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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

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

ITER magnet systems; From qualification to full scale construction

Nakajima, Hideo; Hemmi, Tsutomu; Iguchi, Masahide; Nabara, Yoshihiro; Matsui, Kunihiro; Chida, Yutaka; Kajitani, Hideki; Takano, Katsutoshi; Isono, Takaaki; Koizumi, Norikiyo; et al.

Proceedings of 24th IAEA Fusion Energy Conference (FEC 2012) (CD-ROM), 8 Pages, 2013/03

The ITER organization and 6 Domestic Agencies (DA) have been implementing the construction of ITER superconducting magnet systems. Four DAs have already started full scale construction of Toroidal Field (TF) coil conductors. The qualification of the radial plate manufacture has been completed, and JA and EU are ready for full scale construction. JA has qualified full manufacturing processes of the winding pack with a 1/3 prototype and made 2 full scale mock-ups of the basic segments of TF coil structure to optimize and industrialize the manufacturing process. Preparation and qualification of the full scale construction of the TF coil winding is underway by EU. Procurement of the manufacturing equipment is near completion and qualification of manufacturing processes has already started. The constructions of other components of the ITER magnet systems are also going well towards the main goal of the first plasma in 2020.

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:57.37(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

Procurement of Nb$$_3$$Sn superconducting conductors in ITER

Nabara, Yoshihiro; Isono, Takaaki; Nunoya, Yoshihiko; Koizumi, Norikiyo; Hamada, Kazuya; Matsui, Kunihiro; Hemmi, Tsutomu; Kawano, Katsumi; Uno, Yasuhiro*; Seki, Shuichi*; et al.

Journal of Plasma and Fusion Research SERIES, Vol.9, p.270 - 275, 2010/08

Journal Articles

Qualification of cryogenic structural materials for the ITER toroidal field coils

Nakajima, Hideo; Takano, Katsutoshi; Tsutsumi, Fumiaki; Kawano, Katsumi; Hamada, Kazuya; Okuno, Kiyoshi

Proceedings of 2009 ASME Pressure Vessels and Piping Division Conference (PVP 2009) (CD-ROM), 9 Pages, 2009/07

The Japan Atomic Energy Agency (JAEA) has been evaluating mechanical properties of structural materials for the ITER toroidal field (TF) coils. Newly developed JJ1 forging having thickness of 400 mm, 316LN forging having thickness of 410 mm, and 316LN hot rolled plate having thickness of 200 mm were produced in mass production process to qualify the materials. The distributions of tensile properties at liquid helium temperature (4K) in products have been evaluated to qualify the materials and it has been demonstrated that these materials have good quality and uniform properties, which satisfy the ITER requirements. It is also demonstrated from the results that temperature dependence of strengths are expressed by quadratic curves developed by JAEA, which are expressed as a function of carbon and nitrogen contents and strengths at room temperature. This equation enables to perform quality control of materials at only room temperature. The results obtained from these activities also serve the basis to develop the material material section of "Codes for Fusion Facilities - Rules on Superconducting Magnet Structure (2008)" issued by the Japan Society of Mechanical Engineers (JSME) in October 2008.

Journal Articles

Installation and test programme of the ITER poloidal field conductor insert (PFCI) in the ITER test facility at JAEA Naka

Nunoya, Yoshihiko; Takahashi, Yoshikazu; Hamada, Kazuya; Isono, Takaaki; Matsui, Kunihiro; Oshikiri, Masayuki; Nabara, Yoshihiro; Hemmi, Tsutomu; Nakajima, Hideo; Kawano, Katsumi; et al.

IEEE Transactions on Applied Superconductivity, 19(3), p.1492 - 1495, 2009/06

 Times Cited Count:1 Percentile:12.13(Engineering, Electrical & Electronic)

The ITER Poloidal Field Conductor Insert (PFCI) was constructed to characterize the performance of selected cable-in-conduit NbTi conductors for the ITER Poloidal Field (PF) under relevant operating conditions. The PFCI was installed and tested inside the bore of the ITER CS model coil, which provides the background magnetic field. The PFCI is a single-layer solenoid, wound from about 50 m of a full-size ITER cable-in-conduit conductor. The winding diameter and height are about 1.5 m and 1 m, respectively. The nominal design current of the conductor is 45 kA at 6 T and 5 K. The main items in the PFCI test programme are current sharing temperature (Tcs) measurements, critical current (Ic) measurements and AC loss measurement. The key technology of the installation, the test methods and procedures, and some preliminary results of the testing campaigns are described and discussed in this paper.

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.

Journal Articles

Development of conduits for the ITER central solenoid conductor

Hamada, Kazuya; Nakajima, Hideo; Kawano, Katsumi; Takano, Katsutoshi; Tsutsumi, Fumiaki; Okuno, Kiyoshi; Fujitsuna, Nobuyuki*; Teshima, Osamu*

Teion Kogaku, 43(6), p.244 - 251, 2008/06

Japan Atomic Energy Agency has developed JK2LB conduit for the Nb$$_{3}$$Sn conductor of the ITER Central Solenoid. Mechanical requirements for the CS conductor conduit are 0.2% yield strength of more than 900 MPa and fracture toughness K $$_{IC}$$(J) of more than 130 MPa$$sqrt{m}$$ after a compaction and aging heat treatment (650 $$^{circ}$$C, 240 hours). In the previous work, aged JK2LB conduit has shown high strength and fracture toughness enough to satisfy the requirements. As a next step, work was performed to determine specification of the JK2LB conduit taking account of cold work including compaction and winding, and to simplify its fabrication process. To simulate the cold work effect and aging, mechanical tests were performed at 4.2 K on laboratory scale (20-30kg) ingot samples. It was found that the sum of carbon and nitrogen content should be in a range from 0.11% to 0.18% to achieve the ITER mechanical requirements. To obtain a grain size of conduit as well as that of small ingot sample, applicable solution heat treatment temperature and holding time were studied. In order to simplify the billet production process, we confirmed internal metallurgical qualities of JK2LB cast ingot. Since significant segregation was not observed, we could exclude an electroslag remelting process. Based on above achievements, full size JK2LB conduits were fabricated and satisfied the ITER mechanical requirements.

Journal Articles

Development of jacketing technologies for ITER CS and TF conductor

Hamada, Kazuya; Nakajima, Hideo; Matsui, Kunihiro; Kawano, Katsumi; Takano, Katsutoshi; Tsutsumi, Fumiaki; Okuno, Kiyoshi; Teshima, Osamu*; Soejima, Koji*

AIP Conference Proceedings 986, p.76 - 83, 2008/03

The ITER Toroidal Field (TF) coil and Central Solenoid (CS) use Nb$$_{3}$$Sn cable-in-conduit conductor. Conductor fabrication process are as follows; (1) Fabrication of jacket. (2) Butt welding of jacket to make a long tube (CS: 880 m, TF: 760 m) and insertion of superconducting cable into jacket. (3) Compaction of jacket. (4) Winding for transportation. JAEA has developed jacketing technologies in the cooperation with industries. Major achievements are as follows; (1) Full scale TF and CS jackets were fabricated using low carbon SUS316LN and boron added and high manganese stainless steel (JK2LB), respectively. The jackets satisfied ITER mechanical and dimensional requirement. (2) Butt welding condition was studied to obtain good internal surface condition of welded joint. (3) Compaction machine was constructed. As results of compaction test of TF and CS jacket, compacted jacket dimensions satisfied ITER requirement. Therefore, JAEA demonstrated jacketing technologies for ITER conductor.

Journal Articles

Demonstration of full scale JJ1 and 316LN fabrication for ITER TF coil structure

Hamada, Kazuya; Nakajima, Hideo; Kawano, Katsumi; Takano, Katsutoshi; Tsutsumi, Fumiaki; Okuno, Kiyoshi

Fusion Engineering and Design, 82(5-14), p.1481 - 1486, 2007/10

 Times Cited Count:14 Percentile:72.17(Nuclear Science & Technology)

no abstracts in English

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