H. Hirayama, Katsumi Hayashi, K. Iwanaga, K. Kondo, Seishiro Suzuki
The gamma camera is a useful measuring device for grasping the distribution of contamination, but estimating the radioactivity of contamination is difficult. If the total energy absorption peak count rate of γ-rays can be measured with a pinhole gamma camera, we should be able to estimate the radioactivity. We can calculate it analytically by using the distance between the source and measurement points, and the peak count rate of direct γ-rays. The estimation method of Cs radioactivity using a pinhole gamma camera was studied using Hitachi’s gamma camera (HDG-E1500), which can be used to measure the total energy absorption peak count rate of γ-rays.
{"title":"Measurement of 137Cs Activity with Pinhole Gamma Camera","authors":"H. Hirayama, Katsumi Hayashi, K. Iwanaga, K. Kondo, Seishiro Suzuki","doi":"10.3327/taesj.j19.026","DOIUrl":"https://doi.org/10.3327/taesj.j19.026","url":null,"abstract":"The gamma camera is a useful measuring device for grasping the distribution of contamination, but estimating the radioactivity of contamination is difficult. If the total energy absorption peak count rate of γ-rays can be measured with a pinhole gamma camera, we should be able to estimate the radioactivity. We can calculate it analytically by using the distance between the source and measurement points, and the peak count rate of direct γ-rays. The estimation method of Cs radioactivity using a pinhole gamma camera was studied using Hitachi’s gamma camera (HDG-E1500), which can be used to measure the total energy absorption peak count rate of γ-rays.","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":"69437371","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}
Y. Sugita, N. Taniguchi, Hitoshi Makino, Shinichiro Kanamaru, T. Okumura
A series of structural analyses of disposal containers for the direct disposal of spent fuel was car-ried out to provide preliminary estimates of the required thickness for adequate pressure resistance of the disposal container. Disposal containers were designed to contain either 2, 3 or 4 spent fuel assemblies in linear, triangular or square arrangements, respectively. The required pressure resistance thickness was evaluated using the separation distance of the housing space for each spent fuel assembly as a key model parameter to obtain the required thickness of the body as well as the lid of the disposal container. In this work, we also provide additional technical knowledge on the structural analysis of disposal containers, such as the validity of the setting of the stress evaluation line and the effect of the model length on the analysis. These can then be referred to and used again in the future as a basis for conducting similar evaluations under different conditions or proceeding with more detailed evaluations.
{"title":"Pressure Resistance Thickness of Disposal Containers for Spent Fuel Direct Disposal","authors":"Y. Sugita, N. Taniguchi, Hitoshi Makino, Shinichiro Kanamaru, T. Okumura","doi":"10.3327/taesj.j19.007","DOIUrl":"https://doi.org/10.3327/taesj.j19.007","url":null,"abstract":"A series of structural analyses of disposal containers for the direct disposal of spent fuel was car-ried out to provide preliminary estimates of the required thickness for adequate pressure resistance of the disposal container. Disposal containers were designed to contain either 2, 3 or 4 spent fuel assemblies in linear, triangular or square arrangements, respectively. The required pressure resistance thickness was evaluated using the separation distance of the housing space for each spent fuel assembly as a key model parameter to obtain the required thickness of the body as well as the lid of the disposal container. In this work, we also provide additional technical knowledge on the structural analysis of disposal containers, such as the validity of the setting of the stress evaluation line and the effect of the model length on the analysis. These can then be referred to and used again in the future as a basis for conducting similar evaluations under different conditions or proceeding with more detailed evaluations.","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":"69437127","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}
Integrity confirmation for buildings against collisions of projectiles has been conducted to evalu ate their behavior upon collision between a projectile with a simple shape and a wall using empirical formulas. However, there is a possibility that structures with a complex shape such as a stack may collide with a reactor building. However, there have been few studies of collisions between structures with a complex shape and buildings. In this study, it is necessary to set the physical properties of reinforced concrete to carry out such an evaluation; however, the physical properties depend on the number of reinforcing bars. For this reason, the physical properties of reinforced concrete are set so as to fit conventional empirical formulas recommended. Impact evaluation was carried out using a reac tor building and a stack with a realistic shape and physical properties.
{"title":"Evaluation of Impact with High Temperature Engineering Test Reactor Using Realistic Model of Stack and Reactor Building","authors":"M. Ono, Y. Fujiwara, Tetsuro Matsumoto, K. Iigaki","doi":"10.3327/taesj.j19.005","DOIUrl":"https://doi.org/10.3327/taesj.j19.005","url":null,"abstract":"Integrity confirmation for buildings against collisions of projectiles has been conducted to evalu ate their behavior upon collision between a projectile with a simple shape and a wall using empirical formulas. However, there is a possibility that structures with a complex shape such as a stack may collide with a reactor building. However, there have been few studies of collisions between structures with a complex shape and buildings. In this study, it is necessary to set the physical properties of reinforced concrete to carry out such an evaluation; however, the physical properties depend on the number of reinforcing bars. For this reason, the physical properties of reinforced concrete are set so as to fit conventional empirical formulas recommended. Impact evaluation was carried out using a reac tor building and a stack with a realistic shape and physical properties.","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":"69437088","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}
Various types of flame retardant cable are used in nuclear power plants. The cable insulation and sheath materials are generally categorized into two types: thermosetting （ TS ） and thermoplastic （ TP ） materials. Domestic cable fire damage criteria are set for each material type on the basis of the US guidelines, although they do not clearly define which type a TS / TP composite cable falls into. In this study, cable burning tests were conducted for TP and TS / TP composite cables used in domestic nuclear power plants to evaluate the validity of applying the US cable damage criteria to domestic cables. For varions cable damage scenarios, the tests measured the temperature / time dependence of insulation resistance between conductors and between cable and trays. The results of these tests justified the applicability of the US damage criteria. In addition, it was confirmed that the relationship be tween cable temperature and insulation resistance can be expressed by an Arrhenius plot. These find ings are expected to be utilized for the Safety Improvement Evaluation periodically conducted by nu clear facility operators.
{"title":"Damage Assessment in Case of Cable Fire in Nuclear Facilities Considering Damage Criteria for Each Material Type","authors":"Susumu Tsuchino, A. Matsuda","doi":"10.3327/taesj.j19.021","DOIUrl":"https://doi.org/10.3327/taesj.j19.021","url":null,"abstract":"Various types of flame retardant cable are used in nuclear power plants. The cable insulation and sheath materials are generally categorized into two types: thermosetting （ TS ） and thermoplastic （ TP ） materials. Domestic cable fire damage criteria are set for each material type on the basis of the US guidelines, although they do not clearly define which type a TS / TP composite cable falls into. In this study, cable burning tests were conducted for TP and TS / TP composite cables used in domestic nuclear power plants to evaluate the validity of applying the US cable damage criteria to domestic cables. For varions cable damage scenarios, the tests measured the temperature / time dependence of insulation resistance between conductors and between cable and trays. The results of these tests justified the applicability of the US damage criteria. In addition, it was confirmed that the relationship be tween cable temperature and insulation resistance can be expressed by an Arrhenius plot. These find ings are expected to be utilized for the Safety Improvement Evaluation periodically conducted by nu clear facility operators.","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":"69437320","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}
J. Aihara, Masaki Honda, S. Ueta, H. Ogawa, Koichiro Ohira, Y. Tachibana
Jun AIHARA, Masaki HONDA, Shohei UETA, Hiroaki OGAWA, Koichi OHIRA and Yukio TACHIBANA Japan Atomic Energy Agency, Oarai Research and Development Center, 4002 Narita-cho, Oarai-machi, Higashiibaraki-gun, Ibaraki 311–1393, Japan Nuclear Fuel Industries, Ltd., 41 Muramatsu, Tokai-mura, Naka-gun, Ibaraki 319–1196, Japan Japan Atomic Energy Agency, Tokai Research and Development Center, 2–4 Shirane, Shirakata, Tokai-mura, Naka-gun, Ibaraki 319–1195, Japan (Received March 9, 2018; accepted in revised form October 29, 2018; published online January 17, 2019)
{"title":"Development of Fabrication Technology for Oxidation-Resistant Fuel Elements for High-Temperature Gas-Cooled Reactors","authors":"J. Aihara, Masaki Honda, S. Ueta, H. Ogawa, Koichiro Ohira, Y. Tachibana","doi":"10.3327/TAESJ.J17.027","DOIUrl":"https://doi.org/10.3327/TAESJ.J17.027","url":null,"abstract":"Jun AIHARA, Masaki HONDA, Shohei UETA, Hiroaki OGAWA, Koichi OHIRA and Yukio TACHIBANA Japan Atomic Energy Agency, Oarai Research and Development Center, 4002 Narita-cho, Oarai-machi, Higashiibaraki-gun, Ibaraki 311–1393, Japan Nuclear Fuel Industries, Ltd., 41 Muramatsu, Tokai-mura, Naka-gun, Ibaraki 319–1196, Japan Japan Atomic Energy Agency, Tokai Research and Development Center, 2–4 Shirane, Shirakata, Tokai-mura, Naka-gun, Ibaraki 319–1195, Japan (Received March 9, 2018; accepted in revised form October 29, 2018; published online January 17, 2019)","PeriodicalId":55893,"journal":{"name":"Transactions of the Atomic Energy Society of Japan","volume":" ","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"44509972","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}
Daisuke OKADA, Hiroto ISHII, Yuji OHISHI, Hiroaki MUTA and Ken KUROSAKI Graduate School of Engineering, Osaka University, 2–1 Yamadaoka, Suita-shi, Osaka 565–0871, Japan Research Institute of Nuclear Engineering, University of Fukui, 1–3–33 Kanawacho, Tsuruga-shi, Fukui 914–0055, Japan JST, PRESTO, 4–1–8 Honcho, Kawaguchi-shi, Saitama 332–0012, Japan (Received July 18, 2018; accepted in revised form November 1, 2018; published online January 17, 2019)
{"title":"Thermal and Mechanical Properties of Fe2Zr","authors":"Daisuke Okada","doi":"10.3327/TAESJ.J18.011","DOIUrl":"https://doi.org/10.3327/TAESJ.J18.011","url":null,"abstract":"Daisuke OKADA, Hiroto ISHII, Yuji OHISHI, Hiroaki MUTA and Ken KUROSAKI Graduate School of Engineering, Osaka University, 2–1 Yamadaoka, Suita-shi, Osaka 565–0871, Japan Research Institute of Nuclear Engineering, University of Fukui, 1–3–33 Kanawacho, Tsuruga-shi, Fukui 914–0055, Japan JST, PRESTO, 4–1–8 Honcho, Kawaguchi-shi, Saitama 332–0012, Japan (Received July 18, 2018; accepted in revised form November 1, 2018; published online January 17, 2019)","PeriodicalId":55893,"journal":{"name":"Transactions of the Atomic Energy Society of Japan","volume":" ","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"48519761","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}
H. Ishii, Y. Murakami, Y. Ohishi, H. Muta, M. Uno, K. Kurosaki
The wettability of liquid-state volatile fission products （ FPs ） for solid-state fuels is important for the evaluation of FP release behavior during severe accidents. Previously, we have reported that liquid CsI exhibits extremely high wettability with a contact angle of virtually 0 ° on a polycrystalline UO 2 surface ［ K. Kurosaki et al., Sci. Rep. 7, Article number: 11449 （ 2017 ）］ and a single-crystalline yttria-stabilized zirconia （ YSZ ） surface ［ H. Ishii et al., J. Nucl. Sci. Technol. 55, No. 8, 838 – 842 （ 2018 ）］ . Here, based on the previous studies, we investigate the wettability of liquid cesium halides （ CsCl, CsBr ） on single-crystalline YSZ, TiO 2 , and MgO. We observe the high wettability of liquid cesium halides on single-crystalline YSZ and TiO 2 with contact angles of nearly 0 ° in all cases. However, liquid cesium halides exhibit completely different wettability on single-crystalline MgO, where the contact angles are measured to be 50 ° , 44 ° , and 25 ° for CsCl, CsBr, and CsI, respectively. Similarly to UO 2 , YSZ and TiO 2 are non-stoichiometric compounds containing oxygen defects, while MgO is a stoichiometric line compound. Thus, it is assumed that the oxygen defects play a role in the high wettability of liquid CsI on solid UO 2 .
{"title":"Wettability of Liquid Cesium Halides on Oxide Single Crystals","authors":"H. Ishii, Y. Murakami, Y. Ohishi, H. Muta, M. Uno, K. Kurosaki","doi":"10.3327/TAESJ.J18.002","DOIUrl":"https://doi.org/10.3327/TAESJ.J18.002","url":null,"abstract":"The wettability of liquid-state volatile fission products （ FPs ） for solid-state fuels is important for the evaluation of FP release behavior during severe accidents. Previously, we have reported that liquid CsI exhibits extremely high wettability with a contact angle of virtually 0 ° on a polycrystalline UO 2 surface ［ K. Kurosaki et al., Sci. Rep. 7, Article number: 11449 （ 2017 ）］ and a single-crystalline yttria-stabilized zirconia （ YSZ ） surface ［ H. Ishii et al., J. Nucl. Sci. Technol. 55, No. 8, 838 – 842 （ 2018 ）］ . Here, based on the previous studies, we investigate the wettability of liquid cesium halides （ CsCl, CsBr ） on single-crystalline YSZ, TiO 2 , and MgO. We observe the high wettability of liquid cesium halides on single-crystalline YSZ and TiO 2 with contact angles of nearly 0 ° in all cases. However, liquid cesium halides exhibit completely different wettability on single-crystalline MgO, where the contact angles are measured to be 50 ° , 44 ° , and 25 ° for CsCl, CsBr, and CsI, respectively. Similarly to UO 2 , YSZ and TiO 2 are non-stoichiometric compounds containing oxygen defects, while MgO is a stoichiometric line compound. Thus, it is assumed that the oxygen defects play a role in the high wettability of liquid CsI on solid UO 2 .","PeriodicalId":55893,"journal":{"name":"Transactions of the Atomic Energy Society of Japan","volume":" ","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"47029184","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}
K. Yoshimura, T. Sugii, T. Sano, E. Ishii, Takumi Kitagawa
We developed an examination method that uses computational fluid dynamics (CFD) to investigate the effects of a complex pipe geometry on flow fields. Two kinds of pipe model with different geometries are simulated to test the developed method. The simulation models were split into several computational regions to reduce the computation time. The simulation results showed that the fluctuation of the flow rate depended on the pipe geometry, which qualitatively agreed well with the experimental results. The simulation results of one of the two models showed a swirling flow around the orifice with large fluctuations of the flow rate. It was found that the swirling flow caused velocity fluctuations in the recirculation zone around the tap positions, which resulted in the large fluctuations of the flow rate. We also investigated the mechanisms generating the swirling flow. The simulation results showed that the high velocity of the flow along the wall was caused by the valve and the bend pipe. The high-velocity flow then moves along the pipe wall of the tee, which causes the flow to swirl. These results show that the developed method can be used to evaluate the flow fields in piping systems.
{"title":"Examination Method for Flow Fields in Piping Systems of Nuclear Power Plants Using Computational Fluid Dynamics","authors":"K. Yoshimura, T. Sugii, T. Sano, E. Ishii, Takumi Kitagawa","doi":"10.3327/TAESJ.J18.004","DOIUrl":"https://doi.org/10.3327/TAESJ.J18.004","url":null,"abstract":"We developed an examination method that uses computational fluid dynamics (CFD) to investigate the effects of a complex pipe geometry on flow fields. Two kinds of pipe model with different geometries are simulated to test the developed method. The simulation models were split into several computational regions to reduce the computation time. The simulation results showed that the fluctuation of the flow rate depended on the pipe geometry, which qualitatively agreed well with the experimental results. The simulation results of one of the two models showed a swirling flow around the orifice with large fluctuations of the flow rate. It was found that the swirling flow caused velocity fluctuations in the recirculation zone around the tap positions, which resulted in the large fluctuations of the flow rate. We also investigated the mechanisms generating the swirling flow. The simulation results showed that the high velocity of the flow along the wall was caused by the valve and the bend pipe. The high-velocity flow then moves along the pipe wall of the tee, which causes the flow to swirl. These results show that the developed method can be used to evaluate the flow fields in piping systems.","PeriodicalId":55893,"journal":{"name":"Transactions of the Atomic Energy Society of Japan","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"42100306","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}
R. Nakabayashi, Yasutaka Watanabe, D. Minato, D. Sugiyama
Ryo NAKABAYASHI, Yasutaka WATANABE, Daisuke MINATO and Daisuke SUGIYAMA Radiation Safety Research Center, Nuclear Technology Research Laboratory, Central Research Institute of Electric Power Industry, 2–11–1 Iwadokita, Komae-shi, Tokyo 201–8511, Japan Nuclear Fuel Cycle Backend Research Center, Civil Engineering Research Laboratory, Central Research Institute of Electric Power Industry, 1646 Abiko, Abiko-shi, Chiba 270–1194, Japan (Received May 7, 2018; accepted in revised form August 17, 2018; published online December 13, 2018)
{"title":"Study on Practical Application of Optimization Method for Radioactive Waste Disposal Facility Design Based on Probabilistic Approach: Discussion on Procedure for Setting Radionuclide Migration Parameters","authors":"R. Nakabayashi, Yasutaka Watanabe, D. Minato, D. Sugiyama","doi":"10.3327/TAESJ.J18.003","DOIUrl":"https://doi.org/10.3327/TAESJ.J18.003","url":null,"abstract":"Ryo NAKABAYASHI, Yasutaka WATANABE, Daisuke MINATO and Daisuke SUGIYAMA Radiation Safety Research Center, Nuclear Technology Research Laboratory, Central Research Institute of Electric Power Industry, 2–11–1 Iwadokita, Komae-shi, Tokyo 201–8511, Japan Nuclear Fuel Cycle Backend Research Center, Civil Engineering Research Laboratory, Central Research Institute of Electric Power Industry, 1646 Abiko, Abiko-shi, Chiba 270–1194, Japan (Received May 7, 2018; accepted in revised form August 17, 2018; published online December 13, 2018)","PeriodicalId":55893,"journal":{"name":"Transactions of the Atomic Energy Society of Japan","volume":" ","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://sci-hub-pdf.com/10.3327/TAESJ.J18.003","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"43396483","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}
Tadashi Ikehara, M. Sasagawa, Sho Takano, Teppei Yamana, Naoki Yanagisawa
MCNP whole-core reference solutions with pin-by-pin resolution were generated to numerically complement the operating and/or experimentally measured data of an Advanced BWR MOX core loaded with UO2 to 100% MOX fuels for the verification and validation (V&V) of BWR core design codes. To this end, (1) the degree of preciseness in geometrical and material modeling under the allowable storage limitations of MCNP was explored, (2) the correctness of predicted neutronics phenomena inside BWR cores by MCNP physics was investigated and the necessary enhancement was made for MCNP, (3) the smallness of uncertainties of results in MCNP due to its stochastic treatment and the nuclear data employed was ensured, (4) the representativeness of BWR core characteristics obtained from MCNP reference solutions was qualified. According to the results, the MCNP-generated reference solutions are applicable to validating the neutronics calculation performance of BWR core design codes as an alternative to measured data. In addition, extraction of the detailed neutronics quantities such as fuel-pin-wise flux enables us to verify the physics modules of the design codes such as the neutron flux solver. The MCNP whole-core reference solutions proved to be applicable in performing MCNP-based BWR design code V&V.
{"title":"Generation of MCNP Whole-Core Reference Solutions for Advanced BWR MOX Core and Its Application to Neutronics Calculation Performance V&V of the BWR Nuclear Design Code","authors":"Tadashi Ikehara, M. Sasagawa, Sho Takano, Teppei Yamana, Naoki Yanagisawa","doi":"10.3327/TAESJ.J17.015","DOIUrl":"https://doi.org/10.3327/TAESJ.J17.015","url":null,"abstract":"MCNP whole-core reference solutions with pin-by-pin resolution were generated to numerically complement the operating and/or experimentally measured data of an Advanced BWR MOX core loaded with UO2 to 100% MOX fuels for the verification and validation (V&V) of BWR core design codes. To this end, (1) the degree of preciseness in geometrical and material modeling under the allowable storage limitations of MCNP was explored, (2) the correctness of predicted neutronics phenomena inside BWR cores by MCNP physics was investigated and the necessary enhancement was made for MCNP, (3) the smallness of uncertainties of results in MCNP due to its stochastic treatment and the nuclear data employed was ensured, (4) the representativeness of BWR core characteristics obtained from MCNP reference solutions was qualified. According to the results, the MCNP-generated reference solutions are applicable to validating the neutronics calculation performance of BWR core design codes as an alternative to measured data. In addition, extraction of the detailed neutronics quantities such as fuel-pin-wise flux enables us to verify the physics modules of the design codes such as the neutron flux solver. The MCNP whole-core reference solutions proved to be applicable in performing MCNP-based BWR design code V&V.","PeriodicalId":55893,"journal":{"name":"Transactions of the Atomic Energy Society of Japan","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"69437173","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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