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Journal Articles

Development of safety design technologies for sodium-cooled fast reactor coupled to thermal energy storage system with sodium-molten salt heat exchanger; Project overview

Yamano, Hidemasa; Kurisaka, Kenichi; Takano, Kazuya; Kikuchi, Shin; Kondo, Toshiki; Umeda, Ryota; Shirakura, Shota*

Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2023/09

This project studies investigation on safety design guideline and risk assessment technology for sodium-cooled fast reactor with the molten-salt heat storage system, development of evaluation method for heat transferring performance between sodium and molten-salt and improvement of the performance, and evaluation of chemical reaction characteristic between sodium and molten-salt and improvement of its safety. The project overview is presented in this report.

Journal Articles

Inherent core safety performance of small sodium-cooled fast reactor with oxide fuel

Takano, Kazuya; Oki, Shigeo; Doda, Norihiro; Chikazawa, Yoshitaka; Maeda, Seiichiro

Proceedings of 2023 International Congress on Advanced in Nuclear Power Plants (ICAPP 2023) (Internet), 7 Pages, 2023/04

The MOX fueled SMR-SFRs with lower linear heat rating of 100 W/cm and 50 W/cm, whereas the linear heat rating at rated power is around 400 W/cm in general, were designed to decrease the fuel temperature during its rated power state in order to pursue the inherent core safety for MOX fueled SMR-SFRs. The transient analyses for Anticipated Transient Without Scram (ATWS) events represented by an Unprotected Loss of Flow (ULOF) accident on the lower linear heat rating cores were performed considering their inherent feedback reactivity. Through the transient analysis, the inherent core safety performances for the lower linear heat rating cores were discussed based on the evaluated maximum coolant temperature and Cumulative Damage Fraction (CDF) as criteria to maintain the core and fuel integrity. The feasible design window for MOX fueled SMR-SFRs with the inherent core safety focusing on the linear heat rating was identified based on the transient analysis results.

Journal Articles

Development of safety design technologies for sodium-cooled fast reactor coupled to thermal energy storage system with sodium-molten salt heat exchanger

Yamano, Hidemasa; Kurisaka, Kenichi; Takano, Kazuya; Kikuchi, Shin; Kondo, Toshiki; Umeda, Ryota; Shirakura, Shota*; Hayashi, Masaaki*

Proceedings of 8th International Conference on New Energy and Future Energy Systems (NEFES 2023) (Internet), p.27 - 34, 2023/00

This project studies investigation on safety design guideline and risk assessment technology for sodium-cooled fast reactor with the molten-salt heat storage system, development of evaluation method for heat transferring performance between sodium and molten-salt and improvement of the performance, and evaluation of chemical reaction characteristic between sodium and molten-salt and improvement of its safety. The project overview is presented in this report.

Journal Articles

Core and safety design for France-Japan common concept on sodium-cooled fast reactor

Takano, Kazuya; Oki, Shigeo; Ozawa, Takayuki; Yamano, Hidemasa; Kubo, Shigenobu; Ogura, Masashi*; Yamada, Yumi*; Koyama, Kazuya*; Kurita, Koichi*; Costes, L.*; et al.

EPJ Nuclear Sciences & Technologies (Internet), 8, p.35_1 - 35_9, 2022/12

The France and Japan teams have carried out collaborative works to have common technical views regarding a sodium-cooled fast reactor concept. Japan has studied the feasibility of an enhanced high burnup low-void effect (CFV) core and fuel using oxide dispersion-strengthened steel cladding in ASTRID 600. Regarding passive shutdown capabilities, Japan team has performed a preliminary numerical analysis for ASTRID 600 using a complementary safety device, called a self-actuated shutdown system (SASS), one of the safety approaches of Japan. The mitigation measures of ASTRID 600 against a severe accident, such as a core catcher, molten corium discharge assembly, and the sodium void reactivity features of the CFV core, are promising to achieve in-vessel retention for both countries. The common design concept based on ASTRID 600 is feasible to demonstrate the SFR core and safety technologies for both countries.

Journal Articles

Sodium-cooled Fast Reactors

Ohshima, Hiroyuki; Morishita, Masaki*; Aizawa, Kosuke; Ando, Masanori; Ashida, Takashi; Chikazawa, Yoshitaka; Doda, Norihiro; Enuma, Yasuhiro; Ezure, Toshiki; Fukano, Yoshitaka; et al.

Sodium-cooled Fast Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.3, 631 Pages, 2022/07

This book is a collection of the past experience of design, construction, and operation of two reactors, the latest knowledge and technology for SFR designs, and the future prospects of SFR development in Japan. It is intended to provide the perspective and the relevant knowledge to enable readers to become more familiar with SFR technology.

Journal Articles

Routing study of above core structure with mock-up experiment for ASTRID

Takano, Kazuya; Sakamoto, Yoshihiko; Morohoshi, Kyoichi*; Okazaki, Hitoshi*; Gima, Hiromichi*; Teramae, Takuma*; Ikarimoto, Iwao*; Botte, F.*; Dirat, J.-F.*; Dechelette, F.*

Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 8 Pages, 2019/05

ASTRID has the objective to integrate innovative options in order to prepare the 4th generation reactors. In ASTRID, large number of tubes are installed above each fuel subassembly to monitor the core. These instrumentations such as thermocouples (TC) and Failed Fuel Detection and Location (FFDL) systems are integrated into Above Core Structure (ACS) with various sizes tubes. In the present study, the routing study for TC tubes and FFDL tubes was performed with 3D modeling and mock-up experiment of the ACS designed for ASTRID with 1500 MW thermal power in order to clarify the integration process and secure the design hypotheses. Although some problems on fabricability were found in the mock-up experiment, the possible solutions were proposed. The present study gives manufacturing feedback to design team and will contribute to increase the knowledge for ACS design and fabricability.

Journal Articles

Irradiation induced reactivity in Monju zero power operation

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 $$beta$$ decay of $$^{241}$$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 ($$sim$$10$$^{17}$$ fissions/cm$$^{3}$$). 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.

Journal Articles

IAEA benchmark calculations on control rod withdrawal test performed during Phenix End-of-Life experiments; JAEA's calculation results

Takano, Kazuya; Mori, Tetsuya; Kishimoto, Yasufumi; Hazama, Taira

Proceedings of International Conference on the Physics of Reactors; The Role of Reactor Physics toward a Sustainable Future (PHYSOR 2014) (CD-ROM), 13 Pages, 2014/09

This paper describes details of the IAEA/CRP benchmark calculation by JAEA on the control rod withdrawal test in the Phenix End-of-Life Experiments. The power distribution deviation by the control rod insertion/withdrawal, which is the major target of the benchmark, is well simulated by calculation. In addition to the CRP activities, neutron and photon transport effect is evaluated in the nuclear heating calculation of the benchmark analysis. It is confirmed that the neutron and photon transport effect contributes to the improvement of the absolute power calculation results in the breeder blanket region.

Journal Articles

IAEA benchmark calculations on control rod withdrawal test performed during Phenix End-of-Life experiments; Benchmark results and comparisons

Pascal, V.*; Prulhi$`e$re, G.*; Vanier, M.*; Fontaine, B.*; Devan, K.*; Chellapandi, P.*; Kriventsev, V.*; Monti, S.*; Mikityuk, K.*; Chenu, A.*; et al.

Proceedings of International Conference on the Physics of Reactors; The Role of Reactor Physics toward a Sustainable Future (PHYSOR 2014) (CD-ROM), 16 Pages, 2014/09

no abstracts in English

Journal Articles

Journal Articles

Benchmark calculations on control rod withdrawal tests performed during Phenix End-of-Life experiments

Pascal, V.*; Prulhi$`e$re, G.*; Fontaine, B.*; Devan, K.*; Chellapandi, P.*; Kriventsev, V.*; Monti, S.*; Mikityuk, K.*; Semenov, M.*; Taiwo, T.*; et al.

Proceedings of 2013 International Congress on Advances in Nuclear Power Plants (ICAPP 2013) (USB Flash Drive), 11 Pages, 2013/04

The control rod withdrawal test was one of the various Phenix End-of-Life tests performed in 2009. The main goal was to determine the impact of a rod insertion and/or extraction on the radial power distribution in the fissile core at nominal power. The framework of the Technical Working Group on Fast Reactors (TWG-FR) activities in IAEA, decided to launch a Coordinated Research Project (CRP), devoted to benchmarking analyses on the test. The CRP was performed by experts coming from CEA, ANL, IGCAR, IPPE, IRSN, JAEA, KIT and PSI. After a short description of the test conducted in the Phenix reactor, this paper presents some results obtained in the course of the CRP with special emphasis on control rod efficiencies and power deformation by subassemblies. The paper also discusses the discrepancies found when comparing calculated results with experimental data as well as some preliminary conclusions on the source of these discrepancies.

Journal Articles

Control rod worth evaluation for the Monju restart core

Takano, Kazuya; Fukushima, Masahiro; Hazama, Taira; Suzuki, Takayuki

Nuclear Technology, 179(2), p.266 - 285, 2012/08

 Times Cited Count:9 Percentile:57.16(Nuclear Science & Technology)

The present paper describes the evaluation of the control rod worth data obtained in the Monju restart core. The best-estimate value and its uncertainty are evaluated in detail. As in the criticality evaluation, data obtained in the previous test is evaluated in the same level of detail. The correlation in the uncertainties is also evaluated among different control rods and tests of the previous and the restart cores. Based on the evaluated data, calculation accuracy is investigated with JENDL-3.3 and JENDL-4.0. It is confirmed that the calculation accuracy is within the experimental uncertainty of 2% for each layer and $$^{10}$$B content. A reduction in the uncertainty related to the delayed neutron fraction is effective for a further improvement in the calculation accuracy.

Journal Articles

Adjustment of $$^{241}$$Am cross section with Monju reactor physics data

Hazama, Taira; Takano, Kazuya; Kitano, Akihiro

Proceedings of 2011 International Congress on Advances in Nuclear Power Plants (ICAPP '11) (CD-ROM), p.1527 - 1535, 2011/05

The Japanese prototype fast breeder reactor Monju restarted its reactor physics test in May, 2010 after a 14-year interruption. The accumulation of $$^{241}$$Am due to the $$^{241}$$Pu decay during the interruption reaches 1.5wt% in average. An impact of the reactor physics data obtained in the restart core is investigated by the cross section adjustment technique with JENDL-3.3 and JENDL-4.0. Criticality data obtained before and after the interruption are applied. It is confirmed that Monju reactor physics data, when the two data are used together, effectively adjust $$^{241}$$Am capture cross sections. Consistent results are obtained among JENDL-3.3 after adjustment and JENDL-4.0 before and after the adjustment.

Journal Articles

Monju core physics test analysis with JAEA's calculation system

Takano, Kazuya; Sugino, Kazuteru; Mori, Tetsuya; Kishimoto, Yasufumi*; Usami, Shin

Proceedings of International Conference on the Physics of Reactors, Nuclear Power; A Sustainable Resource (PHYSOR 2008) (CD-ROM), 8 Pages, 2008/09

Monju core physics test analysis was performed using JAEA's neutronics calculation system with various nuclear data libraries (JENDL-3.2, JENDL-3.3, JEFF-3.1, ENDF/B-VII) for the purpose to validate the JAEA's neutronics calculation system, which utilizes JENDL-3.3. Subsequent sensitivity analysis was carried out to clarify the cause of differences in calculation results among nuclear data libraries. It is found that the calculation results obtained by JENDL-3.3 and JAEA's neutronics analysis system showed good agreement with the measured values and its accuracy is identical or better than JEFF-3.1, ENDF/B-VII in most core characteristics. Thus, the validity of JAEA's neutronics analysis system with JENDL-3.3 was confirmed. From the sensitivity analysis, it was identified that Monju can be quite valuable for the verification of the cross sections of such high-order Pu isotopes as $$^{240}$$Pu and $$^{241}$$Pu and also for the validity of temperature dependency of the self-shielding using its property as a power reactor.

Oral presentation

Influence of group constants on the Monju core physics test analysis

Takano, Kazuya; Sato, Wakaei; Usami, Shin; Nishi, Hiroshi

no journal, , 

no abstracts in English

Oral presentation

Development of reactivity feedback effect measurement techniques under sub-critical condition in fast reactor, 1-1; Purpose and program

Okajima, Shigeaki; Nishi, Hiroshi; Yamane, Yoshihiro*; Kanemoto, Shigeru*; Yamane, Tsuyoshi; Mori, Takamasa; Kitamura, Yasunori; Fukushima, Masahiro; Kitano, Akihiro; Ando, Masaki; et al.

no journal, , 

The program was planned and has been carried out to develop of the reactivity feedback effect measurement techniques under sub-critical condition in the start-up experiment for fast reactors. The program includes the demonstration of the developed technique in fast critical assembly (FCA) and the proposal of the appropriate measurement system based on the results. The background and the purposes of the program and the outline of development subjects are presented.

Oral presentation

Prototype FBR Monju system start up test "zero power reactor physics test", 4; Neutron instrumentation soundness measurement

Takano, Kazuya; Miyagawa, Takayuki; Ikegami, Kazunori; Kitano, Akihiro

no journal, , 

no abstracts in English

Oral presentation

Monju core confirmation test, 4; Neutron Instrumentation Soundness Measurement

Takano, Kazuya

no journal, , 

no abstracts in English

Oral presentation

Oral presentation

Results of Monju zero power reactor physics test

Mori, Tetsuya; Takano, Kazuya; Kitano, Akihiro; Morohashi, Yuko; Kato, Yuko; Yabuki, Kentaro; Miyagawa, Takayuki; Okawachi, Yasushi; Hazama, Taira

no journal, , 

Monju restarted safely on May 6, 2010 after 14 years and 5 months suspension. Core Confirmation Test was performed until July 22. The core fuel contains Am-241 because Pu-241 (half-life 14 years) decayed during 14.4 years suspension. Therefore, physics data of the core containing Am-241 are obtained. The mainly test items are criticality, control rod worth and isothermal temperature coefficient. In the criticality, the measured CR position at the criticality was confirmed to be within the predicted CR position range. Criticality was predicted in good accuracy. In the control rod worth measurement, CR worth of CCR1 was measured by the period method. CR worth of other CR was measured by the balancing method. In the isothermal temperature coefficient measurement, the measured value was a little bit smaller than that of the previous test due to the accumulation of Am-241, the decay of Pu-241, and other composition change by refuelling.

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