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JAEA Reports

Analysis of behavior of Ru with nitrogen oxide chemical behavior in accident of evaporation to dryness by boiling of reprocessed high level liquid waste

Yoshida, Kazuo; Tamaki, Hitoshi; Hiyama, Mina*

JAEA-Research 2021-005, 25 Pages, 2021/08

JAEA-Research-2021-005.pdf:2.91MB

An accident of evaporation to dryness by boiling of high level liquid waste (HLLW) is postulated as one of the severe accidents caused by the loss of cooling function at a fuel reprocessing plant. In this case, volatile radioactive materials, such as ruthenium (Ru) are released from the tanks with water and nitric-acid mixed vapor into atmosphere. Accurate quantitative estimation of released Ru is one of the important issues for risk assessment of those facilities. To resolve this issue, an empirical correlation equation of Ru mass transfer coefficient across the vapor-liquid surface, which can be useful for quantitative simulation of Ru mitigating behavior, has been obtained from data analyses of small-scale experiments conducted to clarify gaseous Ru migrating behavior under steam-condensing condition. A simulation study has been also carried out with a hypothetical typical facility building successfully to demonstrate the feasibility of quantitative estimation of amount of Ru migrating in the facility using the obtained correlation equation implemented in SCHERN computer code which simulates chemical behaviors of nitrogen oxide based on the condition also simulated thermal-hydraulic computer code.

JAEA Reports

SCHERN-V2: Technical guide of computer program for chemical behavior in accident of evaporation to dryness by boiling of reprocessed high level liquid waste in Fuel Reprocessing Facilities

Yoshida, Kazuo; Tamaki, Hitoshi; Hiyama, Mina*

JAEA-Data/Code 2021-008, 35 Pages, 2021/08

JAEA-Data-Code-2021-008.pdf:3.68MB

An accident of evaporation to dryness by boiling of high level liquid waste (HLLW) is postulated as one of the severe accidents caused by the loss of cooling function at a fuel reprocessing plant. In this case, volatile radioactive materials, such as ruthenium (Ru) are released from the tanks with water and nitric-acid mixed vapor into atmosphere. In addition to this, nitrogen oxides (NO$$_{rm x}$$) are also released formed by the thermal decomposition of metal nitrates of fission products (FP) in HLLW. It has been observed experimentally that NOx affects to the migration behavior of Ru at the anticipated atmosphere condition in cells and/or compartments of the facility building. Chemical reactions of NO$$_{rm x}$$ with water and nitric acid are also recognized as the complex phenomena to undergo simultaneously in the vapor and liquid phases. The analysis program, SCHERN has been under developed to simulate chemical behavior including Ru coupled with the thermo-hydraulic condition in the flow paths in the facility building. This technical guide for SCHERN-V2 presents the overview of covered accident, analytical models including newly developed models, differential equations for numerical solution, and user instructions.

Journal Articles

Overview and outcomes of the OECD/NEA benchmark study of the accident at the Fukushima Daiichi NPS (BSAF), Phase 2; Results of severe accident analyses for unit 3

Lind, T.*; Pellegrini, M.*; Herranz, L. E.*; Sonnenkalb, M.*; Nishi, Yoshihisa*; Tamaki, Hitoshi; Cousin, F.*; Fernandez Moguel, L.*; Andrews, N.*; Sevon, T.*

Nuclear Engineering and Design, 376, p.111138_1 - 111138_12, 2021/05

 Times Cited Count:3 Percentile:81.22(Nuclear Science & Technology)

This is the third part of the three part paper describing the accidents at the FDNPS as analyzed in the Phase 2 of the OECD/NEA project "Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Plant" (BSAF). In this paper, we describe the accident progression in unit 3. In the BSAF project, eight organizations from five countries analyzed severe accident scenarios for Unit 3 at the Fukushima Daiichi site using different severe accident codes. The present paper for Unit 3 describes the findings of the comparison of the participants' results against each other and against plant data, the evaluation of the accident progression and the final status inside the reactors. Special focus is on the status of the reactor pressure vessel, melt release and fission product release and transport. Unit 3 specific aspects, e.g., the complicated accident progression following repeated containment venting actuations and attempts at coolant injection at the time of the major core degradation, are highlighted and points of consensus as well as remaining uncertainties and data needs will be summarized. FP transport is analyzed, and the calculation results are compared with dose rate measurements in the containment. The release of I-131 and Cs-137 to the environment is compared with analysis conducted by using WSPEEDI code.

Journal Articles

Overview and outcomes of the OECD/NEA benchmark study of the accident at the Fukushima Daiichi NPS (BSAF) Phase 2; Results of severe accident analyses for Unit 1

Herranz, L. E.*; Pellegrini, M.*; Lind, T.*; Sonnenkalb, M.*; Godin-Jacqmin, L.*; L$'o$pez, C.*; Dolganov, K.*; Cousin, F.*; Tamaki, Hitoshi; Kim, T. W.*; et al.

Nuclear Engineering and Design, 369, p.110849_1 - 110849_7, 2020/12

 Times Cited Count:5 Percentile:76.41(Nuclear Science & Technology)

Phase 2 of the OECD/NEA Project "Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Plant (BSAF)" was established in mid-2015. The objectives have been similar to Phase 1 of the project but with an extended analysis period of 3 weeks, a major focus on FP behaviour and releases to the environment and the comparison to various data and results of backwards calculations of the source term. Nine organizations of six countries submitted results of their calculated severe accident scenarios for Unit 1 at the 1F site using different severe accident codes. This paper describes the findings of the comparison of the participants results for Unit1 against each other and against plant data, the evaluation of the accident progression and the final status inside the reactors. Special focus is on RPV status, melt release and FP behaviour and release. Unit specific aspects will be highlighted and points of consensus as well as remaining uncertainties and data needs will be summarised.

Journal Articles

Overview and outcomes of the OECD/NEA benchmark study of the accident at the Fukushima Daiichi NPS (BSAF), Phase 2; Results of severe accident analyses for Unit 2

Sonnenkalb, M.*; Pellegrini, M.*; Herranz, L. E.*; Lind, T.*; Morreale, A. C.*; Kanda, Kenichi*; Tamaki, Hitoshi; Kim, S. I.*; Cousin, F.*; Fernandez Moguel, L.*; et al.

Nuclear Engineering and Design, 369, p.110840_1 - 110840_10, 2020/12

 Times Cited Count:6 Percentile:84.6(Nuclear Science & Technology)

This is the second paper in a series of 3 in which results of severe accident analyses for Unit 2 of Fukushima Daiichi are presented, gained in Phase 2 of the OECD/NEA project "Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Plant (BSAF)". Nine organizations of six countries submitted results of their calculated severe accident scenarios for Unit 2 of Fukushima Daiichi using different severe accident codes. The present paper describes the findings of the comparison of the participants' results for Unit 2 against each other and against plant data, the evaluation of the accident progression and the final status inside the reactors. Special focus is on reactor pressure vessel status, melt release and fission product behavior and release. Unit 2 specific aspects will be highlighted and points of consensus as well as remaining uncertainties and data needs will be summarized.

Journal Articles

Dynamic PRA of flooding-initiated accident scenarios using THALES2-RAPID

Kubo, Kotaro; Zheng, X.; Tanaka, Yoichi; Tamaki, Hitoshi; Sugiyama, Tomoyuki; Jang, S.*; Takata, Takashi*; Yamaguchi, Akira*

Proceedings of 30th European Safety and Reliability Conference and 15th Probabilistic Safety Assessment and Management Conference (ESREL 2020 and PSAM-15) (Internet), p.2279 - 2286, 2020/11

Probabilistic risk assessment (PRA) is one of the methods used to assess the risks associated with large and complex systems. When the risk of an external event is evaluated using conventional PRA, a particular limitation is the difficulty in considering the timing at which nuclear power plant structures, systems, and components fail. To overcome this limitation, we coupled thermal-hydraulic and external-event simulations using Risk Assessment with Plant Interactive Dynamics (RAPID). Internal flooding was chosen as the representative external event, and a pressurized water reactor plant model was used. Equations based on Bernoulli's theorem were applied to flooding propagation in the turbine building. In the analysis, uncertainties were taken into account, including the flow rate of the flood water source and the failure criteria for the mitigation systems. In terms of recovery action, isolation of the flood water source by the operator and drainage using a pump were modeled based on several assumptions. The results indicate that the isolation action became more effective when combined with drainage.

Journal Articles

Enhancement of the treatment of system interactions in a dynamic PRA tool

Tanaka, Yoichi; Tamaki, Hitoshi; Zheng, X.; Sugiyama, Tomoyuki

Proceedings of 30th European Safety and Reliability Conference and 15th Probabilistic Safety Assessment and Management Conference (ESREL 2020 and PSAM-15) (Internet), p.2195 - 2201, 2020/11

Journal Articles

A Comparative study of sampling techniques for dynamic probabilistic risk assessment of nuclear power plants

Kubo, Kotaro; Zheng, X.; Tanaka, Yoichi; Tamaki, Hitoshi; Sugiyama, Tomoyuki; Jang, S.*; Takata, Takashi*; Yamaguchi, Akira*

Proceedings of Joint International Conference on Supercomputing in Nuclear Applications + Monte Carlo 2020 (SNA + MC 2020), p.308 - 315, 2020/10

Dynamic probabilistic risk assessment (PRA) is a method for improving the realism and completeness of conventional PRA. However, enormous calculation costs are incurred by these improvements. One solution is to select an appropriate sampling method. In this paper, we applied the Monte Carlo, Latin hypercube, grid-point, and quasi-Monte Carlo sampling methods to the dynamic PRA of a simplified accident sequence and compared the results for each method. Quasi-Monte Carlo sampling was found to be the most effective method in this case.

Journal Articles

Analysis for the accident at unit 1 of the Fukushima Daiichi NPS with THALES2/KICHE code in BSAF2 project

Tamaki, Hitoshi; Ishikawa, Jun; Sugiyama, Tomoyuki; Maruyama, Yu

Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.72 - 82, 2019/08

JAEA participated in the OECD/NEA BSAF2 project with our integrated severe accident analysis code, THALES2/KICHE, in order to analyze and discuss the accident progression and source term of the accident at the Fukushima Daiichi NPS. One of important characteristics of THALES2/KICHE code is that it has the capability of predicting iodine chemistry based on reaction kinetics in the aqueous phase. JAEA performed the three week analysis for the accident at unit 1 on the basis of the boundary conditions and assumptions proposed by the BSAF2 project and our own assumptions. In addition to the failure of the drywell, it was assumed in the present analysis that continuous leakage occurred through the containment venting line due to incomplete closing of valves in the line. The releases of fission products, especially for iodine and cesium, within three weeks after the earthquake were estimated to be approximately 6% and 1% of the initial inventory, respectively.

Journal Articles

Analysis for the accident at unit 2 of the Fukushima Daiichi NPS with THALES2/KICHE code in BSAF2 project

Tamaki, Hitoshi; Ishikawa, Jun; Sugiyama, Tomoyuki; Maruyama, Yu

Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.100 - 111, 2019/08

JAEA participated in the OECD/NEA BSAF2 project with our integrated severe accident analysis code, THALES2/KICHE, in order to analyze and discuss the accident progression and source term of the accident at the Fukushima Daiichi NPS. One of important characteristics of THALES2/KICHE code is that it has the capability of predicting iodine chemistry based on reaction kinetics in the aqueous phase. JAEA performed the three week analysis for the accident at unit 2 on the basis of the boundary conditions and assumptions proposed by the BSAF2 project and our own assumptions. One of focusing points in the BSAF2 project was the trend of measured data of reactor vessel from 20:00 March 14 to 02:00 March 15. An assumption was made that the lower part of the suppression chamber failed to form a water leakage path. The released iodine and cesium within three weeks after the earthquake were predicted to be approximately 3% and 0.1% of the initial inventory, respectively.

Journal Articles

Analysis for the accident at Unit 3 of the Fukushima Daiichi NPS with THALES2/KICHE Code in BSAF2 project

Ishikawa, Jun; Tamaki, Hitoshi; Sugiyama, Tomoyuki; Maruyama, Yu

Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.536 - 547, 2019/08

JAEA Reports

SCHERN: Analysis program for chemical behavior of nitrogen oxide in accident of evaporation to dryness by boiling of reprocessed high level liquid waste in Fuel Reprocessing Facilities

Hiyama, Mina*; Tamaki, Hitoshi; Yoshida, Kazuo

JAEA-Data/Code 2019-006, 17 Pages, 2019/07

JAEA-Data-Code-2019-006.pdf:1.84MB

An accident of evaporation to dryness by boiling of high level liquid waste (HLLW) is postulated as one of the severe accidents caused by the loss of cooling function at a fuel reprocessing plant. In this case, volatile radioactive materials, such as ruthenium (Ru) are released from the tanks with water and nitric-acid mixed vapor into atmosphere. In addition to this, nitrogen oxides (NOx) are also released formed by the thermal decomposition of metal nitrates of fission products (FP) in HLLW. It has been observed experimentally that NOx affects strongly to the transport behavior of Ru at the anticipated atmosphere condition in cells and/or compartments of the facility building. Chemical reactions of NOx with water and nitric acid are also recognized as the complex phenomena to undergo simultaneously in the vapor and liquid phases. An analysis program has been developed to simulate chemical reaction coupled with the thermo-hydraulic condition in the flow paths in the facility building.

Journal Articles

Analysis of chemical behavior of nitrogen oxide formed by thermal decomposition of FP nitrates in accident of evaporation to dryness by boiling of reprocessed high-level liquid waste

Yoshida, Kazuo; Tamaki, Hitoshi; Yoshida, Naoki; Yoshida, Ryoichiro; Amano, Yuki; Abe, Hitoshi

Nihon Genshiryoku Gakkai Wabun Rombunshi, 18(2), p.69 - 80, 2019/06

An accident of evaporation to dryness by boiling of high level liquid waste (HLLW) is postulated as one of the severe accidents at a fuel reprocessing plant. In this case, volatile radioactive materials, such as ruthenium (Ru) are released from the tanks with water and nitric-acid mixed vapor into atmosphere. In addition to this, nitrogen oxides are also released formed by the thermal decomposition of metal nitrates of fission products (FP) in HLLW. It has been observed experimentally that nitrogen oxide affects strongly to the transport behavior of Ru. Chemical reactions of nitrogen oxide with water and nitric acid are also recognized as the complex phenomena to undergo simultaneously in the vapor and liquid phases. An analysis method has been developed with coupling two types of computer codes to simulate not only thermo-hydraulic behavior but also chemical reactions in the flow paths of carrier gases. A simulation study has been also carried out with a typical facility building.

Journal Articles

Sensitivity analysis of source term in the accident of Fukushima Dai-ichi Nuclear Power Station Unit 1 using THALES2/KICHE

Tamaki, Hitoshi; Ishikawa, Jun; Sugiyama, Tomoyuki; Maruyama, Yu

Proceedings of Asian Symposium on Risk Assessment and Management 2018 (ASRAM 2018) (USB Flash Drive), 6 Pages, 2018/10

In the accidents at Fukushima Dai-ichi Nuclear Power Station, Tsunami caused loss of electric power supply and this event led to core melt and failure of Containment vessel. Finally, fission products were released to the environment. Currently, the activities for understanding of accident progressions are carried out based on the measured data during the accident, accident progression analysis using integrated severe accident analysis codes and investigation of inside of reactor buildings and containment vessels. On the other hand, there are some research activities with combination of accident progression analysis and accident consequence analysis. In Japan Atomic Energy Agency (JAEA), the research project of combination of these analyses using the computational simulation codes has been started. The results obtained from the combination analysis are expected to have broad width of uncertainty because of many uncertainty factors in this combined analysis. In order to perform the analysis efficiently, sensitivity analysis for failure location on containment vessel and its failure size were carried out by THALES2/KICHE developed by JAEA at first. This analysis was performed on unit 1, since it was the first plant to release radioactive materials to the environment during the accident and its consequence had no effect from other plants. The authors focused on the failure of containment vessel head flange, penetration seal and vacuum breaker pipe, and possibility of partial open of vent valve based on the investigations of reactor building inside performed by TEPCO. This paper presents the results obtained from this sensitivity analysis.

Journal Articles

Severe accident scenario uncertainty analysis using the dynamic event tree method

Zheng, X.; Tamaki, Hitoshi; Ishikawa, Jun; Sugiyama, Tomoyuki; Maruyama, Yu

Proceedings of 14th International Conference on Probabilistic Safety Assessment and Management (PSAM-14) (USB Flash Drive), 10 Pages, 2018/09

JAEA Reports

Development of correlation of gaseous ruthenium transfer rate to condensed water in accident of evaporation to dryness by boiling of reprocessed high level liquid waste in Fuel Reprocessing Facilities

Yoshida, Kazuo; Tamaki, Hitoshi; Yoshida, Naoki; Amano, Yuki; Abe, Hitoshi

JAEA-Research 2017-015, 18 Pages, 2018/01

JAEA-Research-2017-015.pdf:3.08MB

An accident of evaporation to dryness by boiling of high level liquid waste is postulated as one of the severe accidents at a fuel reprocessing facility. It was observed at the experiments that a large amount of ruthenium (Ru) is volatilized and transfer to the vapor phase in the tank. The nitric acid and water mixed vapor released from the tank is condensed. Volatilized Ru is expected to transfer into the condensed water at the compartments in the building. Quantitative estimation of the amount of Ru transferred condensed water is key issues to evaluate the reduction the amount of Ru through leak path in the facility building. This report presents that a correlation has been developed for Ru transfer rate to condensed water with vapor condensing rate based on the experimental results and additional thermal-hydraulic simulation of the experiments. Applicability of the correlation has been also demonstrated with the accident simulation of typical facilities in full-scale.

Journal Articles

Application of Bayesian approaches to nuclear reactor severe accident analysis

Zheng, X.; Tamaki, Hitoshi; Shiotsu, Hiroyuki; Sugiyama, Tomoyuki; Maruyama, Yu

Proceedings of Asian Symposium on Risk Assessment and Management 2017 (ASRAM 2017) (USB Flash Drive), 11 Pages, 2017/11

Journal Articles

Current status of research for the accident of evaporation to dryness caused by boiling of reprocessed high level radioactive liquid waste

Tamaki, Hitoshi; Yoshida, Kazuo; Abe, Hitoshi; Sugiyama, Tomoyuki; Maruyama, Yu

Proceedings of Asian Symposium on Risk Assessment and Management 2017 (ASRAM 2017) (USB Flash Drive), 9 Pages, 2017/11

An accident of evaporation to dryness caused by boiling of high level radioactive liquid waste (HLLW) is postulated as one of severe accidents caused by the loss of cooling function at the fuel reprocessing plant. This accident can be divided into early boiling stage, late boiling stage and dry-out stage by characteristics of accident evolution. It is important to estimate the amount of fission product (FP) transport between the liquid and gas phases, and the amount of FP deposition on the walls in each stage in order to estimate the release amount of FP to the environment. Various research activities have been carried out for this issue. This paper reviews these activities and presents the recent activities at JAEA for development of simulation code for this type of accident.

Journal Articles

An Integrated approach to source term uncertainty and sensitivity analysis for nuclear reactor severe accidents

Zheng, X.; Ito, Hiroto; Tamaki, Hitoshi; Maruyama, Yu

Journal of Nuclear Science and Technology, 53(3), p.333 - 344, 2016/03

AA2014-0796.pdf:0.84MB

 Times Cited Count:9 Percentile:75.69(Nuclear Science & Technology)

Journal Articles

Source term uncertainty analysis; Probabilistic approaches and applications to a BWR severe accident

Zheng, X.; Ito, Hiroto; Tamaki, Hitoshi; Maruyama, Yu

Mechanical Engineering Journal (Internet), 2(5), p.15-00032_1 - 15-00032_14, 2015/10

61 (Records 1-20 displayed on this page)