著者
Michio Yamawaki Yuji Arita Takayuki Terai Tadafumi Koyama Koichi Uozumi Yuma Sekiguchi Masami Taira
出版者
The Japan Society of Mechanical Engineers
雑誌
Proceedings of the ... International Conference on Nuclear Engineering. Book of abstracts : ICONE (ISSN:24242934)
巻号頁・発行日
pp._ICONE23-1, 2015-05-17 (Released:2017-06-19)
被引用文献数
1 1

Source term for severe accident analysis of molten salt reactors(MSRs) has been investigated as part of preliminary efforts to develop MSRs. As a severe accident of MSRs, exposure of heated fluoride fuel molten salt to atmosphere was assumed to take place. Vaporization of fluoride molten salt was studied by means of the two methods, the Knudsen effusion mass spectrometry as well as the transpiration method. The former was applied to pseudo-binary fluoride systems to clarify the behaviors of cesium and iodine in the fluoride molten salt. The latter was applied to the mixture of CsI and FLiNaK. These experiments were carried out as the first step of the source term studies, so that interaction with air components has not been covered yet. From this study, useful information related to the source term for MSRs have been obtained. This work suggests how to solve the problem to establish the source term for severe accident analysis of MSRs.
著者
Luca Facciolo Pekka Nuutinen Daniel Welander
出版者
The Japan Society of Mechanical Engineers
雑誌
Proceedings of the ... International Conference on Nuclear Engineering. Book of abstracts : ICONE (ISSN:24242934)
巻号頁・発行日
pp._ICONE23-1, 2015-05-17 (Released:2017-06-19)

The European Utility Requirements organization started the compliance assessment process of the Mitsubishi Heavy Industries EU-APWR Standard Design in 2012. The EU-APWR is an Advanced PWR, 1700 MWe class, 4-loops, 14ft active core fuel length that MHI has developed for the European market. The EU-APWR is an evolution of the Advance PWR currently under the licensing process in Japan for the Tsuruga Power Station. MHI has modified the design applying improvements in safety and economy in order to be adapted for European markets and to comply with the EUR requirements. The EU-APWR Standard Design documentation has been assessed against the EUR Volume 2 - Generic Nuclear Island requirements Revision D, issued in October 2012. The assessment is divided into 20 chapters for a total of over four thousand individual requirements. Each chapter was assigned to Assessment Performers who executed the detailed analysis of the requirements. The assessment of each requirement and the Synthesis Report have been submitted to, and scrutinized by, the Coordination Group, formed by representatives of the European Utilities together with the Vendor, and reviewed by the Administration Group and by the Steering Committee. The Synthesis Reports have been collected in the Volume 3 EU-APWR Standard Design Subset and presented to the Steering Committee, which approved the final draft in October 2014. The overall results of the assessment indicated good compliance of the EU-APWR Standard Design: 77% of the requirements resulted in compliance with EUR. This percentage increases to 85% when taking into account the requirements for which the design has been evaluated in compliance with the objectives. The requirements where the design has been judged not in compliance with EUR are less than 2%. The divergences between the EU-APWR Standard Design and the EUR concern different areas like, for instance, layout, operational capability and performance, outage operations, personal protection and radiation monitoring. Some of the incongruences result from differences in approach to the design process or from differences in the rules and standards in use in Japan and in Europe. Some analyses, like the internal hazards effects, have been performed only partially because, in Japan, such analyses are considered site-specific and are carried out at the detailed design level. The analysis of the consequences of a hydrogen explosion, and the environmental qualification methodology of equipment have not been fully developed yet. While the reactor core has been designed for an operability cycle of 24 months and can be loaded with 50% MOX fuel, no other area of the plant has been designed taking into consideration MOX fuel.
著者
Li Longze Qiu Jinrong Tai Yun Wang Jue Su G.H. Liu Baolin Hou Xiaofan You Ximing
出版者
一般社団法人 日本機械学会
雑誌
Proceedings of the ... International Conference on Nuclear Engineering. Book of abstracts : ICONE
巻号頁・発行日
vol.2019, 2019

The marine nuclear power plant is a floating nuclear power plant which supply power for the offshore oil drill platform. It is designed based on the standard and experience of the traditional onshore nuclear power plant. The reactor of the marine nuclear power plant is a 100 MWt PWR type small modular reactor with 2 loops. Each of the loops contains a main pump, a main check valve, and a steam generator. The pressurizer is set on one of the loops. The engineering safety features in the plant are somewhat different from those in the traditional plants. The special residual heat removal system, the passive residual heat removal, the square steel containment and containment suppression system are designed in the plant. The prevention and mitigation measures for severe accidents are set up on a reasonable and feasible basis to actually eliminate the large release of radioactive products. A MELCOR model and corresponding input deck were developed for the reactor coolant system, the secondary system, the containment system, the engineering safety features. Based on the safety analysis experience in the traditional nuclear power plant, the SBLOCA in the cold leg with the break diameter of 2.5 cm is chose as the initial event of the severe accident in the work. The sequence and important parameters in the accident are analyzed. According to the simulation results, the core exposed and heat up with the coolant release, and finally melt since the emergency core cooling system failure. However, the reactor pressure vessel (RPV) maintained integrity with the mitigation measure, i.e., the external vessel core cooling. The containment also maintained integrity, which prevented the large release of radioactive products to the other cabins and the environment. The work is useful in gaining an insight into the detailed process involved. One of the final goals of this work would be to identify appropriate accident management strategies and countermeasures for the SBLOCA induced severe accidents during the design process of the marine nuclear power plant.
著者
Quang Anh Tho Tran Azola Edson Closon Herve Chares Robert
出版者
一般社団法人 日本機械学会
雑誌
Proceedings of the ... International Conference on Nuclear Engineering. Book of abstracts : ICONE
巻号頁・発行日
vol.2011, pp._ICONE1944, 2011

Operational decisions related to plant performance monitoring mainly result from raw process measurement analysis. When first signs of sub-optimal behavior occur, traditional methods mainly focus on the observation of selective pieces of information. The main disadvantages of these methods are: ・Investigation efforts are required to localize the problems, entailing time losses and costs; ・Validity and reliability of the pieces of information cannot be checked as long as the measurements are observed individually. The problem is not the lack of methods and techniques but rather a lack of reliable and consistent data and information across the entire plant. To overcome drawbacks of traditional methods, measurements are considered as interacting with one another. When related to the other measurements of the plant, the observed information becomes of an interest: its incoherency to the others identifies and localizes a problem. The Data Validation and Reconciliation technology (DVR) is based on an advanced process data coherency treatment. By using all available plant information and by closing the plant heat and mass balances based on rigorous thermodynamics, the method generates: ・A single set of reliable and most accurate plant process data; ・Alarms for incoherent measurements, highlighting potential problems; ・Alarms for equipment and process performance degradation; ・Alarms for faulty and drifting measurements. The use of the advanced DVR software package VALI offers various benefits as it allows to base operational decisions on reliable and accurate data.
著者
HAGA Kazuo Itoh Tomomichi ENDO Hiroshi SHINDO Yoshihisa BAGLIETTO Emilio SASAKI Yasutomo IRIKURA Motoki
出版者
一般社団法人 日本機械学会
雑誌
Proceedings of the ... International Conference on Nuclear Engineering. Book of abstracts : ICONE
巻号頁・発行日
vol.2011, pp._ICONE1943, 2011
被引用文献数
1

When heat-transfer tube failure occurred in SG of LMFBR and the steam leak rate is medium (some 10s g/s-some kg/s), overheating tube rupture would be caused due to the accelerated high-temperature creep in the heat transfer tubes surrounded by the high temperature products of sodium-water reaction (SWR). However, the detailed analysis code to analyze this phenomenon is not established. Adopting a Computational Fluid Dynamics (CFD) code would be a promising candidate for that perpose. At first in developing the analysis tool using a CFD code, STAR-CCM+ and LHM (Locally Homogeneous Multi-phase) model was adapted considering the appropriateness to SWR and the merits of homogeneous model, that is, the light load of computing resources and the minimum usage of models whose applicability is remained in argument. In the second step, a concept of interfacial area density was introduced to the submerged jet analysis. Furthermore, a sodium-water reaction model was added as an external function of the CDF code. A trial calculation was made to the basic SWR experimental data obtained by Hobbes et al. The analysis of the temperature profile formed in the jet region showed a good agreement with the experiment by properly choosing parameters such as Lewis number and the heat transfer coefficient on the sodium droplet. No marked difference was seen between the two-dimensional analysis and the three-dimensional analysis.
著者
Kuang Liuwei Ren Liang Jing Linzhi Wen Bang Liu Huarong
出版者
一般社団法人 日本機械学会
雑誌
Proceedings of the ... International Conference on Nuclear Engineering. Book of abstracts : ICONE
巻号頁・発行日
vol.2015, pp._ICONE23-1-_ICONE23-1, 2015

The significance and development status were introduced about the pneumatic leak test pressure tightness for irradiated fuel rods in this paper, the pneumatic leak test was conducted, and the track mediator and experiment pressure of the pneumatic leak test were defined for irradiated fuel rods. With the consideration of the factors such as sealing, fixing operation, leakage monitoring system, tracer medium and pressure of the irradiated fuel rods in hot cell, the pneumatic leak test device was designed and the gas tightness inside and outside the hot cell was verified. Through the pneumatic leak test for the artificial simulative fuel rods, the effectiveness of the pneumatic leak test device as well as the feasibility of the method were proved, the requirements of pneumatic leak test were met, the technique of pneumatic leak test for irradiated fuel rods was established, the pneumatic leak test for irradiated fuel rods under strong radioactive environment was realized, and the leakage condition and position data of irradiated fuel rods were acquired.
著者
Yang Zhifei Chen Yali Luo Hu Peng Zhenxun Wang Zen
出版者
一般社団法人日本機械学会
雑誌
Proceedings of the ... International Conference on Nuclear Engineering. Book of abstracts : ICONE
巻号頁・発行日
vol.2015, no.23, pp."ICONE23-1266-1"-"ICONE23-1266-4", 2015-05-17

The development of smart support system is to meet the urgent needs and strengthen requirements for the training and application for Full-Scope Severe Accident Management Guidelines (FSSAMG) among utilities, nuclear regulators and research and development (R&D) institute. One of the lessons learned after Fukushima accident is that the existing Severe Accident Management Guidelines (SAMGs) used in Nuclear Power Plants (NPPs) are not sufficient to cope with site wide catastrophic event. Therefore, the development of a full scope SAMGs that update and extend the guidelines to include comprehensive plant conditions and portable equipment is necessary. The smart support system is to alleviate the effort and challenges that encountered in FSSAMG validation, application and training. Besides, it provides intuitive insights to severe accident progression and guidelines execution, and thus to support the decision-making and mitigate the accident consequences.
著者
Tinoco Hernan Ahlinder Stefan Hedberg Peter
出版者
一般社団法人日本機械学会
雑誌
Proceedings of the ... International Conference on Nuclear Engineering. Book of abstracts : ICONE
巻号頁・発行日
vol.2007, no.15, pp."ICONE1510051-1"-"ICONE1510051-12", 2007-04-22

A power uprate of Forsmark's Unit 3 from 109 % to 125 % will be implemented during the 2010 refuelling outage. This uprate implies an increase in gamma heating of the core shroud which could lead to temperatures higher than the design thickness-mean temperature, 300℃, according to ASME regulations. To estimate the temperature distribution in the core shroud, a CFD model of the core bypass has been developed using the commercial code CFX 5. The model consists of the core bypass, from the lower core support plate up to the core grid, and the upper plenum, limited from above by the core shroud cover including the steam separator inlets. The bypass flow enters the computational domain at the level of the core support plate. The two-phase core flow enters the computational domain at the level of the core grid where it entrains and merges with the bypass flow. Both flows have been estimated through the POLCA code. The conjugate heat transfer at the core shroud inner wall comprises the gamma heating from the core, considered as volume distributed heat sources, and subcooled boiling of the bypass flow. The effect of subcooled boiling has been taken care of by using the model by Kurul and Podowski. Using a conservative gamma heat source distribution, leads to local temperatures slightly higher than the design temperature, with a maximum thickness-mean temperature that exceeds the temperature limit by approximately 4℃. If a higher temperature limit is accepted, the ASME regulations are not fulfilled, but the consequence is a minor change in the design stress intensity value Sm according to ASME. Using a somewhat realistic gamma heat source distribution, the results show that the maximum thickness-mean temperature is well below the design temperature.