Sugiyama, Daisuke*; Nakabayashi, Ryo*; Tanaka, Shingo*; Koma, Yoshikazu; Takahatake, Yoko
Journal of Nuclear Science and Technology, 58(4), p.493 - 506, 2021/04
Ono, Masato; Hanawa, Yoshio; Sonobe, Hiroshi; Nishimura, Arashi; Sugaya, Naoto; Iigaki, Kazuhiko
JAEA-Technology 2020-010, 14 Pages, 2020/09
In response to new standard for regulating research and test reactor which is enforced December 18, 2013, it was carried out assessment of the probability of aircraft crashing for HTTR. According to assessment method provided in the Assessment Criteria of the Probability of Aircraft Crashing on Commercial Power Reactor Facilities, assessment was conducted targeting reactor building, spent fuel storage building and cooling tower. As a result, it was confirmed that the probability was 5.9810, which is lower than the assessment criteria 10.
Nakamura, Shoji; Kitatani, Fumito; Kimura, Atsushi; Uehara, Akihiro*; Fujii, Toshiyuki*
Journal of Nuclear Science and Technology, 56(6), p.493 - 502, 2019/06
The thermal-neutron capture cross-section()and resonance integral(I) were measured for the Np(n,)Np reaction by an activation method. A method with a Gadolinium filter, which is similar to the Cadmium difference method, was used to measure the with paying attention to the first resonance at 0.489 eV of Np, and a value of 0.133 eV was taken as a cut-off energy. Neptunium-237 samples were irradiated at the pneumatic tube of the Kyoto University Research Reactor in Institute for Integral Radiation and Nuclear Science, Kyoto University. Wires of Co/Al and Au/Al alloys were used as monitors to determine thermal-neutron fluxes and epi-thermal Westcott's indices at an irradiation position. A -ray spectroscopy was used to measure activities of Np, Np and neutron monitors. On the basis of Westcott's convention, the and I values were derived as 186.96.2 barn, and 100990 barn, respectively.
Iwamoto, Hiroki; Meigo, Shinichiro
Journal of Nuclear Science and Technology, 56(2), p.160 - 171, 2019/02
We present a new model to describe the fission probability of the high-energy fission model, as deduced from the intranuclear cascade calculation with the Intra-Nuclear Cascade model of Lige (INCL) version 4.6 and Prokofiev's phenomenological systematics of the proton-induced fission cross sections. This model is implemented in the de-excitation model of the Generalized Evaporation Model (GEM), and applied to Monte Carlo spallation reaction simulation using the Particle and Heavy Ion Transport code System (PHITS). Comparing with experimental data for subactinide nuclei shows that this model can provide a unified prediction of the proton-, neutron-, and deuteron-induced fission cross sections with markedly improved accuracy. The calculated fission fragments tend to shift to higher mass numbers. To account for the isotopic distributions of fission fragments within the framework of a coupled INCL/GEM, modification of INCL is required, especially for description of the highly-excited states of residual nuclei.
Narukawa, Takafumi; Yamaguchi, Akira*; Jang, S.*; Amaya, Masaki
Journal of Nuclear Materials, 499, p.528 - 538, 2018/02
JAEA-Research 2016-019, 53 Pages, 2017/01
Application of probability generating function for nondestructive nuclear materials assay system was studied. First, high-order neutron correlations were derived algebraically up to septuplet and basic characteristics of the correlations were investigated. It was found that higher-order correlation increases rapidly in response to the increase of leakage multiplication, crosses and leaves lower-order correlations behind, when leakage multiplication is 1.3 that depends on detector efficiency and counter setting. Next, fission rates and doubles count rates by fast neutron and by thermal neutron in their coexisting system were derived algebraically. It was found that the number of induced fissions per unit time by fast neutron and by thermal neutron, the number of induced fissions ( 1) by one source neutron, and individual doubles count rates were possible to be estimated from Rossi-alpha combined distribution and measured ratio of each area obtained by differential die-away self-interrogation and conventional assay data.
Okazaki, Yasuyuki*; Aoyagi, Kazuhei; Kumasaka, Hiroo*; Shinji, Masato*
Doboku Gakkai Rombunshu, F1 (Tonneru Kogaku) (Internet), 72(3), p.I_1 - I_15, 2016/00
In the rational tunnel support design, numerical analysis is powerful tool to know the estimation of the behavior before tunnel construction in the case of the special ground condition and limited similar construction. In order to evaluate the support structure quantitatively, it is necessary to understand the effect of the inhomogeneity of rock mass to the tunnel support stress in advance. In this study, tunnel excavation analysis considering the inhomogeneity of rock mass was carried out. The analysis results were compared with the stress measured in the tunnel support in the Horonobe underground research laboratory. As a result, it was revealed that the local stress measured in the tunnel support can be simulated by considering the inhomogeneity of rock mass stochastically. In addition, this study evaluated the effect of the inhomogeneity of rock mass to the tunnel support stress quantitatively by processing analysis results statistically.
JAEA-Data/Code 2015-015, 162 Pages, 2015/10
MOSRA-SRAC is a lattice calculation module of the Modular code System for nuclear Reactor Analyses (MOSRA). This module performs the neutron transport calculation for various types of fuel elements including existing light water reactors, research reactors, etc. based on the collision probability method with a set of the 200-group cross-sections generated from the Japanese Evaluated Nuclear Data Library JENDL-4.0. It has also a function of the isotope generation and depletion calculation for up to 234 nuclides in each fuel material in the lattice. In these ways, MOSRA-SRAC prepares the burn-up dependent effective microscopic and macroscopic cross-section data to be used in core calculations.
Sugino, Hideharu*; Ito, Hiroto*; Onizawa, Kunio; Suzuki, Masahide
Nihon Genshiryoku Gakkai Wabun Rombunshi, 4(4), p.233 - 241, 2005/12
The purpose of this research is to establish the reliability evaluation method of aged nuclear power components for seismic events from a viewpoint of long-term use of the existing light water reactor nuclear power plants. For this purpose, we developed a piping failure probability evaluation code "PASCAL-SC" based on probabilistic fracture mechanics, and a probabilistic seismic hazard evaluation code "SHEAT-FM" for calculating the seismic occurrence probability of a plant site, paying attention to aging such as fatigue crack progress by the stress corrosion cracking and seismic load in primary coolant piping system. We proposed the reliability evaluation method of aged piping for seismic events by combination of these codes. Using this method, we evaluated the reliability of a weld line in the PLR(Primary Loop Recirculation system) piping of the BWR model plant for seismic events.
Moriyama, Kiyofumi; Takagi, Seiji; Muramatsu, Ken; Nakamura, Hideo; Maruyama, Yu
Proceedings of 2005 International Congress on Advances in Nuclear Power Plants (ICAPP '05) (CD-ROM), 9 Pages, 2005/05
The containment failure probability due to ex-vessel steam explosions were evaluated for a BWR Mk-II model plant. The evaluation was made for two scenarios: a steam explosion in the pedestal area, or in the suppression pool. A probabilistic approach, Latin Hypercube Sampling (LHS), was applied for the evaluation of steam explosion loads, in which a steam explosion simulation code JASMINE was used as a physics model. The fragility curves connecting the steam explosion loads and containment failure were developed based on simplified assumptions on the containment failure scenarios. The mean conditional probabilities of containment failure per occurrence of a steam explosion were for suppression pool and for pedestal area. Note that the results depend on the assumed range of input parameters and fragility curves that involve conservatism and simplification.
Ishikura, Shuichi*; Shiga, Akio*; Futakawa, Masatoshi; Kogawa, Hiroyuki; Sato, Hiroshi; Haga, Katsuhiro; Ikeda, Yujiro
JAERI-Tech 2005-026, 65 Pages, 2005/03
Failure probability analysis was carried out to estimate the lifetime of the mercury target which will be installed into the JSNS (Japan spallation neutron source) in J-PARC (Japan Proton Accelerator Research Complex). The lifetime was estimated as taking loading condition and materials degradation into account. Considered loads imposed on the target vessel were the static stresses due to thermal expansion and static pre-pressure on He-gas and mercury and the dynamic stresses due to the thermally shocked pressure waves generated repeatedly at 25 Hz. Materials used in target vessel will be degraded by the fatigue, neutron and proton irradiation, mercury immersion and pitting damages, etc. The imposed stresses were evaluated through static and dynamic structural analyses. The material-degradations were deduced based on published experimental data. As results, it was quantitatively confirmed that the failure probability for the lifetime expected in the design is very much lower, 10 in the safety hull, meaning that it will be hardly failed during the design lifetime. On the other hand, the beam window of mercury vessel suffered with high-pressure waves exhibits the failure probability of 12%. It was concluded, therefore, that the leaked mercury from the failed area at the beam window is adequately kept in the space between the safety hull and the mercury vessel to detect mercury-leakage sensors.
Nihon Genshiryoku Gakkai Wabun Rombunshi, 3(4), p.396 - 406, 2004/12
The U. S. Nuclear Regulatory Commission has been carrying out the Accident Sequence Precursor (ASP) Program to identify and categorize precursors to potential severe core damage accident sequences using the PSA technique. The ASP Program has identified a lot of risk significant events as precursors that occurred at U. S. nuclear power plants. The results from the ASP Program include valuable information that could be useful for obtaining risk significant insights and for monitoring risk trends in nuclear power industry. The present study examines and discusses quantitative risk trends for the industry level, using two indicators, that is, the occurrence frequency of precursors and the annual core damage probability, deriving from the ASP analysis results. It is shown that the core damage risk at U.S. nuclear power plants has been lowered and the likelihood of risk significant events has been remarkably decreasing. Also, the present study demonstrates that two risk indicators used here can provide quantitative information useful for monitoring risk trends in nuclear power industry.
Nojiri, Naoki; Fujimoto, Nozomu; Mori, Tomoaki; Obata, Hiroyuki*
JAERI-Data/Code 2004-012, 65 Pages, 2004/10
DELIGHT code is a fuel cell burnup analysis code which can produce the group constants necessary for High Temperature Gas-cooled Reactors (HTGR) core analyses. Collision probability method is used to the lattice calculation. The lattice calculation model is a cylinder type fuel or a ball type fuel of the HTGR. This code characterizes the burnup calculation considering the double heterogeneity caused by coated fuel particles of the HTGR fuel. DELIGHT code has updated its nuclear data library to the latest JENDL-3.3 data, and included new burnup chain models in order to calculate high burnup HTGR cores. The material regions of the periphery burnable poisons (BPs) were divided into details in order to improve calculation accuracy of the BP lattice calculation. This report presents the revised points of the DELIGHT-8 and can be used as user's manual.
Shibata, Katsuyuki; Onizawa, Kunio; Li, Y.*; Kato, Daisuke*
International Journal of Pressure Vessels and Piping, 81(9), p.749 - 756, 2004/09
The paper describes the procedure to evaluate the ductile crack extension, where an increase in fracture resistance by a ductile crack extension is considered. Two standard -resistance curves are prepared for applying the elasto-plastic fracture criterion. Case studies concerning the effect of elasto-plastic fracture criterion were carried out using a severe PTS transient. The introduction of the elasto-plastic fracture criterion significantly contributes to remove the over-conservatism in applying the linear elastic fracture criterion. It was also found that the algorithm of the re-evaluation of crack tip characterization also has a significant effect on the failure probability.
Shibata, Katsuyuki; Onizawa, Kunio; Li, Y.*; Kanto, Yasuhiro*; Yoshimura, Shinobu*
Pressure Vessel and Piping Codes and Standards (PVP-Vol.480), p.235 - 242, 2004/00
A screening standard of small flaws that have no significant influence on the structural integrity is prescribed in ASME Code Sec.XI. From the viewpoint of probabilistic methodology, there are some concerns that weather or not the failure probability is uniform for flaws with various aspect ratios and failure frequencies are small enough. Moreover, acceptable flaws may be determined more rationally based the failure probability. A study was performed on the failure probability of RPVs with a surface flaw specified in Sec.XI using the PFM code PASCAL. A PTS transient of NRC/EPRI Benchmark Study was used. Analysis results showed that the conditional failure probability of a RPV with an initial flaw of acceptable depth depended on the aspect ratio. For a circular flaw, the failure probability is high due to the increase in crack initiation at the surface point. The maximum difference in failure probability reaches one order of magnitude. A case study for determining the acceptable flaws based on failure probability was also carried out.
Shibata, Katsuyuki; Kanto, Yasuhiro*; Yoshimura, Shinobu*; Yagawa, Genki*
Proceedings of 5th International Workshop on the Integrity of Nuclear Components, p.99 - 117, 2004/00
no abstracts in English
Hanawa, Satoshi; Ishihara, Masahiro; Shibata, Taiju
Transactions of 17th International Conference on Structural Mechanics in Reactor Technology (SMiRT-17) (CD-ROM), 6 Pages, 2003/08
From a viewpoint of advanced design method of graphite components, it is important to apply the realistic fracture model in the design method. The applicability of the microstructure based brittle fracture model under multiaxial stress condition was, therefore, investigated. The fracture model is possible to treat grain size as well as pore size with fracture mechanics approach taking account of the crystal structure of the graphite. The model was applied to the biaxial strength prediction of near isotropic nuclear graphite using grain/pore related microstructural parameters. Prediction results were compared with biaxial strength data obtained by simultaneous loadings of inner pressure and longitudinal load with thin-walled cylindrical specimen. From this study, it was found that the fracture model predicted fairly good not only mean strength but also strength distribution under biaxial stress condition, and it was concluded that the microstructure based brittle fracture model would be applicable as the advanced design method.
Teraoka, Yuden; Yoshigoe, Akitaka
Oyo Butsuri, 71(2), p.1523 - 1527, 2002/12
The translational kinetic energy of incident molecules is an important parameter for the study of surface chemical reaction mechanisms. New adsorption reactions, induced by the O translational kinetic energy of several eV, have been found in the O/Si(001) system by applying the surface photoemission spectroscopy with supersonic molecular beam techniques and the high-energy-resolution synchrotron radiation. The termination of dangling bonds affected dominantly the oxidation of dimer backbonds. By controlling the translational kinetic energy of incident O molecules, the formation of oxide-layers with a sub-nanometer scale is possible at room temperature.
Uchiyama, Tomoaki; Oikawa, Tetsukuni; Kondo, Masaaki; Watanabe, Yuichi*; Tamura, Kazuo*
JAERI-Data/Code 2002-011, 205 Pages, 2002/03
This report is a user's manual of seismic system reliability analysis code SECOM2 developed at the JAERI for system reliability analysis, which is one of the tasks of seismic probabilistic safety assessment (PSA) of nuclear power plants (NPPs). The SECOM2 code has many functions such as calculation of component and system failure probabilities for given seismic motion levels at the site of an NPP based on the response factor method, calculation of accident sequence frequencies and the core damage frequency (CDF), importance analysis using various indicators, uncertainty analysis, and calculation of the CDF taking into account the effect of the correlations of responses and capacities of components. These analyses require the fault tree (FT) representing the occurrence condition of the system failures and core damage, information about responses and capacities of the components which compose the FT, and seismic hazard curve for the NPP site as input. This report presents calculation method used in the SECOM2 code and how to use those functions in the SECOM2 code.
Shibata, Katsuyuki; Onizawa, Kunio; Li, Y.*; Kato, Daisuke*
Proceedings of 4th International Workshop on the Integrity of Nuclear Components, p.31 - 41, 2002/00
no abstracts in English