New Zr alloys for fuel cladding with different compositions from conventional ones have been de-veloped to increase the safety of nuclear power plants and to utilize existing nuclear power plants more effectively. Since the irradiation growth of fuel cladding is one of the most important parameters regarding the dimensional stability of a fuel rod and / or fuel assembly during irradiation, the irradiation growth behavior of the improved Zr alloys for light-water reactor fuel cladding was investigated. Coupon specimens, which were prepared from fuel cladding tubes with improved Zr alloys, were irra-diated in the Halden reactor in Norway at temperatures of 300 and 320 ℃ under a typical water chem-istry condition of a PWR and at 240 ℃ under the coolant condition of the Halden reactor up to a fast neutron fluence of ～ 8 × 10 25 （ 1 / m 2 , E ＞ 1 MeV ） . During and after the irradiation test, the amount of irradiation growth of each specimen was evaluated. The effect of the difference in alloy composition on irradiation growth behavior seemed insignificant if the other conditions, such as the final heat treat ment condition at fabrication, the irradiation temperature and the amount of hydrogen precharged in the specimen, were the same.
{"title":"Irradiation Growth Behavior of Improved Alloys for Fuel Cladding","authors":"K. Kakiuchi, M. Amaya","doi":"10.3327/taesj.j18.047","DOIUrl":"https://doi.org/10.3327/taesj.j18.047","url":null,"abstract":"New Zr alloys for fuel cladding with different compositions from conventional ones have been de-veloped to increase the safety of nuclear power plants and to utilize existing nuclear power plants more effectively. Since the irradiation growth of fuel cladding is one of the most important parameters regarding the dimensional stability of a fuel rod and / or fuel assembly during irradiation, the irradiation growth behavior of the improved Zr alloys for light-water reactor fuel cladding was investigated. Coupon specimens, which were prepared from fuel cladding tubes with improved Zr alloys, were irra-diated in the Halden reactor in Norway at temperatures of 300 and 320 ℃ under a typical water chem-istry condition of a PWR and at 240 ℃ under the coolant condition of the Halden reactor up to a fast neutron fluence of ～ 8 × 10 25 （ 1 / m 2 , E ＞ 1 MeV ） . During and after the irradiation test, the amount of irradiation growth of each specimen was evaluated. The effect of the difference in alloy composition on irradiation growth behavior seemed insignificant if the other conditions, such as the final heat treat ment condition at fabrication, the irradiation temperature and the amount of hydrogen precharged in the specimen, were the same.","PeriodicalId":55893,"journal":{"name":"Transactions of the Atomic Energy Society of Japan","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2020-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"69437503","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The forward weighted consistent adjoint driven importance sampling (FW-CADIS) method has been proposed as a method for obtaining variance reduction parameters to estimate flux or dose rate distribution over a wide area, or responses at multiple localized detectors for a particle transport calculation based on the stochastic Monte Carlo method with a reasonable and high accuracy. The method has been applied to estimating the dose equivalent rate around the Japanese exclusive ship, “Seiei Maru”, which transports low-level radioactive waste. The particle transport calculation was performed using a mesh tally on the entire surface of the hatch cover above low-level radioactive waste packages stacked in the cargo hold and point detector tallies at each measurement point. The statistical error is spatially uniform for the mesh tally and is reduced only around the vicinity of each point detector tally. It is suggested that the estimated dose equivalent rate obtained by the calculation is equivalent to the measured result, and it is shown that this method is effective for the radiation safety evaluation of low-level radioactive waste transport ships.
{"title":"Evaluation of Dose Equivalent Rates for a New Exclusive Ship “Seiei Maru” for Transporting Low-Level Radioactive Waste by Monte Carlo Particle Transport Calculation Using Variance Reduction Parameter Derived by FW-CADIS Method","authors":"M. Asami, S. Ohnishi, S. Kamada","doi":"10.3327/taesj.j19.024","DOIUrl":"https://doi.org/10.3327/taesj.j19.024","url":null,"abstract":"The forward weighted consistent adjoint driven importance sampling (FW-CADIS) method has been proposed as a method for obtaining variance reduction parameters to estimate flux or dose rate distribution over a wide area, or responses at multiple localized detectors for a particle transport calculation based on the stochastic Monte Carlo method with a reasonable and high accuracy. The method has been applied to estimating the dose equivalent rate around the Japanese exclusive ship, “Seiei Maru”, which transports low-level radioactive waste. The particle transport calculation was performed using a mesh tally on the entire surface of the hatch cover above low-level radioactive waste packages stacked in the cargo hold and point detector tallies at each measurement point. The statistical error is spatially uniform for the mesh tally and is reduced only around the vicinity of each point detector tally. It is suggested that the estimated dose equivalent rate obtained by the calculation is equivalent to the measured result, and it is shown that this method is effective for the radiation safety evaluation of low-level radioactive waste transport ships.","PeriodicalId":55893,"journal":{"name":"Transactions of the Atomic Energy Society of Japan","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2020-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"69437353","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Expert elicitation has traditionally been accepted in some countries as a way to quantify the uncertainty of radionuclide migration parameters in the safety assessment of radioactive waste disposal. However, expert elicitation has not yet been explicitly performed in the field of radioactive waste disposal in Japan. To discuss the applicability of expert elicitation in Japan, here we broadly review the histories and methodologies of expert elicitation in some papers and review in more detail case studies on the utilization of expert elicitation in the safety assessment of radioactive waste disposal in the US, UK, and Sweden. From the literature review, we suggest that it is valuable to adopt expert elicitation to quantify the uncertainty of parameters in Japan. In particular, the documentation of each elicitation step is critical to ensuring the traceability and transparency of expert elicitation. The documentation enables the regulator to evaluate whether the expert judgment including the elicitation process is adequate. Furthermore, we recommend providing not only an aggregated expert judgment for safety assessment but also the distribution of individual expert judgements. Individual expert judgments will be used for related analyses (e.g., sensitivity or uncertainty analyses), leading to increased confidence in the safety assessment.
{"title":"Quantification of Radionuclide Migration Parameters in Safety Assessment of Radioactive Waste Disposal: Review on the Use of Expert Elicitation","authors":"R. Nakabayashi, D. Sugiyama","doi":"10.3327/taesj.j19.004","DOIUrl":"https://doi.org/10.3327/taesj.j19.004","url":null,"abstract":"Expert elicitation has traditionally been accepted in some countries as a way to quantify the uncertainty of radionuclide migration parameters in the safety assessment of radioactive waste disposal. However, expert elicitation has not yet been explicitly performed in the field of radioactive waste disposal in Japan. To discuss the applicability of expert elicitation in Japan, here we broadly review the histories and methodologies of expert elicitation in some papers and review in more detail case studies on the utilization of expert elicitation in the safety assessment of radioactive waste disposal in the US, UK, and Sweden. From the literature review, we suggest that it is valuable to adopt expert elicitation to quantify the uncertainty of parameters in Japan. In particular, the documentation of each elicitation step is critical to ensuring the traceability and transparency of expert elicitation. The documentation enables the regulator to evaluate whether the expert judgment including the elicitation process is adequate. Furthermore, we recommend providing not only an aggregated expert judgment for safety assessment but also the distribution of individual expert judgements. Individual expert judgments will be used for related analyses (e.g., sensitivity or uncertainty analyses), leading to increased confidence in the safety assessment.","PeriodicalId":55893,"journal":{"name":"Transactions of the Atomic Energy Society of Japan","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2020-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"69437081","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The sodium – water reaction caused by failure of the steam generator tube of sodium-cooled fast re-actors causes the wastage phenomenon, which is erosive and corrosive. Self-wastage takes place in the early stage of the sodium – water reaction event when a very small amount of water / steam penetrates a microcrack. When self-wastage proceeds to the inside wall of the tube, the failed area and water leakage rate will increase, whereby the area affected by the sodium – water reaction will be likely to ex-tend. Thus, it is very important to clarify the self-wastage behavior for a locally affected region and detect water leakage in actual nuclear power plants. In this study, the authors performed self-wastage experiments under a high sodium temperature condition to evaluate the effects of the wastage form / geometry using two types of initial defect, i.e., the microfine pinhole and fatigue crack, and of the water leakage rate on the self-wastage rate. Taking into consideration the influence of crack type, we confirmed that the self-wastage rate did not strongly depend on the initial defect geometry. As a mechanism of the self-plug phenomenon, it is speculated that sodium oxide blocks and inhibits the progress of self-wastage. The dependence of the self-wastage rate on the initial sodium temperature was clearly observed, and a new self-wastage correlation was derived considering the initial sodium temperature.
{"title":"Experiments of Self-wastage Phenomena Elucidation in Steam Generator Tube of Sodium-cooled Fast Reactor","authors":"Ryota Umeda, K. Shimoyama, A. Kurihara","doi":"10.3327/taesj.j19.012","DOIUrl":"https://doi.org/10.3327/taesj.j19.012","url":null,"abstract":"The sodium – water reaction caused by failure of the steam generator tube of sodium-cooled fast re-actors causes the wastage phenomenon, which is erosive and corrosive. Self-wastage takes place in the early stage of the sodium – water reaction event when a very small amount of water / steam penetrates a microcrack. When self-wastage proceeds to the inside wall of the tube, the failed area and water leakage rate will increase, whereby the area affected by the sodium – water reaction will be likely to ex-tend. Thus, it is very important to clarify the self-wastage behavior for a locally affected region and detect water leakage in actual nuclear power plants. In this study, the authors performed self-wastage experiments under a high sodium temperature condition to evaluate the effects of the wastage form / geometry using two types of initial defect, i.e., the microfine pinhole and fatigue crack, and of the water leakage rate on the self-wastage rate. Taking into consideration the influence of crack type, we confirmed that the self-wastage rate did not strongly depend on the initial defect geometry. As a mechanism of the self-plug phenomenon, it is speculated that sodium oxide blocks and inhibits the progress of self-wastage. The dependence of the self-wastage rate on the initial sodium temperature was clearly observed, and a new self-wastage correlation was derived considering the initial sodium temperature.","PeriodicalId":55893,"journal":{"name":"Transactions of the Atomic Energy Society of Japan","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2020-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"69437141","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Soma KONO, Koichiro TAKAO and Tsuyoshi ARAI Shibaura Institute of Technology Graduate School of Engineering, 3–7–5 Toyosu, Koto-ku, Tokyo 135–8548, Japan Laboratory for Advanced Nuclear Energy, Institute of Innovative Research, Tokyo Institute of Technology, 2–12–1 N1–32 O-okayama, Meguro-ku, Tokyo 152–8550, Japan Shibaura Institute of Technology, 3–7–5 Toyosu, Koto-ku, Tokyo 135–8548, Japan (Received April 26, 2019; accepted in revised form November 7, 2019; published online May 12, 2020)
日本东京工业大学先进核能研究所创新研究所，2-12-1 n - 32 O-okayama, mekuroku, Tokyo, 152-8550，日本柴浦工业大学，3-7-5 Toyosu, kotoku, Tokyo, 135-8548(收到2019年4月26日;2019年11月7日以修改后的形式接受;2020年5月12日在线发布)
{"title":"Direct and Selective Electrodeposition of Palladium from Betainium Bis(trifluoromethanesulfonyl)imide Ionic Liquid Phase after Solvent Extraction together with Other Platinum Group Metals","authors":"Soma Kono, Koichiro Takao, T. Arai","doi":"10.3327/taesj.j19.001","DOIUrl":"https://doi.org/10.3327/taesj.j19.001","url":null,"abstract":"Soma KONO, Koichiro TAKAO and Tsuyoshi ARAI Shibaura Institute of Technology Graduate School of Engineering, 3–7–5 Toyosu, Koto-ku, Tokyo 135–8548, Japan Laboratory for Advanced Nuclear Energy, Institute of Innovative Research, Tokyo Institute of Technology, 2–12–1 N1–32 O-okayama, Meguro-ku, Tokyo 152–8550, Japan Shibaura Institute of Technology, 3–7–5 Toyosu, Koto-ku, Tokyo 135–8548, Japan (Received April 26, 2019; accepted in revised form November 7, 2019; published online May 12, 2020)","PeriodicalId":55893,"journal":{"name":"Transactions of the Atomic Energy Society of Japan","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2020-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"69437511","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Hidetsugu Nishikawa, M. Takeuchi, T. Kitagaki, Shuho Tsubota, Yuuichi Tooya, I. Sato
Hidetsugu NISHIKAWA, Masayuki TAKEUCHI, Toru KITAGAKI, Shuho TSUBOTA, Yuuichi TOOYA and Isamu SATO Tokyo City University, 1–28–1 Tamazutsumi, Setagaya, Tokyo 158–6557, Japan Japan Atomic Energy Agency, 4–33 Muramatsu, Tokai-mura, Naka-gun, Ibaraki 319–1194, Japan Mitsubishi Heavy Industries, Ltd., 1–1–1 Wadasaki-cho, Hyogo-ku, Kobe 652–8585, Japan Mitsubishi Heavy Industries, Ltd., 2–1–1 Shinhama, Arai-cho, Takasago-shi Hyogo 676–8686, Japan (Received June 6, 2019; accepted in revised form September 5, 2019; published online May 12, 2020)
{"title":"Feasibility of Disassembly of Fast Reactor Fuel Assembly Using Fiber Laser Cutting Technology","authors":"Hidetsugu Nishikawa, M. Takeuchi, T. Kitagaki, Shuho Tsubota, Yuuichi Tooya, I. Sato","doi":"10.3327/taesj.j19.006","DOIUrl":"https://doi.org/10.3327/taesj.j19.006","url":null,"abstract":"Hidetsugu NISHIKAWA, Masayuki TAKEUCHI, Toru KITAGAKI, Shuho TSUBOTA, Yuuichi TOOYA and Isamu SATO Tokyo City University, 1–28–1 Tamazutsumi, Setagaya, Tokyo 158–6557, Japan Japan Atomic Energy Agency, 4–33 Muramatsu, Tokai-mura, Naka-gun, Ibaraki 319–1194, Japan Mitsubishi Heavy Industries, Ltd., 1–1–1 Wadasaki-cho, Hyogo-ku, Kobe 652–8585, Japan Mitsubishi Heavy Industries, Ltd., 2–1–1 Shinhama, Arai-cho, Takasago-shi Hyogo 676–8686, Japan (Received June 6, 2019; accepted in revised form September 5, 2019; published online May 12, 2020)","PeriodicalId":55893,"journal":{"name":"Transactions of the Atomic Energy Society of Japan","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2020-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"69437096","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Light-water reactors (LWRs) are equipped with an emergency core cooling system (ECCS) that is designed to maintain the coolable geometry of the reactor core and finally minimize the release of radioactive fission products to the public and environment even in a loss-of-coolant accident (LOCA). Acceptance criteria for the ECCS of LWRs were determined to evaluate the safety function and performance in the design and to ensure a sufficient safety margin in the results of the evaluation. The latest revision of the criteria was made in 1981 in Japan, referring to the additional knowledge obtained after the previous revision. Fuel burnup has been extended by changing cladding materials, fuel design, etc., since the latest revision. Correspondingly, knowledge has been accumulated through studies on high-burnup fuel behavior under LOCA conditions to confirm the safety during the LOCA. This paper is a summary of the investigation and remaining issues on the applicability of the current Japanese ECCS acceptance criteria to high-burnup fuel, considering the history and basis of the current acceptance criteria. Results of the investigation conducted up to now reveal that the influence of burnup extension is small in terms of the cladding behavior of high-temperature oxidation and the fracture limit in quenching during the LOCA condition, and the current criteria are applicable even in the case of high-burnup fuel.
{"title":"Status of Investigation to Ensure Applicability of ECCS Acceptance Criteria to High-Burnup Fuel","authors":"M. Ozawa, M. Amaya","doi":"10.3327/taesj.j19.020","DOIUrl":"https://doi.org/10.3327/taesj.j19.020","url":null,"abstract":"Light-water reactors (LWRs) are equipped with an emergency core cooling system (ECCS) that is designed to maintain the coolable geometry of the reactor core and finally minimize the release of radioactive fission products to the public and environment even in a loss-of-coolant accident (LOCA). Acceptance criteria for the ECCS of LWRs were determined to evaluate the safety function and performance in the design and to ensure a sufficient safety margin in the results of the evaluation. The latest revision of the criteria was made in 1981 in Japan, referring to the additional knowledge obtained after the previous revision. Fuel burnup has been extended by changing cladding materials, fuel design, etc., since the latest revision. Correspondingly, knowledge has been accumulated through studies on high-burnup fuel behavior under LOCA conditions to confirm the safety during the LOCA. This paper is a summary of the investigation and remaining issues on the applicability of the current Japanese ECCS acceptance criteria to high-burnup fuel, considering the history and basis of the current acceptance criteria. Results of the investigation conducted up to now reveal that the influence of burnup extension is small in terms of the cladding behavior of high-temperature oxidation and the fracture limit in quenching during the LOCA condition, and the current criteria are applicable even in the case of high-burnup fuel.","PeriodicalId":55893,"journal":{"name":"Transactions of the Atomic Energy Society of Japan","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2020-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"69437310","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Understanding the behavior of melted volatile fission products （ FPs ） on the fuel contributes to the precise assessment of the release behaviour during a severe nuclear accident. A previous study revealed that liquid CsI shows abnormally high wettability with measured contact angles of almost zero degrees against the polycrystalline UO 2 solid surface. ［ K. Kurosaki et al., Sci. Rep. 7, Article number: 11449 （ 2017 ） . ］ . In this study, we focus on the melting behavior of CsIO 3 and revealed that liquid CsIO 3 also shows high wettability on the polycrystalline UO 2 solid surface. However, after melting, CsIO 3 decomposed and only Cs reacted with the polycrystalline UO 2 solid surface and I was only ab-sorbed on the solid surface. When the CsI had melted on the polycrystalline UO 2 solid surface, both Cs and I were able to penetrate inside the UO 2 pellets. In short, when Cs and I exist as CsIO 3 , Cs and I will be separately released during severe accidents. These findings suggest that the release mecha nisms of Cs and I could be strongly affected by the chemical species in the irradiated fuels.
了解熔化的挥发性裂变产物(FPs)在燃料上的行为有助于精确评估严重核事故中的释放行为。先前的一项研究表明，液态CsI与多晶uo2固体表面的接触角几乎为零，显示出异常高的润湿性。[K. Kurosaki et al.]， Sci。提案7，文号:11449(2017)。］ . 在本研究中，我们重点研究了csio3的熔融行为，发现液态csio3在多晶uo2固体表面也表现出很高的润湿性。而熔融后csio3分解，只有Cs与多晶uo2固体表面反应，I仅在固体表面被吸收。当CsI在多晶UO 2固体表面熔化时，Cs和I都能够穿透UO 2颗粒内部。简而言之，当Cs和我作为CsIO 3存在时，当发生严重事故时，Cs和我将被分开释放。这些发现表明，辐照燃料中的化学物质可能会强烈影响Cs和I的释放机制。
{"title":"Interaction of Liquid CsIO3 with a Polycrystalline UO2 Solid Surface","authors":"H. Ishii, Y. Ohishi, H. Muta, M. Uno, K. Kurosaki","doi":"10.3327/taesj.j19.017","DOIUrl":"https://doi.org/10.3327/taesj.j19.017","url":null,"abstract":"Understanding the behavior of melted volatile fission products （ FPs ） on the fuel contributes to the precise assessment of the release behaviour during a severe nuclear accident. A previous study revealed that liquid CsI shows abnormally high wettability with measured contact angles of almost zero degrees against the polycrystalline UO 2 solid surface. ［ K. Kurosaki et al., Sci. Rep. 7, Article number: 11449 （ 2017 ） . ］ . In this study, we focus on the melting behavior of CsIO 3 and revealed that liquid CsIO 3 also shows high wettability on the polycrystalline UO 2 solid surface. However, after melting, CsIO 3 decomposed and only Cs reacted with the polycrystalline UO 2 solid surface and I was only ab-sorbed on the solid surface. When the CsI had melted on the polycrystalline UO 2 solid surface, both Cs and I were able to penetrate inside the UO 2 pellets. In short, when Cs and I exist as CsIO 3 , Cs and I will be separately released during severe accidents. These findings suggest that the release mecha nisms of Cs and I could be strongly affected by the chemical species in the irradiated fuels.","PeriodicalId":55893,"journal":{"name":"Transactions of the Atomic Energy Society of Japan","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2020-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"69437297","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Best-estimate evaluation with detailed models of complicated phenomena that occur during accidents has been introduced into the safety evaluation of nuclear power plants. A system analysis code, which has physical models for the realistic prediction of events during accidents, is necessary for the safety evaluation. The analysis code should also be validated for individual phenomena and their combined behaviors at the actual plant scale during accidents. In this study, the system analysis code “AMAGI”, which is applicable to the evaluation of events from anticipated operational occurrences to design extension conditions, has been developed from its basic design. The thermal hydraulic model, heat conduction model, control model, and thermal power model were implemented into AMAGI as primary analytical functions. By conducting analyses of experiments with AMAGI, its fundamental models were validated.
{"title":"Analytical Functions and Development Status of the System Analysis Code for Nuclear Power Plants, AMAGI","authors":"J. Kaneko, Naofumi Tsukamoto","doi":"10.3327/taesj.j19.008","DOIUrl":"https://doi.org/10.3327/taesj.j19.008","url":null,"abstract":"Best-estimate evaluation with detailed models of complicated phenomena that occur during accidents has been introduced into the safety evaluation of nuclear power plants. A system analysis code, which has physical models for the realistic prediction of events during accidents, is necessary for the safety evaluation. The analysis code should also be validated for individual phenomena and their combined behaviors at the actual plant scale during accidents. In this study, the system analysis code “AMAGI”, which is applicable to the evaluation of events from anticipated operational occurrences to design extension conditions, has been developed from its basic design. The thermal hydraulic model, heat conduction model, control model, and thermal power model were implemented into AMAGI as primary analytical functions. By conducting analyses of experiments with AMAGI, its fundamental models were validated.","PeriodicalId":55893,"journal":{"name":"Transactions of the Atomic Energy Society of Japan","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2020-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"69437133","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Takayuki AOKI, Junya MASUKO, Atsuo IKEDA and Takayuki NEGISHI Center for Fundamental Research on Nuclear Decommissioning, Tohoku University, 6–6–01–2 Aoba, Aramaki, Aoba, Sendai 980–8579, Japan The Japan Atomic Power Company, 5-2–1 Ueno, Taito-ku, Tokyo 110–0005, Japan Nuclear Engineering and Services Company, 5-2–1 Ueno, Taito-ku, Tokyo 110–0005, Japan (Received May 14, 2019; accepted in revised form November 7, 2019; published online May 12, 2020)
{"title":"Safety Function Reliability Evaluation and Risk Management during Decommissioning of Accident-Damaged Nuclear Power Plant ～Based on Results of Confinement Function Reliability Evaluation～","authors":"T. Aoki, Junya Masuko, Atsuo Ikeda, T. Negishi","doi":"10.3327/taesj.j19.003","DOIUrl":"https://doi.org/10.3327/taesj.j19.003","url":null,"abstract":"Takayuki AOKI, Junya MASUKO, Atsuo IKEDA and Takayuki NEGISHI Center for Fundamental Research on Nuclear Decommissioning, Tohoku University, 6–6–01–2 Aoba, Aramaki, Aoba, Sendai 980–8579, Japan The Japan Atomic Power Company, 5-2–1 Ueno, Taito-ku, Tokyo 110–0005, Japan Nuclear Engineering and Services Company, 5-2–1 Ueno, Taito-ku, Tokyo 110–0005, Japan (Received May 14, 2019; accepted in revised form November 7, 2019; published online May 12, 2020)","PeriodicalId":55893,"journal":{"name":"Transactions of the Atomic Energy Society of Japan","volume":"30 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2020-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"69437062","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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