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Journal Articles

Depletion calculation of subcritical system with consideration of spontaneous fission reaction

Riyana, E. S.; Okumura, Keisuke; Sakamoto, Masahiro; Matsumura, Taichi; Terashima, Kenichi

Journal of Nuclear Science and Technology, 59(4), p.424 - 430, 2022/04

Journal Articles

Nuclear data processing code FRENDY; A Verification with HTTR criticality benchmark experiments

Fujimoto, Nozomu*; Tada, Kenichi; Ho, H. Q.; Hamamoto, Shimpei; Nagasumi, Satoru; Ishitsuka, Etsuo

Annals of Nuclear Energy, 158, p.108270_1 - 108270_8, 2021/08

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Journal Articles

Numerical investigations on the coolability and the re-criticality of a debris bed with the density-stratified configuration

Li, C.; Uchibori, Akihiro; Takata, Takashi; Pellegrini, M.*; Erkan, N.*; Okamoto, Koji*

Dai-25-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2021/07

The capability of stable cooling and avoiding re-criticality on the debris bed are the main issues for achieving IVR (In-Vessel Retention). In the actual situation, the debris bed is composed of mixed-density debris particles. Hence, when these mixed-density debris particles were launched to re-distribute, the debris bed would possibly form a density-stratified distribution. For the proper evaluation of this scenario, the multi-physics model of CFD-DEM-Monte-Carlo based neutronics is established to investigate the coolability and re-criticality on the heterogeneous density-stratified debris bed with considering the particle relocation. The CFD-DEM model has been verified by utilizing water injection experiments on the mixed-density particle bed in the first portion of this research. In the second portion, the coupled system of the CFD-DEM-Monte-Carlo based neutronics model is applied to reactor cases. Afterward, the debris particles' movement, debris particles' and coolant's temperature, and the k-eff eigenvalue are successfully tracked. Ultimately, the relocation and stratification effects on debris bed's coolability and re-criticality had been quantitatively confirmed.

Journal Articles

Monte Carlo criticality calculation of random media formed by multimaterials mixture under extreme disorder

Ueki, Taro

Nuclear Science and Engineering, 195(2), p.214 - 226, 2021/02

 Times Cited Count:1 Percentile:39.17(Nuclear Science & Technology)

A dynamical system under extreme physical disorder has the tendency of evolving toward the equilibrium state characterized by an inverse power law power spectrum. In this paper, a practically implementable three-dimensional model is proposed for the random media formed by multi-materials mixture under such a power spectrum using a randomized form of Weierstrass function, its extension covering the white noise, and partial volumes pairings of constituent materials. The proposed model is implemented in the SOLOMON Monte Carlo solver with delta tracking. Two sets of numerical results are shown using the JENDL-4 nuclear data libraries.

Journal Articles

Continuous energy Monte Carlo criticality calculation of random media under power law spectrum

Ueki, Taro

Proceedings of International Conference on Mathematics and Computational Methods applied to Nuclear Science and Engineering (M&C 2019) (CD-ROM), p.151 - 160, 2019/00

A dynamical system under extreme physical disorder has the tendency of evolving toward the equilibrium state characterized by an inverse power law spectrum. In this paper, the author proposes a practically implementable modeling of random media under such a spectrum using a randomized form of the Weierstrass function. The proposed modeling is demonstrated by the continuous energy Monte Carlo particle transport with delta tracking for the criticality calculation of a randomized version of the Topsy spherical core in International Criticality Safety Benchmark Evaluation Project.

Journal Articles

A Power spectrum approach to tally convergence in Monte Carlo criticality calculation

Ueki, Taro

Journal of Nuclear Science and Technology, 54(12), p.1310 - 1320, 2017/12


 Times Cited Count:6 Percentile:65.17(Nuclear Science & Technology)

In Monte Carlo criticality calculation, confidence interval estimation is based on the central limit theorem (CLT) for a series of tallies. A fundamental assertion resulting from CLT is the convergence in distribution (CID) of the interpolated standardized time series (ISTS) of tallies. In this work, the spectral analysis of ISTS has been conducted in order to assess the convergence of tallies in terms of CID. Numerical results indicate that the power spectrum of ISTS is equal to the theoretically predicted power spectrum of Brownian motion for effective neutron multiplication factor; on the other hand, the power spectrum of ISTS for local power fluctuates wildly while maintaining the spectral form of fractional Brownian motion. The latter result is the evidence of a case where a series of tallies is away from CID, while the spectral form supports normality assumption on the sample mean.

Journal Articles

Monte Carlo criticality analysis under material distribution uncertainty

Ueki, Taro

Journal of Nuclear Science and Technology, 54(3), p.267 - 279, 2017/03

 Times Cited Count:5 Percentile:58.67(Nuclear Science & Technology)

Analysis framework under material distribution uncertainty is investigated for the Monte Carlo (MC) criticality calculation of continuously mixed media formed via molten core concrete interaction. Deterministic trigonometric functions and randomized Weierstrass functions are utilized to represent the spatially continuous variation. Numerical results indicate that the effective multiplication factor (k$$_{rm eff}$$) under random spatial variation can depart significantly from the k$$_{rm eff}$$ of a reference uniform medium. It is also shown that the deterministic modeling provides an upper-bound measure for extreme results from random realizations.

Journal Articles

Benchmark models for criticalities of FCA-IX assemblies with systematically changed neutron spectra

Fukushima, Masahiro; Kitamura, Yasunori; Kugo, Teruhiko; Okajima, Shigeaki

Journal of Nuclear Science and Technology, 53(3), p.406 - 424, 2016/03

 Times Cited Count:8 Percentile:71.87(Nuclear Science & Technology)

Journal Articles

Assessment of human body surface and internal dose estimations in criticality accidents based on experimental and computational simulations

Sono, Hiroki; Ono, Akio*; Kojima, Takuji; Takahashi, Fumiaki; Yamane, Yoshihiro*

Journal of Nuclear Science and Technology, 43(3), p.276 - 284, 2006/03

 Times Cited Count:1 Percentile:10.64(Nuclear Science & Technology)

For a study on the applicability of a personal dosimetry method to criticality accident dosimetry, an assessment of the human body surface and internal dose estimations was performed by experimental and computational simulations. The experimental simulation was carried out in a criticality accident situation at the TRACY facility. The neutron and $$gamma$$-ray absorbed doses in muscle tissue were separately estimated by a dosimeter set of an alanine dosimeter and a thermoluminescence dosimeter made of enriched lithium tetra borate with a phantom. The computational simulation was conducted with a Monte Carlo code taking account of dose components of neutrons, prompt $$gamma$$-rays and delayed $$gamma$$-rays. The computational simulation was ascertained to be valid by comparison between the calculated dose distributions in the phantom and the measured ones. The assessment based on the experimental and computational simulations confirmed that the personal dosimetry using the dosimeter set provided a first estimation of the body surface and internal doses with precision.

Journal Articles

Extension of effective cross section calculation method for neutron transport calculations in particle-dispersed media

Yamamoto, Toshihiro; Miyoshi, Yoshinori; Takeda, Toshikazu*

Journal of Nuclear Science and Technology, 43(1), p.77 - 87, 2006/01

 Times Cited Count:6 Percentile:43.58(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Evaluation of $$gamma$$-ray dose components in criticality accident situations

Sono, Hiroki; Yanagisawa, Hiroshi*; Ono, Akio*; Kojima, Takuji; Soramasu, Noboru*

Journal of Nuclear Science and Technology, 42(8), p.678 - 687, 2005/08

 Times Cited Count:4 Percentile:32.08(Nuclear Science & Technology)

Component analysis of $$gamma$$-ray doses in criticality accident situations is indispensable for further understanding on emission behavior of $$gamma$$-rays and accurate evaluation of external exposure to human bodies. Such dose components were evaluated, categorizing $$gamma$$-rays into four components: prompt, delayed, pseudo components in the period of criticality, and a residual component in the period after the termination of criticality. This evaluation was performed by the combination of dosimetry experiments at the TRACY facility using a thermoluminescent dosimeter (TLD) made of lithium tetra borate and computational analyses using a Monte Carlo code. The evaluation confirmed that the dose proportions of the above components varied with the distance from the TRACY core tank. This variation was due to the difference in attenuation of the individual components with the distance from the core tank. The evaluated dose proportions quantitatively clarified the contribution of the pseudo and the residual components to be excluded for accurate evaluation of $$gamma$$-ray exposure.

Journal Articles

Annular core experiments in HTTR's start-up core physics tests

Fujimoto, Nozomu; Yamashita, Kiyonobu*; Nojiri, Naoki; Takeuchi, Mitsuo; Fujisaki, Shingo; Nakano, Masaaki*

Nuclear Science and Engineering, 150(3), p.310 - 321, 2005/07

 Times Cited Count:5 Percentile:37.38(Nuclear Science & Technology)

Annular cores were formed in startup-core-physics tests of the High Temperature Engineering Test Reactor (HTTR) to obtain experimental data for verification of calculation codes. The first criticality, control rod positions at critical conditions, neutron flux distribution, excess reactivity etc. were measured as representative data. These data were evaluated with Monte Carlo code MVP that can consider the heterogeneity of coated fuel particles (CFP) distributed randomly in fuel compacts directly. It was made clear that the heterogeneity effect of CFP on reactivity for annular cores is smaller than that for fully-loaded cores. Measured and calculated effective multiplication factors (k) were agreed with differences less than 1%$$Delta$$k. Measured neutron flux distributions agreed with calculated results. The revising method was applied for evaluation of excess reactivity to exclude negative shadowing effect of control rods. The revised and calculated excess reactivity agreed with differences less than 1%$$Delta$$k/k.

Journal Articles

Impact of perturbed fission source on the effective multiplication factor in Monte Carlo perturbation calculations

Nagaya, Yasunobu; Mori, Takamasa

Journal of Nuclear Science and Technology, 42(5), p.428 - 441, 2005/05

 Times Cited Count:50 Percentile:95.28(Nuclear Science & Technology)

A new method to estimate a change in the effective multiplication factor due to the perturbed fission source distribution has been proposed for Monte Carlo perturbation calculations with the correlated sampling and differential operator sampling techniques. The method has been implemented into the MVP code for verification. Simple benchmark problems have been set up for fast and thermal systems and the applicability of the method has been verified with the problems. In consequence, it has been confirmed that the method is very effective to estimate the change. It has been also shown that there are some cases where the perturbed source effect is significant and the change in reactivity cannot be estimated accurately without taking the effect into account. Even in such cases, the new method can estimate the perturbed source effect and the estimation of the change in reactivity has been remarkably improved.

Journal Articles

Examination for neutron dose assessment method from induced sodium-24 in human body in criticality accidents

Takahashi, Fumiaki; Endo, Akira; Yamaguchi, Yasuhiro

Journal of Nuclear Science and Technology, 42(4), p.378 - 383, 2005/04

 Times Cited Count:2 Percentile:18.81(Nuclear Science & Technology)

Experiments were made to verify a dose assessment method from activated sodium in body in criticality accidents. A phantom containing sodium chloride solution was irradiated in the Transient Experiment Critical Facility to simulate activation of sodium. Monte Carlo calculations were performed to obtain quantitative relation between the activity of induced Na-24 and neutron dose in the phantom. In the previous work, conversion coefficients from specific activity of induced Na-24 to neutron dose had been analyzed with the MCNP-4B code concerning neutron spectra at some hypothesized configurations. One of the prepared coefficients was applied to evaluate neutron dose from the measured activity. The estimated dose agreed with the dose analyzed by the Monte Carlo calculation in the present study within an acceptable uncertainty, which is indicated by the IAEA. In addition, the dose calculated with the prepared coefficient was close to the result measured with dosimeters. These results suggest that the prepared coefficients can be applied to dose assessments from induced Na-24 in body.

Journal Articles

Development of fission source acceleration method for slow convergence in criticality analyses by using matrix eigenvector applicable to spent fuel transport cask with axial burnup profile

Kuroishi, Takeshi; Nomura, Yasushi

Journal of Nuclear Science and Technology, 40(6), p.433 - 440, 2003/06

 Times Cited Count:1 Percentile:11.39(Nuclear Science & Technology)

Effective source acceleration method is studied in criticality safety analysis for realistic spent fuel transport cask. Various axial burnup profiles based on in-core flux measurements are proposed in the OECD/NEA/BUC benchmark Phase II-C. In some cases, calculations by ordinary Monte Carlo method show very slow convergence of fission source distribution, and unacceptably large skipped cycles are needed. The matrix eigenvector calculation that has been developed and incorporated in the ordinary Monte Carlo calculation to improve the slow convergence is applied to the benchmark. The efficiency of this method depends on the precision of matrix elements. In a certain stage of insufficient convergence of fission source distribution, especially for this benchmark of very slow convergence, more acceleration procedure causes anomalous results because of large statistical fluctuations of matrix elements corresponding to low source levels. Therefore, we propose effective source acceleration method with less calculation time than increasing histories for the estimation of matrix elements.

JAEA Reports

Revaluation of JACS code system benchmark analyses of the heterogeneous system; Fuel rods in U+Pu nitric acid solution system

Takada, Tomoyuki; Miyoshi, Yoshinori; Katakura, Junichi

JAERI-Tech 2003-036, 80 Pages, 2003/03


In order to perform accuracy evaluation of the critical calculation by the combination of multi-group constant library MGCL and 3-dimensional Monte Carlo code KENO-IV among critical safety evaluation code system JACS, benchmark calculation was carried out from 1980 in 1982. Some cases where the neutron multiplication factor calculated in the heterogeneous system in it was less than 0.95 were seen. In this report, it re-calculated by considering the cause about the heterogeneous system of the U+Pu nitric acid solution systems containing the neutron poison shown in JAERI-M 9859. The present study has shown that the keff value less than 0.95 given in JAERI-M 9859 is caused by the fact that the water reflector below a cylindrical container was not taken into consideration in the KENO-IV calculation model. By taking into the water reflector, the KENO-IV calculation gives a keff value greater than 0.95 and a good agreement with the experiment.

Journal Articles

Dose assessment from activated sodium within a body in criticality accidents

Takahashi, Fumiaki; Endo, Akira; Yamaguchi, Yasuhiro

Radiation Protection Dosimetry, 106(3), p.197 - 206, 2003/00

 Times Cited Count:3 Percentile:26.72(Environmental Sciences)

Some data were derived using recent sophisticated methods to convert rapidly specific activity of induced sodium-24 to average dose over a whole body in criticality accidents. Monte Carlo calculations using the MCNP-4B code were performed to study energy spectra of neutrons and gamma rays for some criticality systems with fissile uranium. Absorbed dose to human body and activation of sodium were also analysed against external radiation by simulations using a phantom. It was found that neutron dose assessment from induced $$^{24}$$Na would be important to give an initial guidance of a treatment. The condition of neutron exposure, however, did not influence the quantitative relation dose by gamma rays induced within a body and activity of $$^{24}$$Na. Analyses were made to clarify the dependence of conversion from $$^{24}$$Na specific activity to dose on the orientation and the size of human body. This study suggested that the size of uranium solution and material around the fuel should be informed to properly estimate dose against external photons from neutron dose.

Journal Articles

Detailed dose assessment for the heavily exposed workers in the Tokai-mura criticality accident

Endo, Akira; Yamaguchi, Yasuhiro; Takahashi, Fumiaki

Radiation Risk Assessment Workshop Proceedings, p.151 - 156, 2003/00

We have developed a new system using numerical simulation technique for analyzing dose distribution in various postures by neutron, photon and electron exposures. The system consists of mathematical human phantoms with movable arms and legs and Monte Carlo codes MCNP and MCNPX. This system was applied to the analysis of dose distribution for the heavily exposed workers in the Tokai-mura criticality accident. The paper describes the simulation technique employed and a summary of the dose analysis.

Journal Articles

New acceleration method of source convergence for loosely coupled multi unit system by using matrix K calculation

Kuroishi, Takeshi; Nomura, Yasushi

Proceedings of International Conference on the New Frontiers of Nuclear Technology; Reactor Physics, Safety and High-Performance Computing (PHYSOR 2002) (CD-ROM), 10 Pages, 2002/10

To accelerate the slow convergence of the fission source distribution, the matrix k calculation has been developed and incorporated in the ordinary Monte Carlo method. The acceleration can be performed by the fission source correction using the eigenvector of the fission matrix, if the coupling coefficients are approximately evaluated in the middle of Monte Carlo calculation. In this paper, we propose two effective applications of the matrix k, that is, the acceleration repetition method and the source generation method. The former simply repeats the matrix k calculation, and the result for the irradiated fuel pin cell shows enough effective to accelerate the fission source on the criticality estimation. However, in some cases of the loosely coupled multi unit system, the repetition of matrix k more than twice could not be carried out to get into convergence because of many units of low source level. The latter is newly devised here to apply to such cases. The checkerboard fuel storage rack is one of the typical cases, and the calculated results show the effectiveness of this method.

Journal Articles

Core calculation of the JMTR using MCNP

Nagao, Yoshiharu

JAERI-Conf 2000-018, p.156 - 167, 2001/01

no abstracts in English

38 (Records 1-20 displayed on this page)