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Hourcade, E.*; Mihara, Takatsugu; Dauphin, A.*; Dirat, J.-F.*; Ide, Akihiro*
Proceedings of 2018 International Congress on Advances in Nuclear Power Plants (ICAPP 2018) (CD-ROM), p.556 - 561, 2018/04
In the framework of the French-Japanese agreement signed in 2014, CEA, AREVA NP, JAEA, and MHI/MFBR is jointly performing components design of ASTRID such as Decay Heat Removal Systems (DHRS). This paper is giving an update concerning ASTRID DHR strategy with description of reference architecture evolution and project objectives. In particular, new developments were made for DHR during normal shutdown and role of Ex-Vessel system. A special focus is made on design process of automatic shutter to hydraulically connect Hot Plenum and cold plenum to enhance primary vessel natural convection.
Hourcade, E.*; Curnier, F.*; Mihara, Takatsugu; Farges, B.*; Dirat, J.-F.*; Ide, Akihiro*
Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.1740 - 1745, 2016/04
In the framework of the French-Japanese agreement signed in 2014, CEA, AREVA NP, JAEA, and MHI/MFBR is jointly performing components design of ASTRID such as Decay Heat Removal Systems (DHRS). This paper is giving highlights of ASTRID DHRS current strategy. Focus is made on operating temperature diversification for in-vessel heat exchanger as well as core catcher coolability by an original features such as heat exchanger located within reactor cold pool, whose design was taken over by Japan team since 2014.
Kotake, Shoji; Sakamoto, Yoshihiko; Mihara, Takatsugu; Kubo, Shigenobu*; Uto, Nariaki; Kamishima, Yoshio*; Aoto, Kazumi; Toda, Mikio*
Nuclear Technology, 170(1), p.133 - 147, 2010/04
Times Cited Count：36 Percentile：91.14(Nuclear Science & Technology)Japan Atomic Energy Agency (JAEA) is now executing "Fast Reactor Cycle Technology Development (FaCT)" project in cooperation with the Japanese electric utilities. In the FaCT project, both the conceptual design study for Japan Sodium-cooled Fast Reactor (JSFR) and the developments of innovative technologies to be adopted to JSFR are now implemented with paying attention to the consistency between the design and the innovative technologies. The current target is that the development will be accomplished around 2015, after that a licensing procedure for the demonstration JSFR will be launched. This paper describes design requirements, design characteristics of JSFR and evaluation on the performances for economic competitiveness. Furthermore, the current status of the key technology development for JSFR is briefly introduced.
Ishikawa, Koki; Takamatsu, Misao; Kawahara, Hirotaka; Mihara, Takatsugu; Kurisaka, Kenichi; Terano, Toshihiro; Murakami, Takanori; Noritsugi, Akihiro; Iseki, Atsushi; Saito, Takakazu; et al.
JAEA-Technology 2009-004, 140 Pages, 2009/05
JAEA-Technology-2009-004.pdf:2.0MB
Probabilistic safety assessment (PSA) has been applied to nuclear plants as a method to achieve effective safety regulation and safety management. In order to establish the PSA standard for fast breeder reactor (FBR), the FBR-PSA for internal events in rated power operation is studied by Japan Atomic Energy Agency (JAEA). The level1 PSA on the experimental fast reactor Joyo was conducted to investigate core damage probability for internal events with taking human factors effect and dependent failures into account. The result of this study shows that the core damage probability of Joyo is 5.0
10
per reactor year (/ry) and that the core damage probability is smaller than the safety goal for existed plants (10 ry) and future plants (10
/ry) in the IAEA INSAG-12 (International Nuclear Safety Advisory Group) basic safety principle.
Uto, Nariaki; Sakai, Takaaki; Mihara, Takatsugu; Toda, Mikio*; Kotake, Shoji; Aoto, Kazumi
Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9298_1 - 9298_11, 2009/05
A conceptual design for JSFR and developments of innovative technologies are implemented. A compact RV has been designed to enhance the economy. The regarding development results have been reflected to the RV design. An innovative CV design has been implemented with elemental tests to reduce the construction cost. SASS and the NC DHRS have been designed to enhance the safety, with the irradiation data acquired in Joyo and the development of a 3-dimensional thermal-hydraulic evaluation method. An approach for ISI/R has been provided to be applicable for FR characteristics, and the developmental studies on innovative inspection technologies have been progressed. Other technologies including double-walled pipes with short elbows, a pump-integrated IHX are also being developed. These results, together with a preliminary conceptual design study on a demonstrative reactor for JSFR, will be utilized as resources in 2010 to determine which innovative technologies should be adopted.
Kotake, Shoji; Mihara, Takatsugu; Kubo, Shigenobu; Aoto, Kazumi; Toda, Mikio*
Proceedings of 2008 International Congress on Advances in Nuclear Power Plants (ICAPP '08) (CD-ROM), p.486 - 495, 2008/06
Japan Atomic Energy Agency (JAEA) is now executing "Fast Reactor Cycle Technology Development (FaCT)" project in cooperation with the Japanese electric utilities. In the FaCT project, both the conceptual design study for JAEA sodium-cooled fast reactor (JSFR) and the developments of the innovative technologies are now implemented with paying attention to the consistency between the design and the innovative technologies. The current target is that the development will have been accomplished around 2015, after that a licensing procedure for JSFR demonstration reactor will be launched. This paper describes the design requirements, design characteristics of JSFR and evaluation on the performances for economic competitiveness. Furthermore, the current status of the key technology development for JSFR is briefly introduced.
Usui, Shinichi; Mihara, Takatsugu; Obata, Hiroyuki; Kotake, Shoji
Proceedings of 2008 International Congress on Advances in Nuclear Power Plants (ICAPP '08) (CD-ROM), p.512 - 518, 2008/06
Refueling operation of sodium fast reactor (SFR) is one of major technical issue due to the chemical activities and opaqueness of sodium coolant properties in comparison with that of LWR. In the Japan Atomic Energy Agency (JAEA) sodium cooled Fast Reactor (JSFR) design study, the further reliable and rational fuel handling system (FHS) has been developing based on the experience of safe and reliable fuel handling operation in the existent SFR plants. Some of advanced concepts for the FHS have being studied in order to increase economic competitiveness further by attempting reduction of the amount of the material and the refueling time, and are scheduled to execute elemental tests and/or mock-up tests to confirm their feasibilities.
Mihara, Takatsugu; Kotake, Shoji
Proceedings of 15th Pacific Basin Nuclear Conference (PBNC-15) (CD-ROM), 6 Pages, 2006/10
The economic competitiveness is one of the crucial points and has been emphasized in the design study of JSFR (JAEA Sodium Fast Reactor) concept. By employing innovative technologies for the NSSS cost reduction, and the complete passive Decay Heat Removal System that leads to BOP cost improvement, it is confirmed that the JSFR design concept is well suited to the development target equivalent to 1,000USD/kWe (as nth-of-a-kind, overnight cost). In this paper, the economical aspect in the JSFR design was summarized.
Mihara, Takatsugu
PNC TN9410 96-273, 36 Pages, 1996/11
The plant configuration control system is being developed as one of the application of Living PSA. In order to evaluate risk measures, such as core damage frequency, among many kinds of plant configurations, shorter Boolean calculation time is favorable. One of the way to reduce the calculation time is the recalculation method based on the minimal cut set (MCS)data, not on the fault tree data. However, some of the truncated cut set terms may be important in some specific plant configurations. Therefore, this method has possibility to miss the important cut set terms and to get invalid risk measures. To avoid invalidity with this method, the way to improve the recalculation method was developed. The step of the development method is shown below. (1)Cut set data preparation. (a)For the safety systems to be evaluated, deducing the minimal pass sets that stand for minimal combinations of the support systems that are needed to make the safety system function. These minimal pass sets are represented in the form of boolean algebra using symbols that refer to the support systems. (b)Calculating the MCS data for the safety systems based on the each plant configuration identified with the each minimal pass set term. (2)Risk evaluation with specific plant configuration. (a)Any plant configurations to be evaluated can be represented in the from of the sum of the some minimal pass set terms identified above. Selecting the such minimal pass set terms matching the specific plant configuration to be evaluated. (b)Identifying the MCS data corresponding to the minimal pass set terms selected in the previous step. (c)Multiplying these MCS and simplifying with Boolean algebra, then we can get the MCS correspond to the configuration. Applying this method to the reliability analysis of the decay heat removal system of an LMFBR model plant, and compared with the fault tree calculation method, the validity of this method was assured.
Okano, Yasushi; Mihara, Takatsugu; Sakai, Takaaki; Nakai, Ryodai; Kubo, Shigenobu*
no journal, ,
The sodium-cooled fast reactor should be deployed to world standard in future, based on the domestic preparations on safety requirements and codes and standards. This study shows the fundamental approach and roadmap for safety harmonization of the SFR.
