ASTRID nuclear island design; Advances in French-Japanese joint team development of decay heat removal systems

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.

Probabilistic safety assessment on experimental fast reactor Joyo; Level1 PSA for internal events

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 5.010 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.

Development of advanced loop-type fast reactor in Japan, 1; Current status of JSFR development

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.

JSFR; Japan's challenge towards the competitive SFR design concept with innovative technologies

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.

Conceptual Design Studies of HLMC Fast Reactors

Mihara, Takatsugu;

Seminar between Japan and Russia On FBR Cycle Development, 0 Pages, 2001/00

None

Heavy liquid Metal cooled fast breeder reactor; Results in 1999

Mihara, Takatsugu; ; ; ;

JNC TN9400 2000-079, 189 Pages, 2000/07

JNC-TN9400-2000-079.pdf:5.94MB

Based on the medium and long-term program of JNC, the feasibility study for fast breeder reactors (FBRs) including related nuclear fuel cycles has been started from the 1999 fiscal year. Various options of FBR plant systems have been studied and a concept of Heavy Liquid Metal cooled FBRs is one of these options. The purpose of this paper is to research and evaluate Heavy Liquid Metal cooled FBRs on the basis of literatures. First, we selected four types of plant concepts listed below. Concept 1: Large-scale pond type reactor with Pb cooled. Concept 2: Large-scale loop type reactor with Pb cooled. Concept 3: Medium-scale module tank type reactor with Pb cooled. Concept 4: Small scale module tank type reactor with Pb-Bi cooled. Concept l and 2 are selected to seek for scale merit on economical aspect. ln Concept 3 and 4, we tried to reduce the inventory of HLMC and to ease the load conditions on structures and seek for competitiveness with module effect such as mass production and learning effect. Through a preliminary design study, we identified some technical features of each concept and roughly evaluated economical competitiveness based on total weight of the NSSSs. From this study, we concluded (1)lngeneral, the large-scale type concepts have little economical advantage because of its huge amount of material needed for its severe load conditions. (Concept 1&2) (2)Even for the large-scale pond type reactor, the conclusion seems to be the same. Total amount of the thermal shielding material became huge. Aseismatic structure makes the amount of material increase under the Japanese seismic condition. (Concept1) (3)For the large-scale loop type reactor, we selected side entry and dual walled piping concept with slide-joint inner wall to cope with thermal expansion of piping system. However, there seemed to be difficulty with compatibility between slide-joint and oxide film corrosion prevention measures. (Concept2) (4)The medium and small modular type ...

A Conceptual Design Study on Various Type of HLMC Fast Reactor Plant

Mihara, Takatsugu; Tanaka, Yoshihiko;

IAEA-AGM on Design and Perfomanceof Reactor and Su, 0 Pages, 2000/00

None

Development of Phased Mission Analysis Program with Monte Carlo Method - Improvement of the variance reduction technique with biasing towards top event -

Yang Jin An*; Mihara, Takatsugu

JNC TN9400 99-013, 89 Pages, 1998/12

JNC-TN9400-99-013.pdf:2.0MB

This report presents a variance reduction technique to estimate the reliability and availability of highly complex systems during phased mission time using the Monte Carlo simulation. In this study, we introduced the variance reduction technique with a concept of distance between the present system state and the cut set configurations. Using this technique, it becomes possible to bias the tansition from the operating states to the failed states of components towards the closest cut set. Therefore a component failure can drive the system towards a cut set configuration more effectively. JNC developed the PHAMMON (Phased Mission Analysis Program with Monte Carlo Method) code which involved the two kinds of variance reduction techniques : (1) forced transition, and (2)failure biasing. However, these techniques did not guarantee an effective reduction in variance. For further improvement, a variance reduction technique incorporating the distance concept was introduced to the PHAMMON code and the numerical calculation was carried out for the different design cases of decay heat removal system in a large fast breeder reactor. Our results indicate that the technique addition of this incorporating distance concept is an effective means of further reducing the variance.

Development of the fast minimal cut sets calculation method for living PSA application; Improving the pre-deduced minimal cut sets method with success pass sets

Mihara, Takatsugu

PNC TN9410 96-273, 36 Pages, 1996/11

PNC-TN9410-96-273.pdf:1.48MB

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.

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Mihara, Takatsugu

IAEA Technical Committee Meeting "Procedures for PSA for Shutdown and Other Low Power Operating Mode, , 

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