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due to perturbed fission-source distributionNagaya, Yasunobu; Mori, Takamasa; Brown, F. B.*
Monte Karuro Keisanho Kodoka No Genjo; Dai-3-Kai Monte Karuro Shimyureshon Kenkyukai Hobunshu, p.105 - 115, 2004/12
The Monte Carlo perturbation method based on the differential operator sampling method has been widely used to obtain a small change in neutronic parameters or sensitivity. The method is very effective for fixed-source problems but a difficulty arises for eigenvalue problems because the fission source distribution is perturbed. Most Monte Carlo codes assume that the source distribution is unchanged after a perturbation is introduced. However, this assumption can lead to a significant error in the perturbation estimate. Recently, a method to estimate the perturbed fission source effect has been proposed. In this method, the additional weights for the differential coefficient of the fission source at fission sites are normalized in each cycle, and the effect is estimated by propagating the normalized additional weight between cycles. The method and benchmark results have been reviewed. It has been found that this method is very effective in perturbation calculations for the effective multiplication factor.
Kumada, Hiroaki; Yamamoto, Kazuyoshi; Murayama, Yoji; Matsumura, Akira*; Nakagawa, Yoshinobu*
Monte Karuro Keisanho Kodoka No Genjo; Dai-3-Kai Monte Karuro Shimyureshon Kenkyukai Hobunshu, p.185 - 194, 2004/12
no abstracts in English
Nakashima, Hiroshi; Safety Group of J-PARC
Monte Karuro Keisanho Kodoka No Genjo; Dai-3-Kai Monte Karuro Shimyureshon Kenkyukai Hobunshu, p.75 - 83, 2004/12
Design policy for radiation safety issues, design criteria, calculation conditions for shielding design, method for design and safety estimation and the present status of shielding design are reported in Japan Proton Accelerator Research Complex.
Sakurai, Takeshi; Kosako, Kazuaki*; Mori, Takamasa
Monte Karuro Keisanho Kodoka No Genjo; Dai-3-Kai Monte Karuro Shimyureshon Kenkyukai Hobunshu, p.168 - 176, 2004/12
no abstracts in English
heating rate in the JMTR core using MCNPNagao, Yoshiharu
Nihon Genshiryoku Gakkai Monte Karuro Ho Ni Yoru Ryushi Shimyureshon No Genjo To Kadai, p.219 - 223, 2002/01
no abstracts in English
Fujimoto, Nozomu; Nojiri, Naoki; Yamashita, Kiyonobu; Shimakawa, Satoshi; Ando, Hiroei; Mori, Takamasa
Nihon Genshiryoku Gakkai Monte Karuro Ho Ni Yoru Ryushi Shimyureshon No Genjo To Kadai, p.201 - 210, 2002/01
no abstracts in English
Shirai, Nobutoshi; Sudo, Toshiyuki; Nojiri, Ichiro
Monte Karuroho Niyoru Ryoshi Shimyureshon No Genjo To Kadai, p.235 - 249, 2002/00
None
Okumura, Keisuke; Nakakawa, Masayuki; Kaneko, Kunio*; *
JAERI-Conf 2000-018, p.31 - 41, 2001/01
Burnup calculation codes based on the conventional deterministic approach often encounter difficult problems because of the constraints on the geometry description, limit of approximation on the effective resonance cross-sections, failing of the diffusion approximation due to extremely strong anisotropic or heterogenity. They are, for example, the prediction of burn characteristics of plutonium spot, core design of ultra-small reactors, analysis of the sample material in an irradiation capsule of the research rector. To deal with these problems any time, a burn-up calculation code (MVP-BURN) was developed by using a continuous energy Monte Carlo code MVP. MVP-BURN was validated by comparison with the results of deterministic codes in the international benchmark problems, and by comparison with the measured values of the spent fuel composition irradiated in a commercial reactor.
Nagao, Yoshiharu
JAERI-Conf 2000-018, p.156 - 167, 2001/01
no abstracts in English
Yokoyama, Kenji
JAERI-Conf 2000-018, p.281 - 296, 2001/01
None
Oyama, Yukio
Nihon Genshiryoku Gakkai Monte Karuro Ho Ni Yoru Ryushi Shimyureshon No Genjo To Kadai, p.183 - 191, 2001/01
no abstracts in English
