Hamon, 30(1), p.7 - 8, 2020/02
Safety review of JRR-3 under the New Regulatory Requirements was completed on 7th November 2018. Neutron beam will come back in early 2021 after reinforcement works of the roof of the reactor building, the peripheral structures like a stack, a cooling tower, and the experimental hall. The future of neutron sciences using the research reactor strongly depends on how many impacted researches using JRR-3 are achieved after restarting JRR-3. At this stage, we can learn a lot of things from the back numbers of HAMON.
Tsuruga Comprehensive Research and Development Center
JAEA-Technology 2019-007, 159 Pages, 2019/07
This report summarizes the history and achievements of the prototype fast breeder reactor Monju. The development of Monju started in 1968 as a prototype reactor following the experimental fast reactor Joyo. The development covers all the activity related to the fast reactor; plant design, mockup tests, construction, operation, and plant management. This report summarizes the history and achievements for 11 technical areas: history and principal achievements, design and construction, operation test, plant safety, core physics, fuel, plant system, sodium technology, materials and mechanical design, plant management, and trouble management.
Murakami, Masashi; Hoshino, Yuzuru; Nakatani, Takayoshi; Sugaya, Toshikatsu; Fukumura, Nobuo*; Sanda, Toshio*; Sakai, Akihiro
JAEA-Technology 2019-003, 50 Pages, 2019/06
Toward the establishment of a common approach to determine the radioactivity concentrations in dismantling wastes arising from research reactors, radionuclide concentrations in the reactor structure materials of aluminum, carbon steel, shield concrete, and graphite of TRIGA Mark II reactor at Rikkyo University, Japan, were evaluated with both radiochemical analysis and theoretical calculation. The measured nuclides by the radiochemical analysis were H, Co, and Ni in aluminum, H, Co, Ni, and Eu in carbon steel, H, Co, and Eu in shield concrete, and H, C, Co, Ni, and Eu in graphite. Neutron-flux distributions and neutron-induced activities were computed with DORT and ORIGEN-ARP codes, respectively. Using the results of material composition analysis, radioactivity concentrations were conservatively predicted with good accuracy except for graphite material.
Nippon Genshiryoku Gakkai-Shi, 61(4), p.270 - 272, 2019/04
no abstracts in English
Genshiryoku No Ima To Ashita, p.264 - 265, 2019/03
no abstracts in English
Kasahara, Shigeki; Fukuya, Koji*; Fujimoto, Koji*; Fujii, Katsuhiko*; Chimi, Yasuhiro
JAEA-Review 2018-013, 171 Pages, 2019/01
For structural integrity assessment of reactor internals of light water reactors, it is important to evaluate and predict the property changes of structural materials, based on existing data obtained from austenitic stainless steel irradiated with neutrons. Compilation of the data into tables is valuable for discussing the representative or the most probable values of the properties applied to the assessment. When the data compilation, the data must be distinguished clearly in consideration of different service conditions of core internals of pressurized water reactors (PWR) and boiling water reactors. Main objective of this work is to provide material property tables of irradiated austenitic stainless steel which will be applicable for assessment of structural integrity of core internals of PWRs. To compile the table, published literature reporting irradiated stainless steel data were surveyed and screened by considering the service conditions of PWRs. In addition to the data, various parameters for the data evaluation, e.g. chemical compositions and pre-treatments of the materials, irradiation and examination conditions, were extracted from the literature, and listed into tables.
Katsuyama, Jinya; Uno, Shumpei*; Watanabe, Tadashi*; Li, Y.
Frontiers of Mechanical Engineering, 13(4), p.563 - 570, 2018/12
For the structural integrity assessments on reactor pressure vessels (RPVs) under pressurized thermal shock (PTS) events, thermal hydraulic (TH) behavior of coolant water is one of the most important influence factors. Configuration of plant equipment and their dimensions, and operator action have large influences on TH behavior. In this study, to investigate the influence of operator action on TH behavior during a PTS event, we developed an analysis model for a typical Japanese plant, and performed TH and structural analyses. Two different operator action times were assumed based on the Japanese and US' rules. From the analysis results, it was clarified that differences in operator action times have a significant effect on TH behavior and loading conditions, that is, following the Japanese rule may lead to lower stresses compared to that when following the US rule because earlier operator action caused lower pressure in the RPV.
Kasahara, Shigeki; Fukuya, Koji*; Koshiishi, Masato*; Fujii, Katsuhiko*; Chimi, Yasuhiro
JAEA-Review 2018-012, 180 Pages, 2018/11
For structural integrity assessment of reactor internals of light water reactors, it is important to evaluate and predict the property changes of structural materials, based on existing data obtained from austenitic stainless steel irradiated with neutrons. Compilation of the data into tables is valuable for discussing the representative or the most probable values of the properties applied to the assessment. In the process of the data compilation, the data must be distinguished clearly in consideration of different service conditions of core internals of boiling water reactors (BWR) and pressurized water reactors. Main objective of this work is to provide material property tables of irradiated austenitic stainless steel which will be applicable for assessment of structural integrity of core internals of BWRs. To compile the table, published literature reporting irradiated stainless steel data were surveyed and screened by considering the service conditions of BWRs. In addition to the data, various parameters for the data evaluation, e.g. chemical compositions and pre-treatments of the materials, irradiation and examination conditions, were extracted from the literature, and listed into the tables.
Soma, Yasutaka; Komatsu, Atsushi; Ueno, Fumiyoshi
Zairyo To Kankyo, 67(9), p.381 - 385, 2018/09
In-situ measurement of electrical conductivity of solution within crevice of SUS316L stainless steel in 288C water has been conducted with newly developed electrochemical sensor system. The sensor measures local electrical conductivity of crevice solution beneath the electrode () with electrochemical impedance method. The sensors were installed at different positions within tapered crevice of SUS316L stainless steel. The crevice specimen with the sensors were immerged into 288C, 8 MPa, pure oxygen saturated high purity water for 100 h. at a position with crevice gap of 59.3m was 8-11S/cm, least deviate from conductivity of 288C pure water (4.4S/cm) and no localized corrosion occurred. On the contrary, at a position with crevice gap of 4.4m increased with time and showed maximum value of 1600S/cm at 70 h. Localized corrosion occurred in the vicinity of this position. Thermodynamic equilibrium calculation showed of 1600S/cm being equivalent to pH of 3 to 3.7. It can be concluded that acidification occurred in tight crevice even under high purity bulk water and resulted in localized corrosion.
Tanaka, Masaaki; Ono, Ayako; Hamase, Erina; Ezure, Toshiki; Miyake, Yasuhiro*
Nippon Kikai Gakkai Kanto Shibu Ibaraki Koenkai 2018 Koen Rombunshu (CD-ROM), 4 Pages, 2018/08
Decay heat removal system (DHRS) by using the natural circulation without depending on the pump as the mechanical equipment is recognized as one of the most effective methodologies for the sodium-cooled fast reactor from the viewpoint of the safety enhancement. The numerical estimation method which can predict thermal hydraulic phenomena in the natural circulation under the plant cooling process by operating the various DHRSs including the severe accident is necessarily required. In this paper, the numerical results of the preliminary analysis for the sodium experiment condition with the apparatus of PLANDTL-2, in which the core and the upper plenum with a dipped-type direct heat exchanger (DHX) were modeled, were discussed, in order to establish an appropriate numerical models for the direct heat exchanger (DHX).
Yamamoto, Akio*; Chiba, Go*; Kirimura, Kazuki*; Miki, Yosuke*; Yokoyama, Kenji
Nippon Genshiryoku Gakkai-Shi, 60(4), p.241 - 245, 2018/04
no abstracts in English
Katsuyama, Jinya; Masaki, Koichi; Miyamoto, Yuhei*; Li, Y.
JAEA-Data/Code 2017-015, 229 Pages, 2018/03
As a part of the structural integrity research for aging light water reactor components, a probabilistic fracture mechanics (PFM) analysis code PASCAL has been developed in JAEA. The PASCAL code can evaluate the conditional failure probabilities and failure frequencies for core region in reactor pressure vessels under the pressurized thermal shock events. In this study, we improved many functions such as the stress intensity factor solutions, the fracture toughness models, or confidence level evaluation function by considering epistemic and aleatory uncertainties related to influence parameters in the structural integrity assessment. We also developed the analysis module PASCAL-Manager which calculates the failure frequency for the entire core region taking into consideration the failure probabilities obtained from PACAL-RV. Based on these improvements, the new analysis code is upgraded to PASCAL Ver.4. This report provides the user's manual and theoretical background of PASCAL Ver.4.
Ebisuzaki, Toshikazu*; Wada, Satoshi*; Saito, Norihito*; Fujii, Takashi*; Nishimura, Akihiko
Reza Kenkyu, 45(10), p.664 - 665, 2017/10
no abstracts in English
Katsuyama, Jinya; Yamaguchi, Yoshihito; Nemoto, Yoshiyuki; Kaji, Yoshiyuki; Osaka, Masahiko
Transactions of 24th International Conference on Structural Mechanics in Reactor Technology (SMiRT-24) (USB Flash Drive), 11 Pages, 2017/08
It is difficult to assess rupture behavior of the lower head of reactor pressure vessel in boiling-water-type nuclear power plants due to severe accident like Fukushima Daiichi because Boiling Water Reactor (BWR) lower heads have geometrically complicated structure with a lot of penetrations. Therefore, we have been developing an analysis method to predict time and location of RPV lower head rupture of BWRs considering creep damage mechanisms based on coupled analysis of three-dimensional Thermal-Hydraulics (TH) and thermal-elastic-plastic-creep analyses. In this study, we performed creep damage evaluations to investigate the effects of the debris depth and heat generation locations on failure behavior of lower head. From the analysis results, we discussed the outflow paths of the relocated molten core to the containment, and it was concluded that failure regions of BWR lower head are only the control rod guide tubes or stub tubes under simulated conditions.
Choi, B.; Nishida, Akemi; Muramatsu, Ken*; Takada, Tsuyoshi*
Transactions of 24th International Conference on Structural Mechanics in Reactor Technology (SMiRT-24) (USB Flash Drive), 10 Pages, 2017/08
In order to clarify the influence of the modeling method on the result of seismic response analysis of nuclear facility, seismic response analysis using various simulated input ground motions was carried out and the uncertainty of response results were statistically analyzed. In particular, we focused on the difference of the response due to the structural modeling method (a conventional sway-rocking model and 3D FE model), and the relations among the input level, floor position, and response results were described and discussed.
Kaku Deta Nyusu (Internet), (117), p.5 - 14, 2017/06
The benchmark calculation is one of the main activities of the Nuclear Science Committee under the Nuclear Energy Agency of the Organization for Economic Cooperation and Development (OECD/NEA/NSC). The international benchmark relatively frequently means the benchmark activity carried out by the NEA. In this manuscript, the author discusses the significance of the international benchmark by describing (i) the current status of the benchmark in the field of the reactor physics conducted by the OECD/NEA/NSC, (ii) revision of the neutronics calculation code system to reflect the results of the benchmark, (iii) the benchmark calculation as the asset for the future research and development, (iv) examples of the benchmark calculation based on the experimental data, and (v) how to propose the benchmark in the OECD/NEA/NSC.
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 9 Pages, 2017/04
Loss-of-heat-removal-system (LOHRS) events are identified as some of most dominant severe accident sequences in a sodium-cooled fast reactor. Safety margins against significant core damage in LOHRS events were therefore studied in this paper assuming large fuel-cladding gap and fuel cladding failure. It was clarified through analyses by the developed code that neither fuel melting nor further mechanical pin failure occurs owing to large fuel-cladding gap and fuel cladding failure. It was therefore concluded that large safety margins against significant core damage are provided during LOHRS events. These results will be effectively used in formulating the safety criteria for severe accidents or beyond-design-basis-accidents as one of the supporting evidences to be seriously considered.
Katsuyama, Jinya; Osakabe, Kazuya*; Uno, Shumpei; Li, Y.
JAEA-Research 2016-022, 40 Pages, 2017/02
For reactor pressure vessels (RPVs) in the light water reactors, the fracture toughness decreases due to the neutron irradiation embrittlement with operating years. In Japan, to prevent RPVs from a nil-ductile fracture, deterministic fracture mechanics methods in accordance with the codes provided by the Japan Electric Association are performed for assessing the structural integrity of RPVs under the pressurized thermal shock (PTS) events by taking the neutron irradiation embrittlement into account. On the other hand, in recent years, probabilistic methodologies for PTS evaluation are introduced into regulations in Europe and the United States. For example, in the United States, a PTS screening criterion related to the reference temperature derived by the probabilistic method is stipulated. If the screening criterion is not satisfied, it is approved to perform the evaluation based on the probabilistic method by calculating numerical index such as through-wall crack frequency (TWCF). To reach the objectives that persons who have knowledge on the fracture mechanics can carry out the PFM analyses and obtain TWCF for a domestic RPVs by referring to this report, we develop the guideline on a structural integrity assessment method based on PFM by reflecting the latest knowledge and expertise.
Ichikawa, Shoichi; Chiba, Yusuke; Ono, Fumiyasu; Hatori, Masakazu; Kobayashi, Takanori; Uekura, Ryoichi; Hashiri, Nobuo*; Inuzuka, Taisuke*; Kitano, Hiroshi*; Abe, Hisashi*
JAEA-Research 2016-021, 32 Pages, 2017/02
In order to reduce the influence on a plant schedule of the MONJU by the maintenance of dew point hygrometers, The JAEA examined a capacitance type dew point hygrometer as an alternative dew point hygrometer for a lithium-chloride type dew point hygrometer which had been used at the CV-LRT in the MONJU. As a result of comparing a capacitance type dew point hygrometer with a lithium-chloride type dew point hygrometer at the CV-LRT (Atmosphere: nitrogen, Testing time: 24 hours), there weren't significant difference between a capacitance type dew point hygrometer and a lithium-chloride type dew point hygrometer. As a result of comparing a capacitance dew point hygrometer with a high-mirror-surface type dew point hygrometer for long term verification (Atmosphere: air, Testing time: 24 months), the JAEA confirmed that a capacitance type dew point hygrometer satisfied the instrument specification (2.04C) required by the JEAC4203-2008.
Yamamoto, Keiichi; Takeuchi, Tomoaki; Hayashi, Takayasu*; Kosuge, Fumiaki*; Tsuchiya, Kunihiko
JAEA-Testing 2016-002, 25 Pages, 2016/11
Development of the reactor measurement system has been carried out to obtain the real-time in-core nuclear and thermal information, where the quantitative measurement of brightness of Cherenkov light was investigated. The system would be applied as a monitoring system in severe accidents and for the advanced operation management technology in existing LWRs. This report summarized the modification of Cherenkov light estimation system described JAEA-Testing 2015-001 and the result of the burn-up evaluation by Cherenkov light image emitted from spent fuel elements of LWRs with the modified system.