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Journal Articles

Replacement of upper core structure in experimental fast reactor Joyo, 1; Existing damaged upper core structure jack-up test

Ito, Hiromichi; Suzuki, Nobuhiro; Kobayashi, Tetsuhiko; Kawahara, Hirotaka; Nagai, Akinori; Sakao, Ryuta*; Murata, Chotaro*; Tanaka, Junya*; Matsusaka, Yasunori*; Tatsuno, Takahiro*

Proceedings of 2015 International Congress on Advances in Nuclear Power Plants (ICAPP 2015) (CD-ROM), p.1058 - 1067, 2015/05

In the experimental fast reactor Joyo (Sodium-cooled Fast Reactor (SFR)), it was confirmed that the top of the irradiation test sub-assembly had bent onto the in-vessel storage rack as an obstacle and had damaged the upper core structure (UCS). There is a risk of deformation of the UCS and guide sleeve (GS) caused by interference between them unless inclination is controlled precisely. To mitigate the risk, special jack-up equipment for applying three-point suspension was developed. The existing damaged UCS (ed-UCS) jack-up test using the jack-up equipment was conducted on May 7, 2014. As a result of this test, it was confirmed that the ed-UCS could be successfully jacked-up to 1000 mm without consequent overload. The experience and knowledge gained in the ed-UCS jack-up test provides valuable insights and prospects not only for UCS replacement but also for further improving and verifying repair techniques in SFRs.

Journal Articles

Development of risk assessment methodology against natural external hazards for sodium-cooled fast reactors; Project overview and strong wind PRA methodology

Yamano, Hidemasa; Nishino, Hiroyuki; Kurisaka, Kenichi; Okano, Yasushi; Sakai, Takaaki; Yamamoto, Takahiro*; Ishizuka, Yoshihiro*; Geshi, Nobuo*; Furukawa, Ryuta*; Nanayama, Futoshi*; et al.

Proceedings of 2015 International Congress on Advances in Nuclear Power Plants (ICAPP 2015) (CD-ROM), p.454 - 465, 2015/05

This paper describes mainly strong wind PRA methodology development in addition to the project overview. In developing the strong wind PRA methodology, hazard curves were estimated by using Weibull and Gumbel distributions based on weather data recorded in Japan. The obtained hazard curves were divided into five discrete categories for event tree quantification. Next, failure probabilities for decay heat removal related components were calculated as a product of two probabilities: i.e., a probability for the missiles to enter the intake or outtake in the decay heat removal system, and fragility caused by the missile impacts. Finally, based on the event tree, the core damage frequency was estimated about 6$$times$$10$$^{-9}$$/year by multiplying the discrete hazard probabilities in the Gumbel distribution by the conditional decay heat removal failure probabilities. A dominant sequence was led by the assumption that the operators could not extinguish fuel tank fire caused by the missile impacts and the fire induced loss of the decay heat removal system.

Journal Articles

Safety improvement in building layout design to meet the safety design criteria for the Generation IV SFR

Kato, Atsushi; Chikazawa, Yoshitaka; Nabeshima, Kunihiko; Iwasaki, Mikinori*; Akiyama, Yo*; Oya, Takeaki*

Proceedings of 2015 International Congress on Advances in Nuclear Power Plants (ICAPP 2015) (CD-ROM), p.593 - 600, 2015/05

Japan sodium cooled fast reactor is the advanced loop type reactor developing in Japan. After the Fukushima-Dai-ichi NPP accident, system enhancement against severe accident have been investigated mainly for residual decay heat removal system, spent fuel storage system and emergency power sources in order to satisfy the safety design criteria for Generation IV SFR. This paper describes principle of the building layout design and the actual approach to be consistent with the recent design enhancement in JSFR. From the perspective of greater ability to withstand severe events, the principles of the building layout design as the measures against aircraft attack and the consequential fire, and tsunami are introduced in order to avoid local event initiating and simultaneous redundant failure of the safety grade facilities and could achieve lowering risk of the loss of all stuck and maintaining the essential power supply.

Journal Articles

Overview of recent methods for the modeling of the uncertainties on the calculations of consequences of a nuclear power plant severe accident

Chevalier-Jabet, K.*; Zheng, X.; Mabrouk, A.*; Maruyama, Yu; Baccou, J.*

Proceedings of 2015 International Congress on Advances in Nuclear Power Plants (ICAPP 2015) (CD-ROM), 13 Pages, 2015/05

Journal Articles

Validation of plant dynamics analysis code using shutdown heat removal test-17 performed at the EBR-II

Ohira, Hiroaki; Doda, Norihiro; Kamide, Hideki; Iwasaki, Takashi*; Minami, Masaki*

Proceedings of 2015 International Congress on Advances in Nuclear Power Plants (ICAPP 2015) (CD-ROM), p.2585 - 2592, 2015/05

IAEA's Coordinated Research Project on Benchmark Analyses of Shutdown Heat Removal Test (SHRT) performed at the Experimental Breeder Reactor-II (EBR-II) has been carried out since 2012. The benchmark specifications were provided by the Argonne National Laboratory (ANL) and the model development for thermal-hydraulics codes and/or plant dynamics codes has been conducted by participating organizations. The experimental data were also provided by the ANL after the calculations have been performed as the blind simulation. JAEA participated in this benchmark analyses, and the plant dynamics analysis code; Super-COPD was applied to the SHRT-17 simulation. The calculated inlet temperature of the high pressure plenum agreed well with the test data in all simulation time. Although the Z-pipe inlet temperature and the IHX intermediate outlet temperature had some discrepancy in the first 400 sec. caused by larger mass flow rate of the primary pump and the perfect mixing model of upper plenum, these temperatures and the flow rate agreed well with the measured data after 400 sec. Hence it was concluded the present analytical model could predict the natural circulation in good accuracy.

Journal Articles

Design study for reactor system of fast reactor JSFR; Concept of reactor system

Kawasaki, Nobuchika; Sakamoto, Yoshihiko; Eto, Masao*; Taniguchi, Yoshihiro*; Kamishima, Yoshio*

Proceedings of 2015 International Congress on Advances in Nuclear Power Plants (ICAPP 2015) (CD-ROM), p.760 - 769, 2015/05

The Japan Sodium-cooled Fast Reactor, JSFR, is currently under conceptual study. The concept of JSFR's reactor system is a compact reactor system to avoid excessive increase of reactor vessel diameter with structural and fluid integrities. To realize this concept, single rotating plug with advanced refueling system is adopted. Advanced refueling system consists of column type Upper Internal Structure and pantograph type Fuel Handling Machine. To realize structural and fluid integrities, top entry piping, sodium dam and flow block/guide structures are adopted. Structural integrities against seismic displacement or thermal stress and fluid integrities against vortex cavitations or cover gas entrainment can be ensured with these designs.

Journal Articles

Performance test of under sodium viewer in Monju

Aizawa, Kosuke; Togashi, Yoshinori; Sasaki, Koei; Chikazawa, Yoshitaka; Fukuie, Masaru*; Jimbo, Noboru*

Proceedings of 2015 International Congress on Advances in Nuclear Power Plants (ICAPP 2015) (CD-ROM), p.808 - 816, 2015/05

Inspection technique in opaque liquid metal coolant is one of the important issues for the safety warranty of Liquid Metal Fast Breeder Reactor (LMFBR) core. A performance test of Under Sodium Viewer (USV) which was developed to detect obstacles in reactor vessel of LMFBR Monju was carried out. The ultrasonic sensors and reflectors are located across the core inside the Monju reactor vessel. The USV detects the obstacle between the core top and the bottom of Upper Core Structure (UCS) by differences of echo signals. This reports showed the USV performance test in Monju before power operation. In the test, the basic echo signals in various conditions were accumulated and signal to noise ratio met with the design value. Measured signals with and without obstacles showed difference clearly. Those experimental results showed that basic performance of the USV to detect an obstacle between the core and UCS.

Journal Articles

Safety operation of chromatography column system with discharging hydrogen radiolytically generated

Watanabe, So; Sano, Yuichi; Nomura, Kazunori; Koma, Yoshikazu; Okamoto, Yoshihiro

Proceedings of 2015 International Congress on Advances in Nuclear Power Plants (ICAPP 2015) (CD-ROM), p.2781 - 2788, 2015/05

Journal Articles

IAEA NAPRO Coordinated Research Project; Physical properties of sodium

Passerini, S.*; Carardi, C.*; Grandy, C.*; Azpitarte, O. E.*; Chocron, M.*; Japas, M. L.*; Bubelis, E.*; Perez-Martin, S.*; Jayaraj, S.*; Roelofs, F.*; et al.

Proceedings of 2015 International Congress on Advances in Nuclear Power Plants (ICAPP 2015) (CD-ROM), p.780 - 790, 2015/05

Journal Articles

Application of the high-temperature gas-cooled reactor to produce tritium for fusion reactors

Nakaya, Hiroyuki*; Matsuura, Hideaki*; Katayama, Kazunari*; Goto, Minoru; Nakagawa, Shigeaki

Proceedings of 2015 International Congress on Advances in Nuclear Power Plants (ICAPP 2015) (CD-ROM), p.398 - 402, 2015/05

The performance of tritium production for fusion reactor using High-Temperature Gas-cooled Reactor (HTGR) is studied. An influence of $$^{6}$$Li concentration on tritium production performance using HTGR is estimated. Li compound is loaded in the reactor core using Li rod consisting cylindrical Li compound in cladding tube. A Gas Turbine High-Temperature Reactor of 300 MWe nominal capacity (GTHTR300) with 600 MW thermal output power is assumed as HTGR. An amount of tritium production is estimated by burn-up calculations using the continuous-energy Monte Carlo transport code MVP-BURN. The amount of tritium outflow is estimated from equilibrium solution for the tritium diffusion equation in the cladding tube. Even if 6Li is enriched, the GTHTR300 can produce 500 g of tritium over 180-day operation without increasing the amount of required $$^{6}$$Li. The amount of tritium outflow is decreased by 20-50%.

Journal Articles

Japan-France collaboration on the astrid program and sodium fast reactor

Rouault, J.*; Le Coz, P.*; Garnier, J.-C.*; Hamy, J.-M.*; Hayafune, Hiroki; Iitsuka, Toru*; Mochida, Haruo*

Proceedings of 2015 International Congress on Advances in Nuclear Power Plants (ICAPP 2015) (CD-ROM), p.832 - 837, 2015/05

The French and international industrial partners already joined the project from 2010 to 2013 and many others are also effective in the Research and Development in support of ASTRID. A new partnership is now effective on both topics with Japan. This collaboration on the ASTRID Program and Sodium Fast Reactor is now fully integrated in the ASTRID program organization. In addition a specific Joint Team, CEA, AREVA, JAEA, MHI and MFBR, has been created to follow specifically Japanese contribution and develop evaluations of a common interest to orientate future work and contribute to ASTRID options confirmation and be of an interest for the future Japanese Fast Breeder reactor.

Journal Articles

ASTRID, the SFR GENIV technology demonstrator project; Where are we, where do we stand for?

Rouault, J.*; Abonneau, E.*; Settimo, D.*; Hamy, J.-M.*; Hayafune, Hiroki; Gefflot, R.*; Benard, R.-P.*; Mandement, O.*; Chauveau, T.*; Lambert, G.*; et al.

Proceedings of 2015 International Congress on Advances in Nuclear Power Plants (ICAPP 2015) (CD-ROM), p.824 - 831, 2015/05

The Preconceptual Design phase of the ASTRID Project ended late 2012, the main goal was to evaluate innovative options. It is now followed by the AVP2 phase planned until the end of 2015 whose objectives are both to focus the design in order to finalize a coherent reactor outline and to finalize by December 2015 the Safety Option Report. The CEA acts as the industrial architect of the project. In 2014, Japan which participates now in the design studies and also in Research and Development in support of the ASTRID Project and VELAN are the latest partners to join the Project. The next important milestone is at the end of 2015 with the release by the Project team of a convincing and coherent Conceptual Design file.

Oral presentation

Study on a suppression of sodium-water reaction in SFR by applying sodium with suspended nanoparticles

Kanda, Hironori; Yoshioka, Naoki*; Ara, Kuniaki; Saito, Junichi; Nagai, Keiichi

no journal, , 

An important safety concern for SFR is the possibility of steam leaking from a tube at high-pressure into surrounding liquid sodium within the steam generator. The steam/liquid sodium pair is very reactive and the steam leak will form a sodium-water reaction jet which may attack and rupture adjacent tubes. A study on the suppression of the reactivity of sodium itself using the concept of suspended nanoparticles in liquid sodium (sodium nanofluid) has carried out. From the experimental results for sodium nanofluid, it was clear that the reaction rate and reaction heat with water were decreased by the atomic interaction of sodium with suspended nanoparticles. Taking the changes of physical and chemical property into account, an analytical model for peak temperature within a sodium nanofluid-water reaction jet has been constructed. The object of this paper is to confirm this analytical model for reaction jet temperature and to predict a mitigation effect on adjacent tube damage in a steam generator by applying sodium nanofluid to the secondary coolant of SFR. Comparing calculation results using the analytical model with the experimental results for a steam injection experiment, we demonstrated that the analytical model for temperature within the reaction jet is appropriate.

Oral presentation

Residual issues on fast reactor development and safety design criteria as global standards

Sagayama, Yutaka

no journal, , 

Fast reactors are vital to achieve nuclear fuel cycle sustainability. Extensive experiences of Sodium cooled fast reactors (SFRs) have been accumulated with past and existing prototype reactors in Japan and the world and SFR technologies are matured from industrial point of view. The residual issues of SFR are the reduction of construction cost and improvement in maintenance and repair. Safety Design Criteria (SDC) is under discussion in the Generation-IV International Forum (GIF) framework to make SDC as global standard.

Oral presentation

Current status and future plans on nuclear fuels & fuel cycle for LWRs and fast neutron nuclear reactors in Japan

Ono, Kiyoshi

no journal, , 

After the Great East Japan Earthquake, Japanese government has maintained the nuclear fuel cycle policy. Regarding the LWR fuel cycle, Rokkasho fuel cycle plants are being steadily constructed as private projects. As for FR fuel cycle, JAEA, as a primary nuclear institute, will proceed the research and development of technologies including MA-bearing MOX fuel fabrication and MA separation for the purpose of volume reduction and mitigation of degree of harmfulness of radioactive waste based on the Strategic Energy Plan adopted by the cabinet in 2014.

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