Tsuruga Comprehensive Research and Development Center
JAEA-Technology 2019-007, 159 Pages, 2019/07
This report summarizes the history and achievements of the prototype fast breeder reactor Monju. The development of Monju started in 1968 as a prototype reactor following the experimental fast reactor Joyo. The development covers all the activity related to the fast reactor; plant design, mockup tests, construction, operation, and plant management. This report summarizes the history and achievements for 11 technical areas: history and principal achievements, design and construction, operation test, plant safety, core physics, fuel, plant system, sodium technology, materials and mechanical design, plant management, and trouble management.
Wakai, Takashi; Machida, Hideo*; Arakawa, Manabu*; Yanagihara, Seiji*; Suzuki, Ryosuke*; Matsubara, Masaaki*
Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 9 Pages, 2018/07
This paper studies crack opening displacement (COD) evaluation methods used in Leak-Before-Break (LBB) assessment of Sodium cooled Fast Reactor (SFR) pipe. For SFR pipe, the continuous leak monitoring will be adopted as an alternative to a volumetric test of the weld joints under conditions that satisfy LBB. The sodium pipes are made of ASME Gr.91 (modified 9Cr-1Mo steel). Thickness of the pipes is small, because the internal pressure is very small. Modified 9Cr-1Mo steel has a relatively large yield stress and small work hardening coefficient comparing to the austenitic stainless steels which are currently used in the conventional plants. In order to assess the LBB behavior of the sodium pipes made of modified 9Cr-1Mo steel, the coolant leak rate from a through wall crack must be estimated properly. Since the leak rate is strongly related to the crack opening displacement (COD), an appropriate COD assessment method must be established to perform LBB assessment. However, COD assessment method applicable for SFR pipes - having thin wall thickness and made of small work hardening material - has not been proposed yet. Thus, a COD assessment method applicable to such a pipe was proposed in this study. In this method, COD was calculated by classifying the components of COD; elastic, local plastic and fully plastic. In addition, the verification of this method was performed by comparing with the results of a series of four-point bending tests using modified 9Cr-1Mo steel pipe having a circumferential through wall notch. As a result, in some cases, COD were over-estimated especially for large cracks. Although the elastic component of COD is still over-estimated for large cracks, leak evaluation from small cracks is much more important in LBB assessment. Therefore, this study recommends that only the elastic component of COD should be adopted in LBB assessment of SFR pipes.
Uranium; Safety, Resources, Separation and Thermodynamic Calculation, p.22 - 48, 2018/05
Safety and economics of uranium utilization for nuclear power generation were investigated and discussed. In order to sustain energy supply with nuclear power generation, uranium resources should be abundant. From the viewpoint of depletion of the resources, FBR, which is breeder reactor of plutonium, has been developed, but that has been not diffused as a commercial reactor yet. Instead of obtaining inexhaustible resources by breeding plutonium, it is known that the inherent safety feature becomes weak in the fast neutron spectrum. As the result of the investigation, it is confirmed with concrete FBR designs that the inherent safety feature and breeding ability are related to the transactions. The amount of uranium resources and electricity generation cost with the resources were investigated. It is concluded that the semi-permanently sustainable energy supply can be established with reasonable cost by using seawater uranium. In addition, the significance of P&T, which is one of the advantages of FBR, was also discussed from the viewpoint of environmental burden from radioactive waste.
Wakai, Takashi; Machida, Hideo*; Arakawa, Manabu*; Kikuchi, Koichi*
Proceedings of ASME Symposium on Elevated Temperature Applications of Materials for Fossil, Nuclear, and Petrochemical Industries, 7 Pages, 2018/04
A simplified J-integral evaluation method applicable to unstable failure analysis in Leak Before Break (LBB) assessment of Sodium-cooled Fast Reactor (SFR) in Japan was proposed. Mod.9Cr-1Mo steel is supposed to be a candidate material for the coolant systems of SFR in Japan. This steel has relatively high yield strength and poor fracture toughness comparing to those of conventional austenitic stainless steels. In addition, SFR pipe has small thickness and large diameter. As a J-integral evaluation method for circumferential through-wall crack in a cylinder, EPRI has proposed a fully plastic solution method. However, the geometry of SFR pipe and material characteristics of Mod.9Cr-1Mo steel exceed the applicable range of EPRI's method. Therefore, a series of elastic, elasto-plastic and plastic finite element analyses (FEA) were performed for a pipe with a circumferential through-wall crack to propose a J-integral evaluation method applicable to such loading conditions. J-integrals obtained from the FEA were resolved into elastic, local plastic and fully plastic components. Each component was expressed as a function of analytical parameter, such as pipe geometries, crack size, material characteristics and so on. As a result, a simplified J-integral evaluation method was proposed. The method enables to conduct 2 parameter failure analysis using J-integral without any fracture mechanics knowledge.
Takano, Kazuya; Maruyama, Shuhei; Hazama, Taira; Usami, Shin
Proceedings of Reactor Physics Paving the Way Towards More Efficient Systems (PHYSOR 2018) (USB Flash Drive), p.1725 - 1735, 2018/04
Irradiation dependence of the core excess reactivity was investigated for the Monju system startup tests at zero-power carried out in 2010. The excess reactivity basically decreases with the decay of Pu in zero-power operation. However, the excess reactivity little changed in the two month period of the startup tests, which suggests a positive reactivity insertion during the period. The investigated irradiation dependence shows that the positive reactivity increases with reactor operation and mostly saturates by the fission-dose attained during the Monju zero-power operation in a month (10 fissions/cm). The saturated positive reactivity is equivalent to approximately 47% of the initially accumulated self-irradiation damage recovery assuming the defects were recovered by the fission-fragment irradiation in the reactor operation.
Ichikawa, Shoichi; Chiba, Yusuke; Ono, Fumiyasu; Hatori, Masakazu; Kobayashi, Takanori; Uekura, Ryoichi; Hashiri, Nobuo*; Inuzuka, Taisuke*; Kitano, Hiroshi*; Abe, Hisashi*
JAEA-Research 2016-021, 32 Pages, 2017/02
In order to reduce the influence on a plant schedule of the MONJU by the maintenance of dew point hygrometers, The JAEA examined a capacitance type dew point hygrometer as an alternative dew point hygrometer for a lithium-chloride type dew point hygrometer which had been used at the CV-LRT in the MONJU. As a result of comparing a capacitance type dew point hygrometer with a lithium-chloride type dew point hygrometer at the CV-LRT (Atmosphere: nitrogen, Testing time: 24 hours), there weren't significant difference between a capacitance type dew point hygrometer and a lithium-chloride type dew point hygrometer. As a result of comparing a capacitance dew point hygrometer with a high-mirror-surface type dew point hygrometer for long term verification (Atmosphere: air, Testing time: 24 months), the JAEA confirmed that a capacitance type dew point hygrometer satisfied the instrument specification (2.04C) required by the JEAC4203-2008.
Kitano, Akihiro; Takegoshi, Atsushi*; Hazama, Taira
Journal of Nuclear Science and Technology, 53(7), p.992 - 1008, 2016/07
A feedback reactivity measurement technique was developed based on a reactivity model featuring components that depend on the reactivity coefficients, denoted as reactor power (K) and reactor vessel inlet temperature (K). This technique was applied to the feedback reactivity experiment conducted in the Monju system start-up test in May 2010. A thorough evaluation considering all possible biases and uncertainties revealed that the reactivity coefficients can be evaluated with a measurement uncertainty smaller than 3%. The evaluated reactivity coefficients were simulated considering the temperature distribution in the core. The C/E value of K showed good agreement between calculated and measured values within the established uncertainty, and the value of K was consistent with that reported in a previous isothermal temperature coefficient experiment. The measured and calculated fuel subassembly outlet temperatures also agreed well within 0.2C.
Hiroi, Hiroshi*; Arai, Masanobu; Kisohara, Naoyuki
Dai-21-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 3 Pages, 2016/06
The purpose of fast breeder reactors (FBR) and the role of Monju were discussed in Ministry of education, culture, sports science and technology-Japan (MEXT) after the Fukushima NNP accident. The discussion has concluded that FBRs contribute to energy security and reduction of high-level radioactive waste, and that Monju is to be utilized to demonstrate these usefulness and to implement international contributions. This paper addresses anticipated R&D results from Monju on the basis of the enforcement of new nuclear regulation, the energy situations in Japan and the international status of FBR development and collaborations.
Sector of Fast Reactor Research and Development
JAEA-Evaluation 2015-005, 77 Pages, 2015/09
Japan Atomic Energy Agency (JAEA) asked the advisory committee "Evaluation Committee of Research and Development Activities for Fast Reactor Cycle" (the Committee) to assess "R&D Programs on FBR/FR Cycle Technologies" and "R&D Programs on Prototype Fast Breeder Reactor Monju and its Related Activities" during the period between FY2010 and FY2014, in accordance with "General Guideline for the Evaluation of Government R&D Activities" by Japanese Cabinet Office, "Guideline Evaluation of R&D in Ministry of Education, Culture, Sports, Science and Technology" and "Regulation on Conduct for Evaluation R&D Activities" by JAEA. This report summarizes results of proposal by the Committee.
Kato, Shinya; Shimomoto, Yoshihiko; Kato, Yuko; Kitano, Akihiro
JAEA-Technology 2014-043, 36 Pages, 2015/02
The core management and operation code system aims to perform core management task efficiently by systematic management of data, analyses and edits, which are needed in the reactor core management and operation. The system consists of the five calculation modules: the reactor constant generation module, the neutronic-thermal calculation module, the radiation analysis module, the core structural integrity estimation module, and the core operation analysis module. In these modules, the neutronic-thermal calculation module is based on the dedicated three-dimensional diffusion and burn-up code HIZER. HIZER can execute core calculations easily for specific design specification and operation patterns of Monju, enabling efficient and accurate evaluation of the Monju core characteristics. This report describes its calculation method and validation results.
Advisory Committee on Monju Safety Requirements
JAEA-Evaluation 2014-005, 275 Pages, 2014/11
In July 2013, Nuclear Regulation Authority (NRA) has enforced new regulatory requirements in consideration of severe accidents for the commercial light water reactors (LWR) and also prototype power generation reactors such as the sodium-cooled fast reactors (SFR) of "Monju" based on TEPCO Fukushima Daiichi Nuclear Power Plant accident occurred in March 2011. Although the regulatory requirements for SFR will be revised by NRA with consideration for public comments, Japan Atomic Energy Agency (JAEA) set up "Advisory Committee on Monju Safety Concept" consists of fast breeder reactor (FBR) and safety assessment experts in order to establish original safety requirements expected to prototype FBR "Monju" considering severe accidents with knowledge from JAEA as well as scientific and technical insights from the experts. This report summarizes the safety requirements expected to Monju discussed by the committee.
Fast Breeder Reactor Research and Development Center, Tsuruga Head Office
JAEA-Review 2014-030, 138 Pages, 2014/08
The prototype fast breeder reactor Monju has accumulated technical achievements in order to establish the fast breeder reactor cycle technology in Japan using the operation and maintenance experience, etc. This annual report summarizes the primary achievements and the data related to the plant management in Monju during fiscal 2013.
Konishi, Satoshi*; Enoeda, Mikio
Purazuma, Kaku Yugo Gakkai-Shi, 90(6), p.332 - 337, 2014/06
Test Blanket Module (TBM) program is to evaluate important functions of prototypical modules of DEMO breeding blankets in the real DT fusion plasma environment of ITER. Therefore, it is regarded as one of the most important milestones toward DEMO blanket. Japan is proposing a Water Cooled Ceramic Breeder (WCCB) TBM as the primary option of TBM program. Japan Atomic Energy Agency (JAEA) is performing the development of the WCCB blanket as the candidate breeding blanket of Japan, with a collaboration of universities and National Institute for Fusion Science (NIFS). Regarding the TBM development, the engineering R and Ds are ongoing, aiming at the demonstration of fabrication technology and structural integrity of the full size mockup of the WCCB TBM. Regarding the test blanket module fabrication technology development, the real scale back wall mockup was successfully fabricated. Also, the design activities are being performed to show the soundness under various loading conditions of electromagnetic force and thermo-mechanical loading. The evaluation of shutdown dose rate behind the TBM test port is also carried out as one of most important design requirement. Furthermore, key technologies toward DEMO blanket, such as, the development of advanced breeder and multiplier pebbles for DEMO blanket, fabrication technology development of Li rich LiTiO pebble and BeTi pebble was performed.
Nakazawa, Tetsuya; Naito, Akira*; Aruga, Takeo; Grismanovs, V.*; Chimi, Yasuhiro; Iwase, Akihiro*; Jitsukawa, Shiro
Journal of Nuclear Materials, 367-370(2), p.1398 - 1403, 2007/08
no abstracts in English
Hoshino, Tsuyoshi; Yasumoto, Masaru*; Tsuchiya, Kunihiko; Hayashi, Kimio; Nishimura, Hidetoshi*; Suzuki, Akihiro*; Terai, Takayuki*
Fusion Engineering and Design, 81(1-7), p.555 - 559, 2006/02
no abstracts in English
Tsuchiya, Kunihiko; Kawamura, Hiroshi; Tanaka, Satoru*
Fusion Engineering and Design, 81(8-14), p.1065 - 1069, 2006/02
no abstracts in English
Sato, Satoshi; Verzilov, Y. M.; Ochiai, Kentaro; Nakao, Makoto*; Wada, Masayuki*; Kubota, Naoyoshi; Kondo, Keitaro; Yamauchi, Michinori*; Nishitani, Takeo
Fusion Engineering and Design, 81(8-14), p.1183 - 1193, 2006/02
no abstracts in English
Olivares, R.*; Oda, Takuji*; Oya, Yasuhisa*; Tanaka, Satoru*; Tsuchiya, Kunihiko
Fusion Engineering and Design, 75-79, p.765 - 768, 2005/11
no abstracts in English
Hoshino, Tsuyoshi; Tsuchiya, Kunihiko; Hayashi, Kimio; Terai, Takayuki*; Tanaka, Satoru*; Takahashi, Yoichi*
Fusion Engineering and Design, 75-79, p.939 - 943, 2005/11
no abstracts in English
Tsuchiya, Kunihiko; Kawamura, Hiroshi; Takayama, Tomoo*; Kato, Shigeru*
Journal of Nuclear Materials, 345(2-3), p.239 - 244, 2005/10
no abstracts in English