information such as uncertainty (covariance) of nuclear data, uncertainty of integral experiment / analysis, sensitivity of integral experiment with respect to nuclear data. ADJ2017R basically has the same performance as ADJ2017, but we conducted an additional investigation on ADJ2017 and revised the following two points. The first is to unify the evaluation method of the correlation coefficient of uncertainty caused by experiments (hereinafter referred to as the experimental correlation coefficient). Because it was found that the common uncertainty used in the evaluation of the experimental correlation coefficient was evaluated by two different methods, the experimental correlation coefficients were revised for all experimental data, and the evaluation method was unified. The second is the review of the integral experiment data used for the cross-section adjustment calculation. It was found that one of the experimental values of composition ratio after irradiation of the Am-243 sample has a problem in uncertainty evaluation because its experimental uncertainty is extremely small compared to the others. The cross-section adjustment calculation was, therefore, redone by excluding the experimental value. In the creation of ADJ2017, a total of 719 data sets were analyzed and evaluated, and eventually adopted 620 integral experimental data sets. In contrast, a total of 61

Journal Articles

Deterioration of pulse characteristics and burn-up effects with an engineering model in Japanese spallation neutron source

Harada, Masahide; Watanabe, Noboru; Teshigawara, Makoto; Kai, Tetsuya; Maekawa, Fujio; Kato, Takashi; Ikeda, Yujiro

LA-UR-06-3904, Vol.2, p.700 - 709, 2006/06

Pulse characteristics data for every neutron beam line are indispensable in designing devices for neutron scattering experiments of JSNS. A detailed model was built and pulse characteristics of each beam line were estimated using the PHITS code and the MCNP-4C code. These results have been disclosed on the J-PARC homepage since September 2004. Due to changes of moderator shapes in a progress of manufacture design, we observed from the calculation that pulse structures of decoupled moderators were deteriorated, especially, those of pulse tail. We found that this deterioration was caused by leakage neutron from gaps between decouplers and absorbing liners of the reflector. For a final stage of the manufacture design, we carefully tried to find other factors which deteriorated the pulse characteristics. Furthermore, pulse structures of poisoned and unpoisoned decoupled moderators were evaluated with the consideration of heterogeneous burn-up and leakage neutron spectra including high-energy region up to GeV were estimated for each neutron beam hole.

JAEA Reports

Evaluation of neutronic characteristics of TRACY water-reflected core system

Sono, Hiroki; Yanagisawa, Hiroshi*; Miyoshi, Yoshinori

JAERI-Tech 2003-096, 84 Pages, 2004/01

JAERI-Tech-2003-096.pdf:3.6MB

Prior to the supercritical experiments using a water-reflected core of the TRACY Facility, neutronic characteristics regarding criticality and reactivity of the core system were evaluated. In the analyses, a continuous energy Monte Carlo code, MVP, and a two-dimensional transport code, TWOTRAN, were used together with a nuclear data library, JENDL-3.3. By comparison to the characteristics in the former-used bare core system of TRACY, the water reflector was estimated not to change the kinetic parameter and to reduce the critical solution level by $$sim$$20 %, the temperature coefficient of reactivity by 6$$sim$$10 %, and the void coefficient of reactivity by $$sim$$18 %, respectively. According to the Nordheim-Fuchs model, the first peak power during a power excursion was evaluated to be $$sim$$15 % smaller than that in the bare system under the same conditions of fuel and reactivity insertion. The influence of the void feedback effect of reactivity, which is left out of consideration in the model, on the power characteristics will be evaluated from the results of the experiments.

JAEA Reports

Neutronics analysis of the improved LEU core in JMTR

Komukai, Bunsaku; Naka, Michihiro; Tabata, Toshio; Nagao, Yoshiharu; Takeda, Takashi*; Fujiki, Kazuo

JAERI-Tech 2002-067, 75 Pages, 2002/08

JAERI-Tech-2002-067.pdf:3.41MB

no abstracts in English

Journal Articles

Neutronic optimization of premoderator and reflector for decoupled hydrogen moderator in 1MW spallation neutron source

Harada, Masahide; Teshigawara, Makoto; Kai, Tetsuya; Sakata, Hideaki*; Watanabe, Noboru; Ikeda, Yujiro

Journal of Nuclear Science and Technology, 39(8), p.827 - 837, 2002/08

 Times Cited Count:19 Percentile:72.74(Nuclear Science & Technology)

For a decoupled hydrogen (super critical) moderator, optimization studies have been performed on a premoderator and reflector material (Pb, Be, Fe and Hg) together with the decoupling energy to realize a higher neutronic performance. The result indicated that the best neutronic performance could be obtained for a decoupled H$$_2$$ moderator in a Pb reflector by optimizing the premoderator and adopting an appropriate decoupling energy, even compared with optimized one in a Be reflector.

JAEA Reports

Neutronic characteristics of coupled moderator proposed in integrated model

Teshigawara, Makoto; Meigo, Shinichiro; Sakata, Hideaki*; Kai, Tetsuya; Harada, Masahide; Ikeda, Yujiro; Watanabe, Noboru

JAERI-Research 2001-022, 33 Pages, 2001/05

JAERI-Research-2001-022.pdf:5.16MB

no abstracts in English

JAEA Reports

Neutronics design of JMTR-LEU core

Nagaoka, Yoshiharu; Komukai, Bunsaku; ; Koike, Sumio; Saito, Minoru;

JAERI-M 92-098, 81 Pages, 1992/07

JAERI-M-92-098.pdf:1.81MB

no abstracts in English

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