Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Choi, B.; Nishida, Akemi; Kawata, Manabu; Shiomi, Tadahiko; Li, Y.
JAEA-Research 2024-001, 206 Pages, 2024/03
JAEA-Research-2024-001.pdf:9.12MB
In the assessment of seismic safety and the design of building structures in nuclear facilities, lumped mass models have been used as standard methods. Recent advances in computer capabilities allow the use of three-dimensional finite element (3D FE) models to account for the 3D behavior of buildings, material nonlinearity, and the nonlinear soil-structure interaction effect. While 3D analysis method has many advantages, it is necessary to ensure its reliability as a new approach. The International Atomic Energy Agency performed an international benchmark study using the 3D FE analysis model for reactor building of Unit 7 at TEPCO's Kashiwazaki-Kariwa Nuclear Power Station based on recordings from the Niigataken Chuetsu-oki Earthquake in 2007. Multiple organizations from different countries participated in this study and the variation in their analytical results was significant, indicating an urgent need to improve the reliability of the analytical results by standardization of the analytical methods using 3D FE models. Additionally, it has been pointed out that it is necessary to understand the 3D behavior in the seismic fragility assessment of buildings and equipment, using realistic seismic response analysis method based on 3D FE models. In view of these considerations, a guideline for the seismic response analysis method using a 3D FE model was developed by incorporating the latest knowledge and findings in this area. The purpose of the guideline is to improve the reliability of the seismic response analysis method using 3D FE model of reactor buildings. The guideline consists of a main body, commentaries, and appendixes. The standard procedures, recommendations, key points to note, and technological bases for conducting seismic response analysis on reactor buildings using 3D FE models are provided in the guideline. In addition, the guideline will be revised reflecting the latest knowledge.
Sr/Y and 
Cs radioactivityTerasaka, Yuta; Uritani, Akira*
Nuclear Instruments and Methods in Physics Research A, 1049, p.168071_1 - 168071_7, 2023/04
Times Cited Count：1 Percentile：68.31(Instruments & Instrumentation)Futagami, Satoshi; Kubo, Shigenobu; Sofu, T.*; Ammirabile, L.*; Gauthe, P.*
Proceedings of International Conference on Topical Issues in Nuclear Installation Safety; Strengthening Safety of Evolutionary and Innovative Reactor Designs (TIC 2022) (Internet), 10 Pages, 2022/10
Kato, Shinya; Matsuba, Kenichi; Kamiyama, Kenji; Akaev, A.*; Vurim, A.*; Baklanov, V.*
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-13) (Internet), 12 Pages, 2022/09
The In-Vessel Retention (IVR) of molten-core in Core Disruptive Accidents (CDAs) is of prime importance in enhancing the safety of sodium-cooled fast reactors. One of the main subjects in ensuring IVR is to design the Control Rod Guide Tube (CRGT) which allows effective discharge of molten core materials from the core region. The effectiveness of the CRGT design is assessed through CDA analyses, and it is reasonable for these analyses to develop a computer code collaborated with experimental researches. Thus, experiments addressing the discharge behavior of the molten-core materials through the CRGT have proceeded as one of the subjects in the collaboration research named the EAGLE-3 project, and the obtained experimental results are reflected in the development of the SIMMER code. In this project, a series of out-of-pile tests using molten-alumina as the fuel simulant was conducted to understand the discharge behavior of molten-core materials through the CRGT. In this study, in order to investigate the effect of an internal structure in the CRGT on the discharge behavior of the molten-core materials, the data of an out-of-pile test in which the molten-alumina penetrated to a duct with the internal structure were analyzed. In addition, the post-test analysis using the SIMMER code was conducted and the results were compared with the test results.
Onoda, Yuichi; Uchita, Masato*; Tokizaki, Minako*; Okazaki, Hitoshi*
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 6 Pages, 2022/08
The safety analyses were carried out to confirm the sufficiency of the function of the plant protection system against the pump/diagrid link rupture. The target plant is a pool-type SFR of about 600 MWe class equipped with an axially homogeneous core currently under development in Japan. In the pool-type SFR, the primary system piping connects primary pump and the high-pressure sodium plenum located at the inlet of fuel sub-assemblies and called "pump/diagrid link". Because this piping is submerged in the reactor vessel, it is difficult to detect small scale sodium leakage in this piping, and thus a certain large pipe break like guillotine should be assumed and evaluated as a design basis event. In order to confirm the detectability of pump/diagrid link rupture by safety protection system signals, a series of analyses of the guillotine break for a pump/diagrid link were carried out. Sensitivity study had also been performed to consider the uncertainty of the reactivity coefficient in the analyses. The sufficiency of the function of the plant protection system against the pump/diagrid link rupture was confirmed by the analysis results that at least two signals are transmitted for the detection of the event, which is the development target of the plant protection system in pool-type SFR.
Choi, B.; Nishida, Akemi; Kawata, Manabu; Shiomi, Tadahiko; Li, Y.
JAEA-Research 2021-017, 174 Pages, 2022/03
JAEA-Research-2021-017.pdf:9.33MB
Standard methods such as lumped mass models have been used in the assessment of seismic safety and the design of building structures in nuclear facilities. Recent advances in computer capabilities allow the use of three-dimensional finite element (3D FE) models to account for the 3D behavior of buildings, material nonlinearity, and the nonlinear soil-structure interaction effect. Since the 3D FE model enables more complex and high-level treatment than ever before, it is necessary to ensure the reliability of the analytical results generated by the 3D FE model. Guidelines for assuring the dependability of modeling techniques and the treatment of nonlinear aspects of material properties have already been created and technical certifications have been awarded in domains other than nuclear engineering. The International Atomic Energy Agency performed an international benchmark study in nuclear engineering. Multiple organizations reported on the results of seismic response studies using the 3D FE model based on recordings from the Niigata-ken Chuetsuoki Earthquake in 2007. The variation in their analytical results was significant, indicating an urgent need to improve the reliability of the analytical results by standardization of the analytical methods using 3D FE models. Additionally, it has been pointed out that it is necessary to understand the 3D behavior in the seismic fragility assessment of buildings and equipment, which requires evaluating the realistic nonlinear behavior of building facilities when assessing their seismic fragility. In view of these considerations, a standard guideline for the seismic response analysis method using a 3D FE model was produced by incorporating the latest knowledge and findings in this area. The purpose of the guideline is to improve the reliability of the seismic response analysis method using 3D FE model of reactor buildings. The guideline consists of a main body, commentaries, and appendixes; it also provides standard procedures
Choi, B.; Nishida, Akemi; Shiomi, Tadahiko; Kawata, Manabu; Li, Y.
Proceedings of 28th International Conference on Nuclear Engineering (ICONE 28) (Internet), 7 Pages, 2021/08
In the seismic safety assessment of building structures in nuclear facilities, lumped mass models are conventionally used. However, they cannot possess the required high-accuracy evaluation of nuclear facilities, such as the local response at the equipment location in a reactor building. In this point of view, a seismic response analysis method using a three-dimensional finite element (3D FE) model is indispensable. Although, it has been reported that the analysis results obtained using 3D FE models vary greatly depending on the experience and knowledge of analysts, the quality of analysis results should be insured by developing a standard analysis method. In the Japan Atomic Energy Agency, we have developed a guideline for seismic response analysis methods that adopt 3D FE models of reactor buildings. The guideline consists of a main body, commentary, and several supplements; it also includes procedures, recommendations, points of attention, and a technical basis for conducting seismic response analysis using 3D FE models of reactor buildings. In this paper, the outline of the guideline and analysis examples based on the guideline are presented.
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.
Oba, Masaki; Miyabe, Masabumi; Akaoka, Katsuaki; Wakaida, Ikuo
Japanese Journal of Applied Physics, 59(6), p.062001_1 - 062001_6, 2020/06
Times Cited Count：8 Percentile：48.16(Physics, Applied)Using a semiconductor microwave source and a coaxial cable for microwave transmission, a compact microwave-assisted, laser-induced breakdown spectroscopy system without a microwave cavity or waveguide was developed. Several types of electrode heads were tested, so that the emission intensity was 50 times larger than without microwave. The limit of the enhancement effect was also found.
Katsuyama, Jinya; Osakabe, Kazuya*; Uno, Shumpei*; Li, Y.; Yoshimura, Shinobu*
Journal of Pressure Vessel Technology, 142(2), p.021205_1 - 021205_10, 2020/04
Times Cited Count：2 Percentile：15.42(Engineering, Mechanical)no abstracts in English
Hashimoto, Makoto
Nihon Genshiryoku Gakkai-Shi ATOMO
, 61(7), p.525 - 528, 2019/07
no abstracts in English
Kato, Shinya; Matsuba, Kenichi; Kamiyama, Kenji; Ganovichev, D. A.*; Baklanov, V. V.*
Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 9 Pages, 2019/05
In order to ensure In-Vessel Retention (IVR) of molten-core in Core Disruptive Accident (CDA), we are investigating the possibility of the molten-core discharge through the control rod guide tube (CRGT) to prevent energetics due to exceeding the prompt criticality. Internal structures of the CRGT, such as a sodium-flow regulator when the CRGT is connected to the high-pressure plenum, may disturb the discharge of molten-core from the core region. Based on above background, an experimental program to clarify characteristics of molten-core discharge through the CRGT has been commenced as one of subjects under a joint study with National Nuclear Center of the Republic of Kazakhstan (NNC-RK) named EAGLE-3 project. An experiment using molten-alumina as fuel simulant and sodium was conducted at the out-of-pile test facility owned by NNC-RK to investigate sodium cooling effect around the sodium flow regulator on its destruction. The experimental result represented that void development at the initiation of molten-alumina discharge eliminated liquid-phase sodium from the discharge path and this also eliminated sodium cooling effect around the sodium flow regulator. As a result, early destruction of the sodium flow regulator and massive discharge of molten alumina occurred in turn.
Miyahara, Kaname; Kawase, Keiichi
Genshiryoku No Ima To Ashita, p.159 - 167, 2019/03
This manuscript overviews lessons learned from decontamination pilot projects towards implementation of regional remediation after the environmental contamination due to the Fukushima Daiichi Nuclear Power Plant Accidents.
Tamura, Itaru
Hamon, 28(4), p.204 - 207, 2018/11
A Neutron guide is one of the devices to transport neutron beam for long distance without sacrificing much neutrons; therefore, it can supply neutrons to many experimental instruments distributed in a large experimental hall. Also, by using a curved guide, only the neutrons in a required energy range can be transported, and 
rays and fast neutrons can be effectively eliminated, therefore the signal to background ratio is improved. In addition, a neutron beam can be branched by applying curved guides. Neutron guides are also used to control the divergence angle and intensity of the neutron beam supplied to the neutron instrument.
Nakai, Ryodai
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 10 Pages, 2017/06
The GIF Safety Design Criteria Task Force (SDC TF) has been developing a set of safety design guidelines (SDG) to support practical application of SDC since the completion of the "SDC Phase I Report" that clarifies safety design requirements for Gen-IV SFR systems. The main objective of the SDG development is to assist SFR developers and vendors to utilize the SDC in their design process for improving the safety in specific topical areas including the use of inherent/passive safety features and the design measures for prevention and mitigation of severe accidents. The first report on "Safety Approach SDGs" aims to provide guidance on safety approaches covering specific safety issues on fast reactor core reactivity and on loss of heat removal. The second report on "SDGs on key Structures, Systems and Components (SSCs)" focuses on the functional requirements for SSCs important to safety; reactor core system, reactor coolant system, and containment system.
Sato, Masayuki*; Muraoka, Koji*; Hozumi, Koki*; Sanada, Yukihisa; Yamada, Tsutomu*; Torii, Tatsuo
Nihon Koku Uchu Gakkai Rombunshu (Internet), 65(2), p.54 - 63, 2017/02
This paper is concerned with the design problem of preview altitude controller for Unmanned Airplane for Radiation Monitoring System (UARMS) to improve its control performance. UARMS has been developed for radiation monitoring around Fukushima Daiichi Nuclear Power Plant which spread radiation contaminant due to the huge tsunamis caused by the Great East Japan Earthquake. The monitoring area contains flat as well as mountain areas. The basic flight controller has been confirmed to have satisfactory performance with respect to altitude holding; however, the control performance for variable altitude commands is not sufficient for practical use in mountain areas. We therefore design preview altitude controller with only proportional gains by considering the practicality and the strong requirement of safety for UARMS. Control performance of the designed preview controller was evaluated by flight tests conducted around Fukushima Sky Park.
Furusawa, Akinori; Nishimura, Akihiko; Takenaka, Yusuke*; Nakamura, Kaori*
Proceedings of 54th Annual Meeting of Hot Laboratories and Remote Handling (HOTLAB 2017) (Internet), 6 Pages, 2017/00
Testing of concrete structures in NPPs is needed to guarantee hereafter workability. Recent work says Core Concrete Reaction advances erosion of the concrete structures of Fukushima NPPs and it's difficult to estimate the correct depth of CCR. In addition, it is clear that seawater intrusion makes the rebar in the concrete structures corroded, thus, advanced remote testing methods for the deterioration should be considered. Gap or decrease of the adhesiveness between rebar and outer concrete appears in its deterioration process. We had a sense of possibility introducing a new testing method based on that. The concept is to propagate laser-excited ultrasonic gathering the information about the deterioration inside and received at distance with LDV. In this work, we investigate and report how it has the effect on propagating ultrasonic along the rebar to decrease adhesiveness between the rebar and the concrete experimentally.
Watanabe, Fumitaka; Okuno, Hiroshi
Proceedings of 18th International Symposium on the Packaging and Transport of Radioactive Materials (PATRAM 2016) (DVD-ROM), 9 Pages, 2016/09
This paper shows our calculations on the effects of a radiological release by assuming a severe accident in nuclear material transportation. Following recalculations of safe distance from the point of a severe accident during transportation of a transportation cask TN12 typically used in France containing spent nuclear fuel, and calculations to replicate the "Regulatory Guide: Emergency Preparedness for Nuclear Facilities", a similar calculation was made for a spent fuel transportation cask NFT-14P that was typically utilized in Japan instead of TN12. The safe distance was calculated to be about 30 m. The above calculations were made with the HotSpot codes which adopted the Gauss plume model and had been developed by the USA. Some additional calculations were made with EyesAct, which was developed and used in Japan, adopting also the Gauss plume model, to compare calculation results.
Yamashita, Shinichiro; Nagase, Fumihisa; Kurata, Masaki; Kaji, Yoshiyuki
Proceedings of Annual Topical Meeting on LWR Fuels with Enhanced Safety and Performance (TopFuel 2016) (USB Flash Drive), p.21 - 30, 2016/09
Fuel rod, channel box, and control rod designed with new materials and concepts have been developed in Japan for increasing accident tolerance of LWRs. In order to efficiently and properly implement the accident tolerant fuels (ATFs) and the other components, it is necessary not only to accumulate fundamental and practical data but also to consider technology readiness, recognize knowledge gaps, and establish strategy for design and fabrication. The Japan Atomic Energy Agency (JAEA) has established the above "technical basis" and drafted a research plan towards implementation of the ATFs and components as a program sponsored and organized by the Ministry of Economy, Trade and Industry (METI). It is useful to take advantage of the experiences in commercial uses of zirconium-base alloys in LWRs and, therefore, JAEA has conducted this METI project in cooperation with power plant providers, fuel venders, research institutes and universities who have been involved in the development of the ATF materials. The present paper describes the main results of the project conducted to establish the technical basis of the ATFs and components.
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.
