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JAEA Reports

Evaluation of the minimum critical amount for heterogeneous lattice systems composed of fuel rods utilized in low-power water-moderated research and test reactors by using continuous-energy Monte Carlo code MVP with JENDL-4.0

Yanagisawa, Hiroshi

JAEA-Technology 2021-023, 190 Pages, 2021/11

JAEA-Technology-2021-023.pdf:5.25MB

Computational analyses on nuclear criticality characteristics were carried out for heterogeneous lattice systems composed of water moderator and fuel rods utilized in low-power research and test reactors, in which the depletion of fuel due to burnup is relatively small, by using the continuous-energy Monte Carlo code MVP Version 2 with the evaluated nuclear data library JENDL-4.0. In the analyses, the minimum critical number of fuel rods was evaluated using calculated neutron multiplication factors for the heterogeneous systems of the uranium dioxide fuel rod in the Static Experiments Critical Facility (STACY) and the Tank-type Critical Assembly (TCA), and the uranium-zirconium hydride fuel rod in the Nuclear Safety Research Reactor (NSRR). In addition, six sorts of the ratio of reaction rates, which are components of neutron multiplication factors, were calculated in the analyses to explain the variation of neutron multiplication factors with the ratio of water moderator to fuel volume in a unit fuel rod cell. Those results of analyses are considered to be useful for the confirmation of reasonableness and validity of criticality safety measures as data showing criticality characteristics for water-moderated heterogeneous lattice systems composed of the existing fuel rods in research and test reactors, of which criticality data are not sufficiently provided by the Criticality Safety Handbook.

JAEA Reports

Interim activity status report of "the group for investigation of reasonable safety assurance based on graded approach" (from September, 2019 to September, 2020)

Yonomoto, Taisuke; Nakashima, Hiroshi*; Sono, Hiroki; Kishimoto, Katsumi; Izawa, Kazuhiko; Kinase, Masami; Osa, Akihiko; Ogawa, Kazuhiko; Horiguchi, Hironori; Inoi, Hiroyuki; et al.

JAEA-Review 2020-056, 51 Pages, 2021/03

JAEA-Review-2020-056.pdf:3.26MB

A group named as "The group for investigation of reasonable safety assurance based on graded approach", which consists of about 10 staffs from Sector of Nuclear Science Research, Safety and Nuclear Security Administration Department, departments for management of nuclear facility, Sector of Nuclear Safety Research and Emergency Preparedness, aims to realize effective graded approach (GA) about management of facilities and regulatory compliance of JAEA. The group started its activities in September, 2019 and has had discussions through 10 meetings and email communications. In the meetings, basic ideas of GA, status of compliance with new regulatory standards at each facility, new inspection system, etc were discussed, while individual investigation at each facility were shared among the members. This report is compiled with expectation that it will help promote rational and effective safety management based on GA by sharing contents of the activity widely inside and outside JAEA.

JAEA Reports

Outline of Regional Workshops held in 2006 - 2017 by the International Atomic Energy Agency in the proposal of Nuclear Emergency Preparedness Group of the Asian Nuclear Safety Network

Okuno, Hiroshi; Yamamoto, Kazuya

JAEA-Review 2020-066, 32 Pages, 2021/02

JAEA-Review-2020-066.pdf:3.01MB

The International Atomic Energy Agency (abbreviated as IAEA) has been implementing the Asian Nuclear Safety Network (abbreviated as ANSN) activities since 2002. As part of this effort, Topical Group on Emergency Preparedness and Response (abbreviated as EPRTG) for nuclear or radiation disasters was established in 2006 under the umbrella of the ANSN. Based on the EPRTG proposal, the IAEA conducted 23 Asian regional workshops in the 12 years from 2006 to 2017. Typical topical fields of the regional workshops were nuclear emergency drills, emergency medical care, long-term response after nuclear/radiological emergency, international cooperation, national nuclear disaster prevention system. The Japan Atomic Energy Agency has produced coordinators for EPRTG since its establishment and has led its activities since then. This report summarizes the Asian regional workshops conducted by the IAEA based on the recommendations of the EPRTG.

JAEA Reports

Introduction of a new framework of safety, maintenance and quality management activities in Japan Atomic Energy Agency under the new nuclear regulatory inspection system since FY 2020

Sono, Hiroki; Sukegawa, Kazuhiro; Nomura, Norio; Okuda, Eiichi; Study Team on Safety and Maintenance; Study Team on Quality Management; Task Force on New Nuclear Regulatory Inspection Systems

JAEA-Technology 2020-013, 460 Pages, 2020/11

JAEA-Technology-2020-013.pdf:13.46MB

Japan Atomic Energy Agency (JAEA) has completed the introduction of a new frame work of safety, maintenance and quality management activities under the new acts on the Regulation of nuclear source material, nuclear fuel material and reactors since April 2020, in consideration of variety, specialty and similarity of nuclear facilities of JAEA (Power reactor in the research and development stage, Reprocessing facility, Fabrication facility, Waste treatment facility, Waste burial facility, Research reactor and Nuclear fuel material usage facility). The JAEA task forces on new nuclear regulatory inspection systems prepared new guidelines on (1) Safety and maintenance, (2) Independent inspection, (3) Welding inspection, (4) Free-access response, (5) Performance indicators and (6) Corrective action program for the JAEA's nuclear facilities. New Quality management systems and new Safety regulations were also prepared as a typical pattern of these facilities. JAEA will steadily improve these guidelines, quality management systems and safety regulations, reviewing the official activities under the new regulatory inspection system together with the Nuclear Regulation Authority and other nuclear operators.

Journal Articles

Considerations on phenomena scaling for BEPU

Nakamura, Hideo

Proceedings of ANS International Conference on Best Estimate Plus Uncertainties Methods (BEPU 2018) (USB Flash Drive), 8 Pages, 2018/00

no abstracts in English

Journal Articles

Investigation of countermeasure against local temperature rise in vessel cooling system in loss of core cooling test without nuclear heating

Ono, Masato; Shimizu, Atsushi; Kondo, Makoto; Shimazaki, Yosuke; Shinohara, Masanori; Tochio, Daisuke; Iigaki, Kazuhiko; Nakagawa, Shigeaki; Takada, Shoji; Sawa, Kazuhiro

Journal of Nuclear Engineering and Radiation Science, 2(4), p.044502_1 - 044502_4, 2016/10

In the loss of forced core cooling test using High Temperature engineering Test Reactor (HTTR), the forced cooling of reactor core is stopped without inserting control rods into the core and cooling by Vessel Cooling System (VCS) to verify safety evaluation codes to investigate the inherent safety of HTGR be secured by natural phenomena to make it possible to design a severe accident free reactor. The VCS passively removes the retained residual heat and the decay heat from the core via the reactor pressure vessel by natural convection and thermal radiation. In the test, the local temperature was supposed to exceed the limit from the viewpoint of long-term use at the uncovered water cooling tube by thermal reflectors in the VCS, although the safety of reactor is kept. Through a cold test, which was carried out by non-nuclear heat input from gas circulators with stopping water flow in the VCS, the local higher temperature position was specified although the temperature was sufficiently lower than the maximum allowable working temperature, and natural circulation of water had insufficient cooling effect on the temperature of water cooling tube below 1$$^{circ}$$C. Then, a new safe and secured procedure for the loss of forced core cooling test was established, which will be carried out soon after the restart of HTTR.

Journal Articles

Progress of the general control system for the Materials and Life Science Experimental Facility in J-PARC

Sakai, Kenji; Oi, Motoki; Watanabe, Akihiko; Kai, Tetsuya; Kato, Yuko; Meigo, Shinichiro; Takada, Hiroshi

JAEA-Conf 2015-002, p.593 - 598, 2016/02

For safe and stable beam operation, a MLF general control system (GCS) consists of several subsystems such as an integral control, interlock, server, network, and timing distribution systems. Since the first beam injection in 2008, the GCS has operated stably without any serious troubles in comparison with upgrade of target devices for ramping up beam power and increment of user apparatuses year by year. In recent years, however, it has been improved significantly in view of sustainable long-term operation and maintenance. The monitor and operation system of the GCS has been upgraded by changing its framework software to improve potential flexibility in its maintenance. Its interlock system was also modified in accordance with the re-examination of the risk management system of J-PARC. This paper reports recent progress of the MLF-GCS.

Journal Articles

ROSA/LSTF tests and RELAP5 posttest analyses for PWR safety system using steam generator secondary-side depressurization against effects of release of nitrogen gas dissolved in accumulator water

Takeda, Takeshi; Onuki, Akira*; Kanamori, Daisuke*; Otsu, Iwao

Science and Technology of Nuclear Installations, 2016, p.7481793_1 - 7481793_15, 2016/00

AA2016-0048.pdf:5.15MB

 Times Cited Count:1 Percentile:14.27(Nuclear Science & Technology)

JAEA Reports

Collection of measurement data from in-situ experiment for performance confirmation of engineered barrier system at Horonobe Underground Research Laboratory (FY2014)

Nakayama, Masashi; Ono, Hirokazu; Nakayama, Mariko*; Kobayashi, Masato*

JAEA-Data/Code 2015-013, 53 Pages, 2015/09

JAEA-Data-Code-2015-013.pdf:9.78MB
JAEA-Data-Code-2015-013(errata).pdf:0.37MB
JAEA-Data-Code-2015-013-appendix(CD-ROM).zip:5.76MB

The Horonobe Underground Research Laboratory (URL) Project has being pursued by Japan Atomic Energy Agency (JAEA) to enhance the reliability of relevant disposal technologies through investigations of the deep geological environment within the host sedimentary formation at Horonobe, northern Hokkaido. The URL Project consists of two major research areas, "Geoscientific Research" and "Research and Development on Geological Disposal Technologies", and proceeds in three overlapping phases, "Phase I: Surface-based investigations", "Phase II: Investigations during tunnel excavation" and "Phase III: Investigations in the underground facilities", over a period of around 20 years. Phase III investigation was started in 2010 fiscal year. The in-situ experiment for performance confirmation of engineered barrier system (EBS experiment) had been prepared from 2013 to 2014 fiscal year at G.L.-350m gallery, and heating by electric heater in simulated overpack had started in January, 2015. One of objectives of the experiment is acquiring data concerned with Thermal-Hydrological-Mechanical-Chemical (THMC) coupled behavior. These data will be used in order to confirm the performance of engineered barrier system. This report summarizes the measurement data acquired from the EBS experiment. The period of data acquisition is from December, 2014 to March, 2015. It will be periodically published summarized data of EBS experiment.

Journal Articles

Scaling issues for the experimental characterization of reactor coolant system in integral test facilities and role of system code as extrapolation tool

Mascari, F.*; Nakamura, Hideo; Umminger, K.*; De Rosa, F.*; D'Auria, F.*

Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.4921 - 4934, 2015/08

Journal Articles

JSFR design progress related to development of safety design criteria for generation IV sodium-cooled fast reactors, 4; Balance of plant

Chikazawa, Yoshitaka; Kato, Atsushi; Nabeshima, Kunihiko; Otaka, Masahiko; Uzawa, Masayuki*; Ikari, Risako*; Iwasaki, Mikinori*

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 8 Pages, 2015/05

Design study and evaluation for SDC and safety SDG on the BOP of the demonstration JSFR including fuel handling system, power supply system, component cooling water system, building arrangement are reported. For the fuel handling system, enhancement of storage cooling system has been investigated adding diversified cooling systems. For the power supply, existing emergency power supply system has been reinforced and alternative emergency power supply system is added. For the component cooling system and air conditioning, requirements and relation between safety grade components are investigated. Additionally for the component cooling system, design impact when adding decay heat removal system by sea water has been investigated. For reactor building, over view of evaluation on the external events and design policy for distributed arrangement is reported. Those design study and evaluation provides background information of SDC and SDG.

Journal Articles

Investigation of characteristics of natural circulation of water in vessel cooling system in loss of core cooling test without nuclear heating

Takada, Shoji; Shimizu, Atsushi; Kondo, Makoto; Shimazaki, Yosuke; Shinohara, Masanori; Seki, Tomokazu; Tochio, Daisuke; Iigaki, Kazuhiko; Nakagawa, Shigeaki; Sawa, Kazuhiro

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 5 Pages, 2015/05

In the loss of forced core cooling test using High Temperature engineering Test Reactor (HTTR), the forced cooling of reactor core is stopped without inserting control rods into the core and cooling by Vessel Cooling System (VCS) to demonstrate the inherent safety of HTGR be secured by natural phenomena to make it possible to design a severe accident free reactor. In the test, the local temperature was supposed to exceed the limit from the viewpoint of long-term use at the uncovered water cooling tube by thermal reflectors in the VCS, although the safety of reactor is kept. The local higher temperature position was specified although the temperature was sufficiently lower than the maximum allowable working temperature, and natural circulation of water had insufficient cooling effect on the temperature of water cooling tube below 1$$^{circ}$$C. Then, a new safe and secured procedure for the loss of forced core cooling test was established, which will be carried out soon after the restart of HTTR.

Journal Articles

Performance test of HTTR

Nakagawa, Shigeaki; Tachibana, Yukio; Takamatsu, Kuniyoshi; Ueta, Shohei; Hanawa, Satoshi

Nuclear Engineering and Design, 233(1-3), p.291 - 300, 2004/10

 Times Cited Count:6 Percentile:42.52(Nuclear Science & Technology)

The High Temperature Gas-cooled Reactor (HTGR) is particularly attractive due to its capability of producing high temperature helium gas and due to its inherent safety characteristics. The High Temperature Engineering Test Reactor (HTTR), which is the first HTGR in Japan, was successfully constructed at the Oarai Research Establishment of the Japan Atomic Energy Research Institute. The HTTR achieved full power of 30MW at a reactor outlet coolant temperature of about 850$$^{circ}$$C on December 7, 2001 during the "rise-to-power tests". Two kinds of tests were carried out during the "rise-to-power tests". One is commissioning test to get operation permit by the government and another is test to confirm a performance of the reactor, heat exchanger, control system. From the test results of the "rise-to-power tests" up to 30MW, the functionality of the reactor and the cooling system were confirmed, and it was also confirmed that an operation of reactor facility can be performed safely.

JAEA Reports

Probabilistic safety assessment of inflammable gas leakage in the HTTR hydrogen production system (Contract research)

Shimizu, Akira; Nishihara, Tetsuo; Moriyama, Koichi*

JAERI-Tech 2004-051, 69 Pages, 2004/06

JAERI-Tech-2004-051.pdf:4.68MB

HTTR of JAERI will be connected with a hydrogen production system by steam reforming of methane for development of nuclear heat utilization technology. This facility will handle much inflammable gas near the nuclear reactor so that special safety consideration is necessary. This report describes the Probabilistic Safety Assessment (PSA) of inflammable gas leakage in the HTTR hydrogen production system. Vessels and pipes, which contain flammable gas, were divided into several systems. Probability of gas leakage were calculated at all candidate places. As a result of assessment, the counter measures such as double-covered inflammable gas pipes, small diameter instrument pipes, leakage detector and emergency shut off valves, are confirmed to be very effective to minimize the scale of explosion and to prevent the damage on nuclear plant.

JAEA Reports

Research and development on reduced-moderation light water reactor with passive safety features (Contract research)

Iwamura, Takamichi; Okubo, Tsutomu; Akie, Hiroshi; Kugo, Teruhiko; Yonomoto, Taisuke; Kureta, Masatoshi; Ishikawa, Nobuyuki; Nagaya, Yasunobu; Araya, Fumimasa; Okajima, Shigeaki; et al.

JAERI-Research 2004-008, 383 Pages, 2004/06

JAERI-Research-2004-008.pdf:21.49MB

The present report contains the achievement of "Research and Development on Reduced-Moderation Light Water Reactor with Passive Safety Features", which was performed by Japan Atomic Energy Research Institute (JAERI), Hitachi Ltd., Japan Atomic Power Company and Tokyo Institute of Technology in FY2000-2002 as the innovative and viable nuclear energy technology (IVNET) development project operated by the Institute of Applied Energy (IAE). In the present project, the reduced-moderation water reactor (RMWR) has been developed to ensure sustainable energy supply and to solve the recent problems of nuclear power and nuclear fuel cycle, such as economical competitiveness, effective use of plutonium and reduction of spent fuel storage. The RMWR can attain the favorable characteristics such as high burnup, long operation cycle, multiple recycling of plutonium (Pu) and effective utilization of uranium resources based on accumulated LWR technologies.

JAEA Reports

Chaotic behavior in a system simulating the pressure balanced injection system; Analysis of passive safety reactor behavior, JAERI's nuclear research promotion program, H12-012 (Contract research)

Madarame, Haruki*; Okamoto, Koji*; Tanaka, Gentaro*; Morimoto, Yuichiro*; Sato, Akira*; Kondo, Masaya

JAERI-Tech 2003-017, 156 Pages, 2003/03

JAERI-Tech-2003-017.pdf:5.31MB

no abstracts in English

JAEA Reports

Report on the water leakage from instrumentation pipe in JMTR

JMTR Pressure Measurement Pipe Investigation Committee

JAERI-Review 2003-014, 117 Pages, 2003/03

JAERI-Review-2003-014.pdf:27.62MB

On December 10、2002, the leak was found at the pressure measurement pipe attached to the exit pipe of No.1 filing pump of the refining system of a primary cooling system at JMTR in Oarai Research establishment JAERI. Investigation Committee for Water Leakage from Instrumentation Pipe in JMTR was established and organized by specialists from inside and outside JAERI on December 16 and its meeting was held in public 3 times by 6th January, 2003. They investigated the cause and countermeasures of cracks, and also investigated enhancement of safety management. This is the report on the cause and countermeasures of cracks and enhancement of safety management.

JAEA Reports

Study on system layout and component design in the HTTR hydrogen production system (Contract research)

Nishihara, Tetsuo; Shimizu, Akira; Tanihira, Masanori*; Uchida, Shoji*

JAERI-Tech 2002-101, 46 Pages, 2003/01

JAERI-Tech-2002-101.pdf:2.6MB

no abstracts in English

Journal Articles

Design of the ITER tritium plant, confinement and detritiation facilities

Yoshida, Hiroshi; Glugla, M.*; Hayashi, Takumi; L$"a$sser, R.*; Murdoch, D.*; Nishi, Masataka; Haange, R.*

Fusion Engineering and Design, 61-62, p.513 - 523, 2002/11

 Times Cited Count:25 Percentile:83.26(Nuclear Science & Technology)

ITER tritium plant is composed of tokamak fuel cycle systems, tritium confinement and detritation systems. The tokamak fuel cycle systems, composed of various tritium sumsystems such as vacuum vessel cleaning gas processing, tokamak exhaust processing, hydrogen isotope separation, fuel storage, mixing and delivery, and external tritium receiving and long-term storage, has been designed to meet not only ITER operation scenarios but safety requirements (minimization of equipment tritium inventory and reduction of environmental tritium release at different off-normal events and accident scenarios). Multiple confinement design was employed because tritium easily permeates through metals (at $$>$$ 150 $$^{circ}$$C) and plastics (at ambient temperature) and mixed with moisture in room air. That is, tritium process equipment and piping are designed to be the primary confinement barrier, and the process equipments (tritium inventory $$>$$ 1 g) are surrounded by the secondary confinement barrier such as a glovebox. Tritium process rooms, which contains these facilities, form the tertiary confinement barrier, and equipped with emergency isolation valves in the heating ventillation and air conditioning ducts as well as atmosphere detritiation systems. This confinement approach has been applied to tokamak building, tritium building, and hotcell and radwaste building.

Journal Articles

Safety shutdown of the High Temperature Engineering Test Reactor during loss of off-site electric power simulation test

Takeda, Takeshi; Nakagawa, Shigeaki; Homma, Fumitaka*; Takada, Eiji*; Fujimoto, Nozomu

Journal of Nuclear Science and Technology, 39(9), p.986 - 995, 2002/09

 Times Cited Count:4 Percentile:30.46(Nuclear Science & Technology)

no abstracts in English

73 (Records 1-20 displayed on this page)