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Nishida, Masayuki*; Harjo, S.; Kawasaki, Takuro; Yamashita, Takayuki*; Gong, W.
Quantum Beam Science (Internet), 7(1), p.8_1 - 8_15, 2023/03
Sato, Hiroyuki; Nishida, Akemi; Ohashi, Hirofumi; Muramatsu, Ken*; Muta, Hitoshi*; Itoi, Tatsuya*; Takada, Tsuyoshi*; Hida, Takenori*; Tanabe, Masayuki*; Yamamoto, Tsuyoshi*; et al.
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 7 Pages, 2017/04
JAEA, in conjunction with Tokyo City University, The University of Tokyo and JGC Corporation, have started development of a PRA method considering the safety and design features of HTGR. The primary objective of the project is to develop a seismic PRA method which enables to provide a reasonably complete identification of accident scenario including a loss of safety function in passive system, structure and components. In addition, we aim to develop a basis for guidance to implement the PRA. This paper provides the overview of the activities including development of a system analysis method for multiple failures, a component failure data using the operation and maintenance experience in the HTTR, seismic fragility evaluation method, and mechanistic source term evaluation method considering failures in core graphite components and reactor building.
Matsuda, Kosuke*; Muramatsu, Ken*; Muta, Hitoshi*; Sato, Hiroyuki; Nishida, Akemi; Ohashi, Hirofumi; Itoi, Tatsuya*; Takada, Tsuyoshi*; Hida, Takenori*; Tanabe, Masayuki*; et al.
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 7 Pages, 2017/04
This paper proposes a set of procedures for accident sequence analysis in seismic PRAs of HTGRs that can consider the unique accident progression characteristics of HTGRs. Main features of our proposed procedure are as follows: (1) Systematic analysis techniques including Master Logic Diagrams are used to ensure reasonable completeness in identification of initiating events and classification of accident sequences, (2) Information on factors that govern the accident progression and source terms are effectively reflected to the construction of event trees for delineation of accident sequences, and (3) Frequency quantification of seismically-initiated accident sequence frequencies that involve multiplepipe ruptures are made with the use of the Direct Quantification of Fault Trees by Monte Carlo (DQFM) method by a computer code SECOM-DQFM.
Nishida, Akemi; Matsubara, Hitoshi; Tian, R.; Hazama, Osamu; Suzuki, Yoshio; Araya, Fumimasa; Nakajima, Norihiro; Tani, Masayuki; Kondo, Makoto
Nihon Genshiryoku Gakkai Wabun Rombunshi, 6(3), p.376 - 382, 2007/09
Unexpected accidents such as oil-tank fires caused by the earthquake and breakage of pipes of nuclear plants have occurred over the past several years. Higher reliability is thus now increasingly expected to maintain the safety of infrastructures. We have been intensely focused on the construction of an analysis system called the "three-dimensional virtual vibration testbed," which is a numerical simulation system for a nuclear plant which considers the interconnection of machines, pipes, buildings, and their foundations under real operating conditions. In this paper, the "part-wise analysis method" is proposed in which each structural component is treated independently and analyzed as an assembly structure. Further, the system configurations in a parallel distribution environment are described. This study shows one of the successful examples of the application of this method to a nuclear-plant cooling system that has tens of millions of degrees of freedom.
Tani, Masayuki; Nakajima, Norihiro; Nishida, Akemi; Suzuki, Yoshio; Matsubara, Hitoshi; Araya, Fumimasa; Kushida, Noriyuki; Hazama, Osamu; Kondo, Makoto; Kawasaki, Kozo
Proceedings of Joint International Topical Meeting on Mathematics & Computations and Supercomputing in Nuclear Applications (M&C+SNA 2007) (CD-ROM), 12 Pages, 2007/04
Hazama, Osamu; Suzuki, Yoshio; Matsubara, Hitoshi; Tian, R.; Nishida, Akemi; Tani, Masayuki; Nakajima, Norihiro
Proceedings of 7th MpCCI User Forum, p.132 - 136, 2006/00
Nuclear power plants are large in scale and functionally very complex structures. For safety precautions, they are maintained under very strict rules. Yet, no controlled experiments are possible to deal with full-scale nuclear reactors and its cooling systems in its entirety. In order to maintain the safety of these nuclear power plants against extra-large earthquakes and aging, Japan Atomic Energy Agency (JAEA) is currently constructing a fully three-dimensional virtual earthquake testbed on the ITBL Grid infrastructure. Currently, we have developed a high-performance finite element elastostatic simulation system based on component and part-wise assembly. Using the program, we were able to construct a finite element model of an experimental high-temperature gas reactor (HTGR) called HTTR, or High Temperature engineering Test Reactor.
Fujiwara, Masayuki*; Nishida, Toshio*; *; *
PNC TJ9058 89-003, 67 Pages, 1989/05
The high Cr ferritic/martensitic steels with a superior resistance to swelling are thought to be preferrable for the long-life wrapper tube materials of the large-scale FBR's. To evaluate the applicability of the ferritic/martensitic steels strengthened by solid-solution and carbide precipitation for the wrapper tubes, the material design to fullfil the requirements for the tensile and the impact properties, and the trial-manufacturing test were performed. The composition selected as a basis was 0.14C-0.5Ni-1Cr-0.5Mo-2.0W-0.25V-0.07Nb-0.06N, and the effects of N, V, and Nb contents and the heat-treatment conditions on the tensile and the impact properties were examined. (1)The tensile and yield strength at elevated-temperatures was largely affected by the heat-treatment conditions, while the effect of the chemical compositions on the strength was small. (2)The impact properties were largely affected by both chemical composition and the heat-treatment condition. The higher contents of N, V, and Nb than those of the basic composition increased the DBTT. (3)To fullfil the requirements for the tensile and the impact properties of the wrapper tubes, the basic composition selected in this study and the heat-treatment condition, normalized at 1050
C and tempered at 710C, were thought to be suitable. (4)From the manufacturing test of the wrapper tubes applying the selected chemical composition and the heat-treatment condition, the wrapper tubes having high strength and superior impact properties exceed the requirements were obtained.
Fujiwara, Masayuki*; Nishida, Toshio*; *; Soshiroda, Tetsuo*
PNC TJ9058 88-009, 77 Pages, 1989/02
The Spray-Dispersion (SD) Method is a recently developed technique which can produce a steel containing dispersed oxide particles sprayed from outside into the molten steel stream. To evaluate the applicability of the oxide dispersion strengthened (ODS) ferritic steels made by this new process for the long life core materials of the large scale FBR's, a fundamental study using Fe-13Cr ferritic steels was conducted. The main results are summarized as follows : (1)From the results of the simulation tests utilizing water and molten Sn, it was cleared that the molten metal diameter, the throat diameter and the number of the spray gas nozzles, the spray gas nozzle angle, and the spray gas pressure are the important to get a favorable spray-dispersion condition. (2)An experimental apparatus for the SD method installed in a vacuum induction furnace was manufactured and the SD tests utilizing ZrO
oxide particles were performed on the 10kg ingots of Fe-13Cr ferritic steels containing the dispersion-controlled elements such as Ti, Nb, V. The maximum quantity of ZrO
added in the steels was about 0.5%. But the oxide particle sizes in the steels were large and the distribution of the oxide particles was not homogenious. It is considered that the higher pressure of the spray gas and the higher feed rate of the oxide powders are needed for the improvement of these problems.
Fujiwara, Masayuki*; Nishida, Toshio*; Soshiroda, Tetsuo*; *
PNC TJ9058 88-007, 189 Pages, 1988/08
As the Oxide Dispersion Strengthened (ODS) ferritic steels have excellent resistance to swelling as well as superior strength at elevated temperatures, they are expected for the long life fuel claddings of the large scale FBR's. To evaluate the applicability of the ODS ferritic steels for the fuel claddings, we investigated the effects of the manufacturing conditions, the additional elements (Nb, V, Zr), the combined strengthening of solutin (W,Mo) and dispersion hardenings, and the kinds of the oxide particles (YO
, ZrO) on the properties, especially the elevated - temperature strength, of the ODS ferritic steels. (1)The conditions of the mechanical alloying(MA) process using the attritor showed the strong effect on the properties of the ODS steels. The higher energy the alloy powders are given, the better dispersion of the oxide particles is obtained. The higher agitater speed of the attritor is thought to be the most efficient condition among the MA conditions. (2)The mixed oxied were formed in the Nb and Zr added steels, but in the V added steel the formation of the mixed oxides was not observed. The additions of Nb and V raised the elevated - temperature strength, while the addition of Zr showed little effect on the strength. (3)The additions of the solution hardening elements, Mo and W, up to about 2% increased the creep rupture strength. The effect of W was larger than that of Mo. (4)We chose ZrO instead of YO, and studied the effect of ZrO dispersion on the elevated - temperature strength of the ODS ferritic steels. Although the effect of ZrO on the strength is rather smaller than that of YO, the dispersion strengthening by ZrO particles was observed.
Fujiwara, Masayuki*; Nishida, Toshio*
PNC TJ9058 88-008, 56 Pages, 1988/07
The inner surface corrosion of the cladding as a result of FCCI is an important problem for the long-life fuel of the FBR. To decrease the corrosion, Ti was selected as an oxygen getter, and the various methods for the inner surface coating of the cladding were surveyed. The electroless Ni-Ti complex deposition process was selected as one of the methods that are thought to be commercially available and relatively inexpensive, and some preliminary tests were performed using type 316 stainless steel plates. As the test results, the uniform-thickness layer (about 20
m) of the Ni-Ti complex deposition free of defects could be obtained. The Ti particles were dispersed in the Ni deposition matrix with the maximum content of about 30% in the area, but the reaction between Ni and Ti, the compositional change and the formation of the intermetallic compounds occurred at the high temperatures. It is necessary to clarify the effectiveness of the electroless Ni-Ti complex deposition against the FCCI corrosion for the evaluation of the availability of this method.
Fujiwara, Masayuki*; Soshiroda, Tetsuo*; Nishida, Toshio*
PNC TJ9058 88-004, 61 Pages, 1988/03
Elevated-temperature tensile tests and internal creep rupture tests were performed on three kinds of fuel cladding tubes (61FK, 61FSF, 61FS) of high strength ferritic/martensitic steels, domestically test-manufactured in the fiscal year 1986. Diametrical changes during internal creep were also measured on modified type 316 stainless steel tubes test-manufactured for Monju (55MK). The base composition of the high strength ferritic/martensitic tubes was Fe-11%Cr and the additional elements were 0.5%Mo and 2.0%W on the 61FS tubes, 2.2%W on the 6IFSF tubes, and 2.0%Mo on the 61FK tubes, respectively. The structures of the 61FSF and 61FS tube were single phase of tempered martensite, while that of the 61FK tubes with low C content was dual phases of tempered martentsite and 
-ferrite. The hardness of the 61FK tubes was higher than those of the 61FSF and 61FS tubes. The test results are sumarized as follows: (1)The tensile strength and the 0.2% proof stress of the 61FK tubes were rather higher than those of the 61FSF and 61FS tubes at the test temperature range from R.T. to 800C. The elongation was not different among the tubes. (2)The 61FK tubes showed the highest creep rupture strength among the three kinds of tubes and the strength of the 61FS tubes was rather higher than that of the 61FSF tubes at 600C, while the difference in the strength was not observed at 650C and at a longer period of time. (3)The creep rupture strength at 600 C of the three kinds of tubes was almost equivalent to that of the 59FK tubes test-manufactured in the fiscal year 1984, but the strength was higher than that of the 59FK tubes at 650C. (4)The effect of the interrupted period of time for measuring the diameter on the creep elongation was not found during internal creep of the 55MK tubes at 650C and ...
Matsubara, Hitoshi; Minami, Takahiro; Hazama, Osamu; Nishida, Akemi; Tian, R.; Nakajima, Norihiro; Tani, Masayuki
no journal, ,
A nuclear power plant is made up of numerous components. In previous techniques, structural analyses of entire nuclear power plant have not been achieved because it is assumed to be united structure. In this work, through our new model approach called the assembled-structure analysis, simulation of an entire nuclear power plant by assembling of individual components was made possible.
Tani, Masayuki; Suzuki, Yoshio; Nishida, Akemi; Nakajima, Norihiro
no journal, ,
no abstracts in English
Suzuki, Yoshio; Kushida, Noriyuki; Yamagishi, Nobuhiro; Minami, Takahiro; Matsumoto, Nobuko; Nakajima, Kohei; Nishida, Akemi; Matsubara, Hitoshi; Tian, R.; Hazama, Osamu; et al.
no journal, ,
no abstracts in English
Sato, Hiroyuki; Nishida, Akemi; Furuya, Osamu*; Muramatsu, Ken*; Itoi, Tatsuya*; Takada, Tsuyoshi*; Tanabe, Masayuki*; Yamamoto, Tsuyoshi*
no journal, ,
The proposed research aims to establish a probabilistic risk assessment method for high temperature gas-cooled reactors fully utilizing their design and safety characteristics. The method will be developed for the incorporation of a graded approach as well as a component failure evaluation model using the operation and maintenance experience in the high temperature engineering test reactor into an accident frequency analysis. In addition, a source term evaluation method considering failures in core graphite components will be developed.
Honda, Yuki; Sato, Hiroyuki; Ohashi, Hirofumi; Nishida, Akemi; Muta, Hitoshi*; Muramatsu, Ken*; Itoi, Tatsuya*; Tanabe, Masayuki*
no journal, ,
We have been conducting a source term evaluation method development for high temperature gas-cooled reactors considering structural failures in the major components. We will present the outline of evaluation method and results of transient analysis for unprotected depressurized loss-of-forced cooling accident which may be initiated by Beyond-Design-Basis Earthquake.
Muramatsu, Ken*; Muta, Hitoshi*; Matsuda, Kosuke*; Sato, Hiroyuki; Nishida, Akemi; Ohashi, Hirofumi; Itoi, Tatsuya*; Tanabe, Masayuki*
no journal, ,
As part of PRA method development for High temperature Gas-cooled Reactor, domestic and international PRA standards are investigated to identify requirements related to accident scenario assessment and reliability database development. It was concluded that there are three key thing to note for the PRA method development, that is, detail analysis for passive components, integration of mechanistic source term evaluation and supplementation of reliability data.
Sato, Hiroyuki; Nishida, Akemi; Muramatsu, Ken*; Muta, Hitoshi*; Itoi, Tatsuya*; Takada, Tsuyoshi*; Tanabe, Masayuki*; Yamamoto, Tsuyoshi*
no journal, ,
Japan Atomic Energy Agency, in conjunction with Tokyo City University, The University of Tokyo and JGC Corporation, have started development of a probabilistic risk assessment method considering the safety and design features of High Temperature Gas-cooled Reactor (HTGR). The presentation will focus on project overview and progress in FY2017 and FY2018.
Sato, Hiroyuki; Nishida, Akemi; Muramatsu, Ken*; Muta, Hitoshi*; Itoi, Tatsuya*; Takada, Tsuyoshi*; Tanabe, Masayuki*; Yamamoto, Tsuyoshi*
no journal, ,
The primary objective of the research is to develop a PRA method considering safety and design features of HTGRs. This presentation provides the overview and achievements of activities including development of a system analysis method and a source term evaluation method for multiple failures in passive SSCs as well as a seismic fragility evaluation method which enables identification of practical accident scenario.
Sato, Hiroyuki; Nishida, Akemi; Ohashi, Hirofumi; Muramatsu, Ken*; Muta, Hitoshi*; Matsuda, Kosuke*; Itoi, Tatsuya*; Takada, Tsuyoshi; Tanabe, Masayuki*; Yamamoto, Tsuyoshi*; et al.
no journal, ,
