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Mori, Takero; Ohira, Hiroaki; Sotsu, Masutake; Fukano, Yoshitaka
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 9 Pages, 2017/04
Since safety measures against severe accidents (SAs) such as a long-term station blackout (SBO) are required for Japanese prototype fast breeder reactor Monju, a validation is necessary for the plant dynamics code during SBO. In order to take into account the phenomena in natural circulation: a heat transfer among subassemblies and a flow redistribution, a whole core model has been developed for the plant dynamics code, Super-COPD. This model has been validated by test results of natural circulation in actual facility. In this study, this whole core model was applied to Monju core to evaluate safety measures against SBO, and the pressure loss model of Monju was validated by comparing with results of the plant trip test from the power of 40%. In addition, an analysis was conducted for SBO to investigate the applicability of this model to Monju. The applicability of this model was confirmed by comparing with analytical results using the model without heat transfer between assemblies.
Yamashita, Takuya; Nagae, Yuji; Kikuchi, Koichi*; Yamamoto, Kenji*
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 8 Pages, 2017/04
Kubota, Ryuzaburo; Koyama, Kazuya*; Moriwaki, Hiroyuki*; Yamada, Yumi*; Shimakawa, Yoshio*; Suzuki, Toru; Kawada, Kenichi; Kubo, Shigenobu; Yamano, Hidemasa
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 Pages, 2017/04
This paper describes an analysis study on the initiating phase of the ATWS events with SAS4A in order to confirm the appropriateness of the core design for the medium-scale SFR (750MWe-1765MWt). Not using a conventional lumping method that multiple fuel sub-assemblies having a similar characteristic were assigned to one channel (representing fuel assembly in SAS4A), each channel represents only the sub-assemblies of identical operating condition. In addition, the detailed power and reactivity distribution were set reflecting the change of insertion position of control rods. Applying these detailed analysis conditions, the SAS4A analyses were performed for unprotected loss-of-flow (ULOF) and unprotected transient overpower (UTOP) during both of the nominal power and the partial power operation. As a result, more proper event progression including incoherency of events especially fuel dispersion after fuel failure was successfully evaluated and then this analysis study suggested that the power excursion with prompt criticality leading to large mechanical energy release can be prevented in the initiating phase of the current design.
Mukai, Yasunobu; Nakamichi, Hideo; Kobayashi, Daisuke; Nishimura, Kazuaki; Fujisaku, Sakae; Tanaka, Hideki; Isomae, Hidemi; Nakamura, Hironobu; Kurita, Tsutomu; Iida, Masayoshi*; et al.
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 8 Pages, 2017/04
TRP has stored the plutonium in solution state for long-term since the last PCDF operation in 2007 was finished. After the great east Japan earthquake in 2011, JAEA had investigated the risk against potential hazard of these solutions which might lead to make hydrogen explosion and/or boiling of the solution accidents with the release of radioactive materials to the public when blackout. To reduce the risk for storing Pu solution (about 640 kg Pu), JAEA planned to perform the process operation for the solidification and stabilization of the solution by converted into MOX powder at PCDF in 2013. In order to perform PCDF operation without adaption of new safety regulation, JAEA conducted several safety measures such as emergency safety countermeasures, necessary security and safeguards (3S) measures with understanding of NRA. As a result, the PCDF operation had stared on 28th April, 2014, and successfully completed to convert MOX powder on 3rd August, 2016 for about 2 years as planned.
Kudo, Atsunari; Kurabayashi, Kazuaki; Yanagibashi, Futoshi; Sasaki, Shunichi; Sato, Takehiko; Fujimoto, Ikuo; Obu, Tomoyuki
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 6 Pages, 2017/04
The Co-processing process is the extraction process to recover Pu/U mixed product solution with given Pu/U ratio for improving of nuclear proliferation resistance. In addition, Np is also recovered with U and Pu because Np is one of minor actinides and a long-lived radionuclide and Np has the extractability into TBP solvent. Development of its flowsheet achieves to decrease environmental effect of waste materials. The orientation of development about Co-processing process is to demonstrate of reprocessing the future spent fuels from a LWR, a LWR-MOX hybrid, and a FR-MOX with one cycle. We demonstrated by use of miniature reflux-type centrifugal contactors at the partitioning unit. The test conditions of the Pu/U ratio in the loaded solvents were 1%, 3%, and 5% considering the composition of spent fuels. We used the HAN as the reductant of Np (VI) for back extraction. The results of these tests were very good. We got the prospect of U, Pu, and Np Co-processing flowsheet.
Ohgama, Kazuya; Ikeda, Kazumi*; Ishikawa, Makoto; Kan, Taro*; Maruyama, Shuhei; Yokoyama, Kenji; Sugino, Kazuteru; Nagaya, Yasunobu; Oki, Shigeo
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 Pages, 2017/04
Ohgama, Kazuya; Aliberti, G.*; Stauff, N. E.*; Oki, Shigeo; Kim, T. K.*
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 8 Pages, 2017/04
Stauff, N. E.*; Ohgama, Kazuya; Aliberti, G.*; Oki, Shigeo; Kim, T. K.*
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 Pages, 2017/04
Ohgama, Kazuya; Ota, Hirokazu*; Ikusawa, Yoshihisa; Oki, Shigeo; Ogata, Takanari*
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 6 Pages, 2017/04
Imaizumi, Yuya; Kamiyama, Kenji; Matsuba, Kenichi; Isozaki, Mikio; Suzuki, Toru; Emura, Yuki
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 5 Pages, 2017/04
In order to simulate the typical accident conditions of the fuel relocation phase in SFRs, the molten alloy of low melting point was discharged into a shallow water pool. The distance between the nozzle exit and the bottom plate was set to a value which was indicated to be insufficient to fragment. As the experimental result, the melt jet reached the bottom plate, and dispersed in all directions along the plate, together with the progress of fragmentation. In addition, the melt temperature on the bottom plate decreased rapidly along the radius direction. These results suggest that the fragmentation which would accompany this rapid cooling would be enhanced by the plate. This enhancement would be caused by the extension of the melt-water interface when the melt was dispersed forcibly by the plate. The solidified debris remained after the discharge showed remarkable fragmentation which was assumed to be caused by the formations of small vapor bubbles in the interface.
Takeuchi, Masayuki; Sano, Yuichi; Watanabe, So; Nakahara, Masaumi; Aihara, Haruka; Kofuji, Hirohide; Koizumi, Tsutomu; Mizuno, Tomoyasu
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 6 Pages, 2017/04
Fujimura, Koji*; Itooka, Satoshi*; Oki, Shigeo; Takeda, Toshikazu*
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 6 Pages, 2017/04
Oki, Shigeo; Maruyama, Shuhei; Chikazawa, Yoshitaka; Ohtaki, Akira; Kubo, Shigenobu; Hibi, Koki*; Kan, Taro*
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 9 Pages, 2017/04
Kan, Taro*; Ogura, Masashi*; Hibi, Koki*; Oki, Shigeo; Maeda, Seiichiro; Maruyama, Shuhei; Ohgama, Kazuya
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 Pages, 2017/04
Arai, Masanobu; Kunogi, Kosuke; Aizawa, Kosuke; Chikazawa, Yoshitaka; Takaya, Shigeru; Kubo, Shigenobu; Kotake, Shoji*; Ito, Takaya*; Yamaguchi, Akira*
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 6 Pages, 2017/04
Applications for maintenance program on piping support of prototype fast breeder reactor Monju are studied. Based on degradation mechanism, snubbers in Monju primary cooling system showed lifetime more than the plant lifetime of 30 years by experiments conservatively. For the first step during construction, visual inspection on accessible all supports could be available. In that visual inspection, mounting conditions and damages of all accessible supports could be monitored. One of major features of the Monju primary piping system is large thermal expansion due to large temperature difference between maintenance and operation conditions. Thanks to that large thermal expansion, integrity of the piping support could be monitored by measuring piping displacement. When technologies of piping displacement monitoring are matured in Monju, visual inspection on piping support could be shifted to piping displacement monitoring. At that stage, the visual inspection could be limited only on representative supports.
Chikazawa, Yoshitaka; Kubo, Shigenobu; Shimakawa, Yoshio*; Kaneko, Fumiaki*; Shoji, Takashi*; Nakata, Shuhei*
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 7 Pages, 2017/04
Sodium-cooled reactor (SFR) has superior characteristics thanks to sodium coolant features such as low pressure and high natural convection capability. Involving lessons learned from the 1F accident, requirements on design base DHRS have been modified. In that modification, safety requirements on design extended conditions have been clarified and sodium temperature criteria have been changed taking into account design margin even for design extended conditions. With the new DHRS configuration including ACS, designs of component cooling water system and emergency power supply have been updated.
Honda, Yuki; Sato, Hiroyuki; Ohashi, Hirofumi
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 8 Pages, 2017/04
We have been conducting a source term evaluation method development for high temperature gas-cooled reactors considering structural failures in the major components. This paper present the results of transient analysis for depressurized loss-of-forced cooling accident with ruptures of the cross cut ducts and standpipe, which may be initiated by earthquake. The sequences accounts failures of mitigation systems such as core heat removal by Vessel Cooling System (VCS) and reactor shut down by control rod systems. We will show the effect of mitigation system failure to depressurized loss-of-forced cooling accident in the view point of fuel temperature and natural circulation flow rate which is important for source term evaluation. The major findings obtained in this study showed that multiple failures in mitigation systems for a representative HTGR plant do not aggravate the accident. The result demonstrated that a simplification of event sequence analysis and source term analysis can be achieved with a design fully utilizing the safety characteristics of HTGR.
Hanari, Toshihide; Takebe, Toshihiko*; Yamada, Tomonori; Daido, Hiroyuki; Ishizuka, Ippei*; Omori, Shinya*; Kurosawa, Koichi*; Sasaki, Go*; Nakada, Masahiro*; Sakai, Hideaki*
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 3 Pages, 2017/04
In decommissioning of Fukushima Daiichi Nuclear Power Station, a retrieval process of fuel debris in the Primary Containment Vessel by a remote operation is one of the key issues. In this process, prevention of spreading radioactive materials is one of the important considerations. Furthermore, an applicable technique to the process requires keeping of reasonable processing-efficiency. We propose to use the combined technique including a laser light and a water jet as a retrieval technique of the fuel debris. The laser processing technique combined with a repetitive pulsed water jet could perform an efficient retrieval processing. Our experimental result encourages us to promote further development of the technique towards a real application at Fukushima Daiichi Nuclear Power Station.
Takeuchi, Tomoaki; Tsuchiya, Kunihiko; Komanome, Hirohisa*; Miura, Kuniaki*; Ishihara, Masahiro
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 6 Pages, 2017/04
After the accident at the Fukushima Dai-ichi (1F) Nuclear Power Plant (NPP), the Japanese Government referred to "Enhancement of instrumentation to identify the status of the reactors and PCVs", in the report of Japanese government to the IAEA ministerial conference in June 2011. In response to these provisions, a research and development of a monitoring system for NPPs situations during severe accidents started in November 2012. The objectives of the R&D are composed of radiation-resistant monitoring camera, radiation-resistant in-water transmission system, and heat-resistant signal cable. For all the three objectives, the elemental technologies have been already developed and now trial system are being fabricated and tested under simulated conditions of severe accidents. The results will enable us to determine the basic specifications of the systems and to provide the information about application limits for users.
Tada, Kenichi; Suyama, Kenya
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 4 Pages, 2017/04
JAEA provides the evaluated nuclear data library JENDL. Usually, the integral experiments are used for the validation. Since this validation process takes long time and much effort, the automated system has been desired. To realize the automated system, nuclear data processing, analysis of the integral experiments and editing calculation results are required. With regard to the nuclear data processing, JAEA has started to develop the new nuclear data processing system FRENDY. Using FRENDY, the nuclear data can be automatically processed. Taking advantage of FRENDY, we developed the automatic nuclear data validation system VACANCE. VACANCE has many functions, e.g., searching and modifying the input file, available for the parallel computation and restart calculation, editing the calculation results. Combination of FRENDY and VACANCE enables us to carry out the efficient nuclear data validation cycle. In this presentation, the outline and functions of VACANCE are demonstrated.