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Shimizu, Atsushi; Sato, Hiroyuki; Nakagawa, Shigeaki; Ohashi, Hirofumi
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 8 Pages, 2017/04
Present paper provides an approach to update PRA parameters using the operational and maintenance experience obtained from the HTTR. Firstly, components subject to investigation are selected with the following criteria; The component has safety function in commercial HTGR, the component is utilized in high temperature-irradiated condition, structure or mechanism of the action for the component is unique, and the component is installed in the HTTR. Secondly, component boundaries are clarified and raw data is collected from maintenance records, monthly surveillance test records, operation and maintenance database, etc. As a preliminary study, selected PRA parameters are updated using Bayesian methods to confirm the effectiveness of the use of the HTTR experience. The results showed that the use of HTTR operational and maintenance data is effective for HTGR reliability database development.
Sato, Hiroyuki; Nishida, Akemi; Ohashi, Hirofumi; Muramatsu, Ken*; Muta, Hitoshi*; Itoi, Tatsuya*; Takada, Tsuyoshi*; Hida, Takenori*; Tanabe, Masayuki*; Yamamoto, Tsuyoshi*; et al.
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 7 Pages, 2017/04
JAEA, in conjunction with Tokyo City University, The University of Tokyo and JGC Corporation, have started development of a PRA method considering the safety and design features of HTGR. The primary objective of the project is to develop a seismic PRA method which enables to provide a reasonably complete identification of accident scenario including a loss of safety function in passive system, structure and components. In addition, we aim to develop a basis for guidance to implement the PRA. This paper provides the overview of the activities including development of a system analysis method for multiple failures, a component failure data using the operation and maintenance experience in the HTTR, seismic fragility evaluation method, and mechanistic source term evaluation method considering failures in core graphite components and reactor building.
Fukano, Yoshitaka
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 9 Pages, 2017/04
Loss-of-heat-removal-system (LOHRS) events are identified as some of most dominant severe accident sequences in a sodium-cooled fast reactor. Safety margins against significant core damage in LOHRS events were therefore studied in this paper assuming large fuel-cladding gap and fuel cladding failure. It was clarified through analyses by the developed code that neither fuel melting nor further mechanical pin failure occurs owing to large fuel-cladding gap and fuel cladding failure. It was therefore concluded that large safety margins against significant core damage are provided during LOHRS events. These results will be effectively used in formulating the safety criteria for severe accidents or beyond-design-basis-accidents as one of the supporting evidences to be seriously considered.
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
Kotake, Shoji*; Chikazawa, Yoshitaka; Takaya, Shigeru; Otaka, Masahiko; Kubo, Shigenobu; Arai, Masanobu; Kunogi, Kosuke; Ito, Takaya*; Yamaguchi, Akira*
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 6 Pages, 2017/04
A maintenance management required to prototype nuclear power reactors is proposed. Monitoring and control of sodium impurity and thermal transient are extremely important for sodium boundary maintenance for sodium-cooled fast reactors. At the fast stage of the prototype reactor Monju operation, degradation mechanism on the piping should be demonstrated based on operation experiences. Therefore inspection on a representative position for crack indication and pipe thickness is proposed. Due to less experience of SFR plants, early detection of boundary failure is considered. For a matured operation stage, when degradation mechanism is well demonstrated based on inspection data, inspection cycle could be extended. And for commercial reactors, maintenance without inspection will be established based on accumulated operation experiences including those of the prototype reactor Monju.
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.
Irisawa, Keita; Taniguchi, Takumi; Namiki, Masahiro; Garc
a-Lodeiro, I.*; Osugi, Takeshi; Sakakibara, Tetsuro; Nakazawa, Osamu; Meguro, Yoshihiro; Kinoshita, Hajime*
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 6 Pages, 2017/04
A solidification technique with minimized water content is being developed using phosphate cements for the safe storage of secondary radioactive wastes in the Fukushima Daiichi Nuclear Power Plant. Conventional cement systems become solidified via hydration reactions, and need a certain water content. Phosphate cement systems, however, become solidified via an acid-base reaction, and so they only require water mainly for reasons of workability. A reduced water content of phosphate cement systems is beneficial for the immobilization of the radioactive wastes from mitigating the potential to generate hydrogen gas by the radiolysis of water by radioactive wastes. The current study investigated the water content and mineralogy of calcium aluminate cement (CAC) and phosphate-modified CAC (CAP) cured in open systems at 60, 90 and 120 
C and in a closed system at 20 C as a reference case. Water contents in both the CAC and the CAP were seen to decrease as curing progressed. For 
90 C, the CAP contained less water than CAC. Free water in CAC converted to structural water by heat treatment, but this was not the case for CAP. An orthophosphate hydrate salt, a precursor phase of hydroxyapatite, was found in CAP when cured at 20 and 60 C, and a mixture of the orthophosphate hydrate salt and hydroxyapatite, Ca
(PO
)
(OH)
, were formed in the CAP when cured at 90 C. Phosphate products in CAP cured at 120 C appears to consist of a different phosphate phase compared with the CAP cured at 20, 60 and 90 C.
Uwaba, Tomoyuki; Yano, Yasuhide; Otsuka, Satoshi; Naganuma, Masayuki; Tanno, Takashi; Oka, Hiroshi; Kato, Shoichi; Kaito, Takeji; Ukai, Shigeharu*; Kimura, Akihiko*; et al.
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 7 Pages, 2017/04
Tolerance of fast rector fuel elements to failure in the typical accident conditions was evaluated for the oxide-dispersion-strengthened (ODS) ferritic steel claddings that are candidate of the cladding material for advanced fast reactors. The evaluation was based on the cladding creep damage, which was quantified by the cumulative damage fractions (CDFs). It was shown that the CDFs of the ODS ferritic steel cladding were substantially lower than the breach limit of 1.0 in the loss of flow and transient over power conditions until a passive reactor shutdown system operates.
Kaji, Yoshiyuki; Katsuyama, Jinya; Yamaguchi, Yoshihito; Nemoto, Yoshiyuki; Osaka, Masahiko
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 6 Pages, 2017/04
To investigate the inhomogeneous temperature and stress distribution by geometrical complex of BWR lower head, the detailed 3D model of RPV lower head with control rod guide tubes and shroud supports are constructed and the 3D thermal hydraulic analysis of simulated molten pool and creep deformation analysis of lower head are performed using ANSYS Fluent / Mechanical finite element code. It is found that failure for BWR lower head might be caused by combination between melting failure in inner surface of lower head and creep failure in outer surface of lower head.
Kawabata, Kuniaki; Tanifuji, Yuta; Mori, Fumiaki; Shirasaki, Norihito
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 4 Pages, 2017/04
This paper describes to develop test methods for evaluation of remotely operated robots and operator proficiency for nuclear emergency response and decommissioning tasks. We summarized representative robot's behaviors in the actual operations by the time analysis approach. We also examined environmental factors from the view point of the operation efficiency. Based on these examinations, we currently design some modules of the field for testing remotely operated machines. The approach and progress of the test method development are reported.
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.
Yamashita, Susumu; Uesawa, Shinichiro; Yoshida, Hiroyuki
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 Pages, 2017/04
In order to precisely and unified investigate molten core relocation behavior in the Fukushima Daiichi Nuclear Power Station, we have developed the detailed and phenomenological numerical simulation code named JUPITER for predicting the molten core behavior including solidification and relocation based on the three-dimensional multiphase thermal-hydraulic simulation models. At the moment, the fundamental framework for melt relocation behavior and natural convection around accumulated debris in the JUPITER has been developed, that is, core melt and its relocation, corium spreading behavior in a pedestal region, simulation of air cooling evaluation method. In this paper, preliminary analyses, e.g., numerical simulation for core melt relocation behavior, corium spreading behavior and air cooling analysis around debris are shown.
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.
Hamamoto, Shimpei; Takada, Shoji
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 4 Pages, 2017/04
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.