Koga, Kazuhiro*; Suzuki, Kazunori*; Takagi, Tsuyohiko; Hamano, Tomoharu
FAPIG, (196), p.8 - 15, 2020/02
The prototype fast breeder reactor Monju has already started (from June 2017) the unloading operation period (about 5.5 years: until the end of 2022) of the fuel assembly, which is the first stage of decommission. Among them, the first "Processing of fuel assembly" operation (86 in total) was conducted from August 2018 to January 2019 as the first handling of the fuel assembly. Fuji Electric provided technical support, such as dispatching technicians throughout the period, in cooperation with Japan Atomic Energy Agency for the "Processing of fuel assembly" operation, and contributed to the completion of the operation while experiencing various troubles. This manuscript introduces the contents of the first "Processing of fuel assembly" operation and the overview of the trouble status. This manuscript is a sequel to FAPIG No.194 "Prototype Fast Breeder Reactor Monju Decommissioning and Unloading Operation of the Fuel Assembly from the Core", please refer to it.
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.
Sector of Fast Reactor and Advanced Reactor Research and Development
JAEA-Evaluation 2019-004, 47 Pages, 2019/06
Japan Atomic Energy Agency (hereafter referred to as "JAEA") consulted with the "Evaluation Committee of Research and Development Activities for Fast Reactor Cycle" (hereinafter referred to as "Committee"), which consists of specialists in the fields of the evaluation subjects of fast reactor cycle technologies, for interim assessment of R&D activities of fast reactor cycle in the 3rd Mid- and Long-Term Plan (from April 2015 to March 2022) in accordance with "General Guideline for the Evaluation of Government Research and Development (R&D) Activities" by Cabinet Office, Government of Japan, "Guideline for Evaluation of R&D in Ministry of Education, Culture, Sports, Science and Technology" and Regulation on Conduct for Evaluation of R&D Activities" by JAEA. In response to the JAEA's request, the Committee assessed the R&D program of fast reactor cycle technologies during the period of four years from April 2015 to March 2018. The Committee evaluated the management and R&D activities based on the explanatory documents and oral presentations by JAEA. The results of the evaluation were compiled in assessment report that was organized including the reasons for evaluation and the opinions and recommendations. This report is issued for the purpose of actively disseminate evaluation information to the people of Japan (based on General Guideline), which lists the members of the Committee and outlines the assessment items and the review process for procedure of the assessment. The assessment report which was issued by the Committee is attached.
Journal of Nuclear Engineering and Radiation Science, 5(1), p.011001_1 - 011001_13, 2019/01
Local subassembly faults (LFs) have been considered to be of greater importance in safety evaluation in sodium-cooled fast reactors (SFRs) because fuel elements were generally densely arranged in the subassemblies (SAs) in this type of reactors, and because power densities were higher compared with those in light water reactors. A hypothetical total instantaneous flow blockage at the coolant inlet of an SA (HTIB) gives most severe consequences among a variety of LFs. Although an evaluation on the consequences of HTIB using SAS4A code was performed in the past study, SAS4A code was further developed by implementing analytical model of power control system in this study. An evaluation on the consequences of HTIB in an SFR by this developed SAS4A code clarified that the conclusion in the past study was almost same as that in this study. Furthermore SAS4A code was newly validated using four in-pile experiments which simulated HTIB events. The validity of SAS4A application to safety evaluation on the consequence of HTIB was further enhanced in this study. Thus the methodology of HTIB evaluation was established in this study together with the past study and is applicable to HTIB evaluations in other SFRs.
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.
Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-11) (USB Flash Drive), 12 Pages, 2016/10
Probabilistic and deterministic safety assessments and experimental studies on local fault (LF) propagation in sodium-cooled fast reactors (SFRs) have been performed in many countries because LFs have been historically considered as one of the possible causes of severe accidents. Adventitious fuel pin failures have been considered to be the most dominant initiators of LFs in these probabilistic assessments because of its high frequency of occurrence during reactor operation and possibility of subsequent pin-to-pin failure propagation. The four possible mechanisms of fuel element failure propagation from adventitious fuel pin failure (FEFPA) were identified in the past study. All the mechanisms for FEFPA analysis including thermal, mechanical and chemical propagation were modeled into a safety assessment code which was applicable to arbitrary SFRs. Safety analyses on FEFPA of Japanese experimental fast reactor (JOYO), Japanese prototype fast breeder reactor (Monju), Japanese prototype fast breeder reactor with upgraded reactor core (Upgraded Monju) and Japan sodium-cooled fast reactor (JSFR) were performed using this methodology. Although analytical results were different owing to the different core designs in four SFRs, it was clarified in this study that FEFPA was highly unlikely in these SFRs. These results also suggest future possibility of long-term run-beyond-cladding-breach operation which would enhance the economic efficiency in SFRs.
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.
Morohashi, Yuko; Suzuki, Satoshi
JAEA-Technology 2014-045, 116 Pages, 2015/03
The failed fuel detection and location (FFDL) system collects the tagging gas that migrates into the reactor cover gas from a failed pin. The tagging gas is made of stable isotopes of Kr and Xe. The isotopic composition of the tagging gas can be made specific to each assembly. The assembly containing a failed fuel pin can be identified by analyzing the isotopic composition. The FFDL system is comprised of two tagging gas concentration devices. The concentration rate is designed to be higher than 200. Past examinations demonstrated that the concentration rate meets the requirement with a noble gas concentration of 1ppm. However, the actual noble gas concentration emitted from a failed fuel is assumed to be much lower. In the present study, the performance of FFDL system was investigated by measuring low concentration gas of the actual fuel failure level. As a result, the concentration rate was confirmed to be more than tens of thousands, which sufficiently satisfies the design demand.
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.
JNC-TN1400 2001-014, 437 Pages, 2001/10
no abstracts in English
Ohno, Shuji; Matsuki, Takuo*
JNC-TN9400 2000-106, 132 Pages, 2000/12
Sodium fire analyses were performed on 7 kinds of sodium leak tests using a computer code ASSCOPS which has been developed to evaluate thermal consequences of sodium leak accident in an FBR plant. By the comparison between the calculated and the test results of gas pressure, gas temperature, sodium catch pan temperature, wall temperature, and of oxygen concentration, it was confirmed that the ASSCOPS code and the parameters used in the analysis give valid or conservative results on thermal consequences of sodium leak and fire.
; Ohno, Shuji;
JNC-TN2400 2000-006, 56 Pages, 2000/12
Sodium combustion analyses were performed using ASSCOPS version 2.1 in order to obtain background data for evaluating the validity of the mitigation system against secondary sodium leak of MONJU. The calculated results are summarized as follows. (1)Peak atmospheric pressure 4.3 kPa[gage] (2)Peak floor liner temperature 870C, Maximum thinning of liner 2.6mm (3)Peak hydrogen concentration <2% (4)Peak floor liner temperature in the spilt sodium storage eell 400C , Peak floor concrete temperature in the spilt sodium storage cell 140C.
; ; Ueno, Fumiyoshi; ; ; ;
JNC-TN2400 2000-005, 103 Pages, 2000/12
Inelastic analyses of the floor liner subjected to thermal loading due to sodium leakage and combustion were carried out, considering thinning of the liner plate due to molten salt type corrosion. Because the inelastic strain obtained by the analyses stayed below the ductility limit of the material, mechanical integrity, i.e., there exist no through-wall crack on the floor liner, was confirmed. Partial structural model tests were conducted, with a band of local thinning of the liner plate. Displacements were controlled to give specimens much larger strains than those obtained by the inelastic analyses above. No through-wall crack was observed by these tests. Mechanical integrity of the floor liner was confirmed by these results of the inelastic analyses and the partial structural model tests.
JNC-TN1400 2000-012, 250 Pages, 2000/11
no abstracts in English
JNC-TN8410 2000-015, 7 Pages, 2000/10
Some falsification has been detected in the results of quality control data relating to the diameter of samples of pellets produced in the BNFL's MOX Demonstration Facility (MDF) on September 1999. This document is the outlines of inspection procedure for the MONJU fuel pellet in plutonium fuel center of JNC.