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Journal Articles

Evaluation of fixed absorber reactivity measurement in the prototype fast reactor Monju

Ohgama, Kazuya; Katagiri, Hiroki; Takegoshi, Atsushi*; Hazama, Taira

Nuclear Technology, 207(12), p.1810 - 1820, 2021/12

 Times Cited Count:1 Percentile:48.83(Nuclear Science & Technology)

Journal Articles

Investigation of thermal expansion model for evaluation of core support plate reactivity in ATWS event

Sotsu, Masutake

Journal of Energy and Power Engineering, 14(8), p.251 - 258, 2020/08

Thermal expansion behavior was investigated for evaluation of the core support plate expansion reactivity in the Unprotected Loss of Heat Sink reactor trip failure event. A possibility of mechanical restraint was investigated in thermal expansion of the core structure for the prototype fast breeder reactor Monju. The reactor core expansion was simulated in a three-dimensional finite element analysis model of the reactor vessel considering detailed temperature distribution of the sodium coolant based on the thermal-hydraulic analysis result of the whole core model. It was found that the thermal expansion of the core was not restrained in the ULOHS evert, although part of the core structure is mechanically restrained.

Journal Articles

Now is the time of fast reactor

Negishi, Hitoshi; Kamide, Hideki; Maeda, Seiichiro; Nakamura, Hirofumi; Abe, Tomoyuki

Nihon Genshiryoku Gakkai-Shi ATOMO$$Sigma$$, 62(8), p.438 - 441, 2020/08

Prototype Fast Breeder Reactor, Monju, was under decommission since April, 2018. It is the first time for Japan to make a sodium cooled reactor into decommission. It is significant work and will take 30 years. The Monju has provided wide spectrum and huge amount of findings and knowledge, e.g., design, R&D, manufacturing, construction, and operation up to 40% of full power over 50 years of development history. It is significant to utilize such findings and knowledge for the development and commercialization of a fast rector in Japan.

JAEA Reports

Prototype fast breeder reactor Monju; Its history and achievements (Translated document)

Mitsumoto, Rika; Hazama, Taira; Takahashi, Keita; Kondo, Satoru

JAEA-Technology 2019-020, 167 Pages, 2020/03

JAEA-Technology-2019-020.pdf:21.06MB
JAEA-Technology-2019-020-high-resolution1.pdf:47.3MB
JAEA-Technology-2019-020-high-resolution2.pdf:34.99MB
JAEA-Technology-2019-020-high-resolution3.pdf:48.74MB
JAEA-Technology-2019-020-high-resolution4.pdf:47.83MB
JAEA-Technology-2019-020-high-resolution5.pdf:18.35MB
JAEA-Technology-2019-020-high-resolution6.pdf:49.4MB
JAEA-Technology-2019-020-high-resolution7.pdf:39.78MB

The prototype fast breeder reactor Monju has produced valuable technological achievements through design, construction, operation and maintenance over half a century since 1968. This report compiles the reactor technologies developed for Monju, including the areas: history and major achievements, design and construction, commissioning, safety, reactor physics, fuel, systems and components, sodium technology, materials and structures, operation and maintenance, and accidents and failures.

JAEA Reports

Evaluation of decay heat used for effectiveness evaluations of countermeasures against severe accidents in the prototype FBR Monju

Usami, Shin; Kishimoto, Yasufumi*; Taninaka, Hiroshi; Maeda, Shigetaka

JAEA-Technology 2018-003, 97 Pages, 2018/07

JAEA-Technology-2018-003.pdf:12.54MB

The decay heat used for effectiveness evaluation of the prevention measures against severe accidents in the prototype fast breeder reactor Monju was evaluated by applying the updated nuclear data libraries based on JENDL-4.0, reflecting the realistic core operation pattern, and setting the rational extent of uncertainty. The decay heats of fission products, the actinide nuclides such as Cm-242, and radioactive structural materials were calculated by FPGS code. The decay heat of U-239 and Np-239 was evaluated based on ANSI/ANS-5.1-1994. The calculation uncertainty of each decay heat was evaluated based on summation of uncertainty factors, C/E values of reaction rates obtained in Monju system startup test, and so on. Furthermore, the decay heat evaluation method based on the FPGS90 was verified by the comparison of the results of the decay heat measurement of the two spent MOX fuel subassemblies in the experimental fast reactor Joyo MK-II core.

Journal Articles

Evaluation of feedback reactivity coefficients by inverse kinetics in Monju

Kitano, Akihiro; Nakajima, Ken*

Proceedings of 2018 International Congress on Advances in Nuclear Power Plants (ICAPP 2018) (CD-ROM), p.1205 - 1210, 2018/04

The feedback reactivity is taken into account in fast reactor core design especially in order to make the power coefficient negative, which is required to be confirmed in the operation. In the feedback reactivity experiment, the positive reactivity was inserted in the critical state at zero power, and the thermal data, such as reactor power and the R/V inlet temperature, was acquired until the power got stable by the feedback reactivity. In the conventional study, only two critical points in an experiment are available for evaluation of the feedback reactivity coefficients. This method needs three days for evaluation. The advanced method based on the inverse kinetics is newly applied in this work using the more extensive data. It is confirmed that this approach can evaluate the feedback reactivity coefficients in one experiment.

Journal Articles

Prototype fast breeder reactor "Monju" decommissioning and unloading operation of the fuel assembly from the core

Koga, Kazuhiro*; Suzuki, Kazunori*; Hamano, Tomoharu; Takagi, Tsuyohiko

FAPIG, (194), p.6 - 11, 2018/02

The prototype fast breeder reactor "Monju" was decided to decommission by Japanese government on 21 December 2016. After that, Japan Atomic Energy Agency (JAEA) submitted "Monju decommissioning basic plan" to MEXT (Charge ministry) on 13 June 2017, then the unloading operation period (about 5.5 years) of the fuel assembly, which is the first stage of decommission, has started. Fuji Electric is proceeding with various preparations in cooperation with JAEA for the safe work of the unloading operation. This manuscript introduces the outline of preparation situation such as maintenance and inspection for the unloading operation.

JAEA Reports

Verification of alternative dew point hygrometer for CV-LRT in MONJU; Short- and long-term verification for capacitance-type dew point hygrometer (Translated document)

Ichikawa, Shoichi; Chiba, Yusuke; Ono, Fumiyasu; Hatori, Masakazu; Kobayashi, Takanori; Uekura, Ryoichi; Hashiri, Nobuo*; Inuzuka, Taisuke*; Kitano, Hiroshi*; Abe, Hisashi*

JAEA-Research 2017-001, 40 Pages, 2017/03

JAEA-Research-2017-001.pdf:5.19MB

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 ($$pm$$2.04$$^{circ}$$C) required by the JEAC4203-2008.

Journal Articles

Validation of core cooling capability analysis in Monju during guillotine pipe break at primary heat transport system

Yamada, Fumiaki; Arikawa, Mitsuhiro*; Fukano, Yoshitaka

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 7 Pages, 2015/05

In sodium-cooled fast reactor, since the coolant does not need to be pressurized, a pipe break due to the internal pressure does not occur physically. For safety margin in Japanese prototype fast breeder reactor (Monju), the guillotine pipe break accident, i.e., loss of integrity (LOPI) has been analyzed as an extreme assumption for beyond design basis accidents (B-DBAs) in the licensing application for the permit. The cooling capability of the core was re-evaluated in this paper during a large-scale, more specifically guillotine pipe break at the primary heat transport system (PHTS) in Monju, newly considering the following latest findings: (a) Experimental data on sodium boiling in fuel assemblies, (b) Actual PHTS pump coast-down characteristics, and (c) Transient burst test data on irradiated fuel claddings. The analysis models were the validated and simulations were re-performed also using the actual Monju data such as the response time to the trip signals, etc. As a result, it was clarified that the ratio of failed fuel claddings does not exceed around 3% of all of fuel assemblies, as in the past licensing analysis. The safety has been reconfirmed to be secured without significant core damage even under an extreme assumption of a double-ended guillotine pipe break at the PHTS in Monju.

Journal Articles

Water flow simulation test on flow-induced oscillation of thermowell in prototype fast breeder reactor "MONJU"

Kondo, Masaya; Anoda, Yoshinari

JSME International Journal, Series B, 44(4), p.688 - 694, 2001/11

no abstracts in English

JAEA Reports

Water flow simulation of the flow-induced vibration phenomenon of the thermowell in the prototype-FBR Monju

Anoda, Yoshinari; *; ; Murata, Hideo

JAERI-Tech 96-028, 71 Pages, 1996/06

JAERI-Tech-96-028.pdf:3.04MB

no abstracts in English

JAEA Reports

Critical Assembly Experiment on FCA Assembly VI-2

Iijima, Tsutomu; ; ; ; ; ; ; ; *; M.Cho*

JAERI-M 7888, 50 Pages, 1978/10

JAERI-M-7888.pdf:1.1MB

no abstracts in English

Oral presentation

The R&D Programs of Maintenance Technology for FBR Monju Using the Sodium Engineering Research Facility

Saruta, Koichi; Yamaguchi, Toshihiko; Ueda, Masashi

no journal, , 

Oral presentation

Evaluation of double leakage at primary heat transport systems of Monju with passive safety features

Yoshimura, Kazuo; Ikeda, Makinori; Enuma, Yasuhiro; Aizawa, Kosuke

no journal, , 

JAEA has conducted a safety evaluation of impacts of double leakage at the PHTSs considering the passive safety features of Monju as the best-estimate evaluation for a DEC. The result shows that the total amount of leaked sodium can be reduced by the depressurization of the cover gas resulting from decrease in coolant inventory, i.e. negative pressure effects. The reactor coolant level required for decay heat removal, therefore, can be maintained even under double leakage at the PHTSs.

Oral presentation

Trial of domain segmentation of granite based on fracture orientations using Dip Azimuth Vector plot

Terusawa, Shuji*; Shimada, Koji

no journal, , 

The Dip Azimuth Vector plots used in resource exploration was utilized to visualize the dominant fracture direction trends at each depth section of borehole data taken from the Monju site. This procedure is one of the domain analysis method of granite. Domain boundaries are intersections of the two dominant fracture orientations providing pipe-like flow paths of groundwater. So, utilization of geological data accumulated at the Monju in this way, contributes application for geological survey for disposal of HLW as well as ground safety assessment for large earthquakes.

Oral presentation

Current status of Monju decommissioning

Nagaoki, Yoshihiro

no journal, , 

no abstracts in English

Oral presentation

Consultancy Meeting on preparation for the International Workshop on policy and strategy for decommissioning of research reactors & fast reactors and management of spent fuel

Nagaoki, Yoshihiro; Sekine, Takashi

no journal, , 

According to the Backend Roadmap, JAEA is proceeding with the decommissioning of research facilities. Monju is proceeding with defueling work, and Joyo is preparing to restart. Thought Japan is promoting the nuclear fuel cycle, the future power supply has uncertain. Therefore, it is necessary to strategically manage the spent fuel of the fast reactor.

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