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JAEA Reports

Radiation monitoring using manned helicopter around the Nuclear Power Station in the fiscal year 2020 (Contract research)

Futemma, Akira; Sanada, Yukihisa; Ishizaki, Azusa; Kawasaki, Yoshiharu*; Iwai, Takeyuki*; Hiraga, Shogo*; Sato, Kazuhiko*; Haginoya, Masashi*; Matsunaga, Yuki*; Kikuchi, Hikaru*; et al.

JAEA-Technology 2021-029, 132 Pages, 2022/02

JAEA-Technology-2021-029.pdf:24.58MB

By the nuclear disaster of Fukushima Daiichi Nuclear Power Station (FDNPS), Tokyo Electric Power Company (TEPCO), caused by the Great East Japan Earthquake and the following tsunami on March 11, 2011, a large amount of radioactive material was released from the FDNPS. After the nuclear disaster, airborne radiation monitoring using manned helicopter has been conducted around FDNPS. The results of the airborne radiation monitoring and the evaluation for temporal change of dose rate in the fiscal 2020 were summarized in this report. Analysis considering topographical effects was applied to the result of the airborne monitoring to improve the accuracy of conventional method. In addition, technique for discriminating gamma rays from the ground and those from the airborne Rn-progenies was also utilized to evaluate their effect on airborne radiation monitoring.

JAEA Reports

Manufacture of substitutive assemblies for MONJU reactor decommissioning

Sakakibara, Hiroshi; Aoki, Nobuhiro; Muto, Masahiro; Otabe, Jun; Takahashi, Kenji*; Fujita, Naoyuki*; Hiyama, Kazuhiko*; Suzuki, Hirokazu*; Kamogawa, Toshiyuki*; Yokosuka, Toru*; et al.

JAEA-Technology 2020-020, 73 Pages, 2021/03

JAEA-Technology-2020-020.pdf:8.26MB

The decommissioning is currently in progress at the prototype fast breeder reactor Monju. Fuel assemblies will be taken out of its core for the first step of the great task. Fuel assemblies stand on their own spike plugged into a socket on the core support plate and support with adjacent assemblies through their housing pads each other, resulting in steady core structure. For this reason, some substitutive assemblies are necessary for the purpose of discharging the fuel assemblies of the core. Monju side commissioned, therefore, Plutonium Fuel Development Center to manufacture the substitutive assemblies and the Center accepted it. This report gives descriptions of design, manufacture, and shipment in regard to the substitutive assemblies.

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.

Journal Articles

A New critical assembly: STACY

Araki, Shohei; Gunji, Satoshi; Tonoike, Kotaro; Kobayashi, Fuyumi; Izawa, Kazuhiko; Ogawa, Kazuhiko

Proceedings of European Research Reactor Conference 2020 (RRFM 2020) (Internet), 7 Pages, 2020/10

Critical experiments of thermal neutron system are still expected to be playing an important role for wide technical issues. The Japan Atomic Energy Agency (JAEA) is renovating the Static Experimental Critical Facility (STACY) to maintain the experimental capability. The new STACY is designed as a general-purpose criticality facility. Its core mainly consists of low enriched UO$$_{2}$$ fuel rods, grid plates, and light water moderator. The first experiment campaign in the new STACY aims to obtain criticality characteristics of fuel debris, which will be used in validation of criticality analysis methods. The designs of the experimental core configurations are in progress.

Journal Articles

Progress of criticality control study on fuel debris by Japan Atomic Energy Agency to support Secretariat of Nuclear Regulation Authority

Tonoike, Kotaro; Watanabe, Tomoaki; Gunji, Satoshi; Yamane, Yuichi; Nagaya, Yasunobu; Umeda, Miki; Izawa, Kazuhiko; Ogawa, Kazuhiko

Proceedings of 11th International Conference on Nuclear Criticality Safety (ICNC 2019) (Internet), 9 Pages, 2019/09

Criticality control of the fuel debris in the Fukushima Daiichi Nuclear Power Station would be a risk-informed control to mitigate consequences of criticality events, instead of a deterministic control to prevent such events. The Nuclear Regulation Authority of Japan has administrated a research and development program to tackle this challenge since 2014. The Nuclear Safety Research Center of Japan Atomic Energy Agency, commissioned by the authority, is conducting activities such as computations of criticality characteristics of the fuel debris, development of a criticality analysis code, preparation of criticality experiments, and development of a criticality risk analysis method.

Journal Articles

Neutronic design of basic cores of the new STACY

Izawa, Kazuhiko; Ishii, Junichi; Okubo, Takuya; Ogawa, Kazuhiko; Tonoike, Kotaro

Proceedings of 11th International Conference on Nuclear Criticality Safety (ICNC 2019) (Internet), 9 Pages, 2019/09

Japan Atomic Energy Agency, JAEA, is conducting the renewal program of the heterogeneous water moderated critical assembly STACY (Static Experiment Critical Facility) in order to verify the criticality calculation considering fuel debris which have been produced in the accident of Fukushima Daiichi Nuclear Power Station. The first criticality of the new STACY is scheduled at the beginning of 2021. After the first criticality, it is necessary to perform a series of critical experiments with a series of basic experimental core in order to gain a proficiency of operators and grasp the uncertainty that accompanies the result of critical experiments in STACY. Prior to the construction of the new STACY, a series of neutronic calculation was carried out for licensing and planning first series of critical experiment. In this paper, possible core configuration of the basic experimental core and their limitations are discussed and presented.

Journal Articles

Study on reduction of potential radiotoxicity for spent fuel by using HTGR

Fukaya, Yuji; Kunitomi, Kazuhiko; Ogawa, Masuro

Nihon Genshiryoku Gakkai Wabun Rombunshi, 14(3), p.189 - 201, 2015/09

A study on reduction of potential radiotoxicity for spent fuel by using High Temperature Gas-cooled Reactors (HTGRs) have been performed. Unlike Partitioning and Transmutation (P&T), the reactor concept is investigated from the viewpoint of reduction of the radiotoxicity generation itself. To reduce the radiotoxicity, $$^{238}$$U, which generates Pu, Am and Cm, should be excluded. Therefore, we proposed HTGR fueled by new concept fuels with alternative fuel matrix instead of $$^{238}$$U. Those are Yttria Stabilized Zirconia (YSZ) and thorium, and the fissile material is Highly Enriched Uranium (HEU) with the enrichment of 93%. With the HEU, the radiotoxicity can be significantly reduced, and the cooling time to decay to a natural uranium level can be shorted down to approximately 800 years. Fuel integrity and proliferation resistance can be remained by the dilution using YSZ, and neutronic characteristics of self-regulation is remained by the loading of erbium. The fuel can generate heat as same amount as ordinary uranium fuel. The electricity generation cost is as cheap as ordinary GTHTR300. It is concluded that the proposed reactor concept can reduce the cooling time less than 1% from 100 thousand years to a 800 years without additional technology development.

Journal Articles

Thermal analysis of heated cylinder simulating nuclear reactor during loss of coolant accident

Sato, Hiroyuki; Ohashi, Hirofumi; Tachibana, Yukio; Kunitomi, Kazuhiko; Ogawa, Masuro

Journal of Nuclear Science and Technology, 51(11-12), p.1317 - 1323, 2014/11

 Times Cited Count:7 Percentile:55.77(Nuclear Science & Technology)

Transient analyses are presented of temperature behavior of reactor during loss-of-coolant accident with scram. The influence of reactor thermal properties, operating power density, geometry of active core and selection of fuel type on the capability of decay heat removal against the accident are studied. It is shown that the reactor design envelope is fully determined by the key parameters. The range of the envelope is shown to enlarge considerably by selecting high refractory fuel. High temperature gas-cooled reactor (HTGR), a graphite-moderated reactor with TRISO coated fuel particle, is the primary candidate which can fulfill the requirement to the design concept of nuclear reactor independent of coolant for decay heat removal.

JAEA Reports

Economic evaluation of HTGR IS process hydrogen production system

Iwatsuki, Jin; Kasahara, Seiji; Kubo, Shinji; Inagaki, Yoshiyuki; Kunitomi, Kazuhiko; Ogawa, Masuro

JAEA-Review 2014-037, 14 Pages, 2014/09

JAEA-Review-2014-037.pdf:8.84MB

Thermochemical iodine-sulfur (IS) process is one of the promising technologies, which harnesses heat energy of high temperature gas-cooled reactors (HTGRs). An economic estimation of hydrogen production by a future commercial HTGR-IS process hydrogen production system was performed on the basis of economic evaluation data of an existing commercial hydrogen production plant using fossil fuel as a raw material. Hydrogen production cost was estimated at 25.4 JPY/Nm$$^{3}$$ under this estimation conditions. Capital cost and energy cost account for 13% and 78% of the total hydrogen production cost, respectively. To decrease HTGR construction cost, to increase HTGR availability, to improve hydrogen production thermal efficiency are important for cost reduction of hydrogen. The cost will be competitive with estimated costs by fossil fuel hydrogen production methods. It is appropriate that the hydrogen production cost is set for a goal of present R&Ds.

Journal Articles

Feasibility study on Naturally Safe HTGR (NSHTR) for air ingress accident

Ohashi, Hirofumi; Sato, Hiroyuki; Tachibana, Yukio; Kunitomi, Kazuhiko; Ogawa, Masuro

Nuclear Engineering and Design, 271, p.537 - 544, 2014/05

 Times Cited Count:3 Percentile:28.59(Nuclear Science & Technology)

The concept of the Naturally-safe HTGR is that the release of radioactive materials is kept at very low level and no harmful effect on people and the environment is ensured by only physical phenomena without any engineered safety features. In this study, the CO concentration and the heat generated by graphite oxidation inside the circular tube were evaluated parametrically using a steady-state one-dimensional model to confirm the feasibility of the Naturally-safe HTGR at a severe condition of the air ingress accident (i.e., a massive air ingress by simultaneous rupture of two primary pipes). It was confirmed that the CO concentration at the outlet of coolant channel can be maintained below the explosion limit due to the reaction with oxygen in the air, and the reaction heat can be removed with the decay heat by physical phenomena under certain conditions of the coolant channel geometry without any engineered safety features.

Journal Articles

Analysis of core heat removal capability under DLOFC accidents for HTGRs

Sato, Hiroyuki; Ohashi, Hirofumi; Tachibana, Yukio; Kunitomi, Kazuhiko; Ogawa, Masuro

Nuclear Engineering and Design, 271, p.530 - 536, 2014/05

 Times Cited Count:2 Percentile:19.81(Nuclear Science & Technology)

Design envelope of prismatic High Temperature Gas-cooled Reactors in terms of core heat removal capability under depressurized loss-of-forced-circulation accidents without operating active or passive decay heat removal systems are investigated. Lumped element models consist of core, reactor pressure vessel and cavity wall are presented in order to evaluate transient response of core temperature. Parametric calculations changing the core height, initial core temperature, thickness of side reflector, cavity size and peaking factor are performed. A series of calculation provides relationships of core radius to average power density and reactor thermal power which can remove the heat in core without reliance on specific design features. The results clarified the design envelope for the Naturally Safe HTGR in terms of core decay heat removal.

Journal Articles

Inherently-safe high temperature gas-cooled reactor for hydrogen production

Ogawa, Masuro; Kunitomi, Kazuhiko; Sato, Hiroyuki

Keynote Lecture for International Conference of PM2.5 & Energy Security 2014 (PMES 2014), p.72 - 74, 2014/03

The present paper reviewed the overview of HTGR technologies as well as excellent characteristics, such as environmental friendliness, safety, minimum waste, economic competitiveness, sustainability, usability and proliferation resistance, suitable for an energy supply system to be deployed in the field of heat utilization.

Journal Articles

Concept on inherent safety in high-temperature gas-cooled reactor

Ohashi, Hirofumi; Sato, Hiroyuki; Kunitomi, Kazuhiko; Ogawa, Masuro

Nihon Genshiryoku Gakkai Wabun Rombunshi, 13(1), p.17 - 26, 2014/03

A new concept on inherent safety in High Temperature Gas-cooled Reactor (HTGR) was proposed to accomplish no harmful consequences for people and the environment even in the failures in safety and safety related systems. The safety concept is that the progression of the events that lead the loss or degradation of the confinement function of the coated fuel particle is suppressed and the release of radioactive materials is kept at very low level by only physical phenomena. The feasibility studies and related information revealed that the reactor design on this safety concept is technically feasible. The physical phenomena can be appeared with the cause of event (i.e., temperature increase, oxidation of fuel cladding and explosion of carbon monoxide) and can prevent or mitigate the events.

Journal Articles

Analysis of core heat removal capability under DLOFC accidents for HTGRs

Sato, Hiroyuki; Ohashi, Hirofumi; Tachibana, Yukio; Kunitomi, Kazuhiko; Ogawa, Masuro

Proceedings of 6th International Topical Meeting on High Temperature Reactor Technology (HTR 2012) (USB Flash Drive), 9 Pages, 2012/10

Design criteria of prismatic High Temperature Gas-cooled Reactors (HTGRs) in terms of core heat removal capability under depressurized loss-of-forced-circulation accidents without operating active or passive decay heat removal systems are investigated. Lumped element models consist of core, RPV and surroundings and soil are presented in order to evaluate transient response of core and RPV temperatures. The results clarified the design criteria for the Inherently-safe HTGR in terms of core heat removal under DLOFC accidents.

Journal Articles

Feasibility study on naturally-safe HTGR for air ingress accident

Ohashi, Hirofumi; Sato, Hiroyuki; Tachibana, Yukio; Kunitomi, Kazuhiko; Ogawa, Masuro

Proceedings of 6th International Topical Meeting on High Temperature Reactor Technology (HTR 2012) (USB Flash Drive), 10 Pages, 2012/10

The concept of the Naturally-safe HTGR is that the release of radioactive materials is kept at very low level and no harmful effect on people and the environment is ensured by only physical phenomena without any engineered safety features. At an air ingress accident, possible physical events to loss of the confinement function of the fuel coating layers are the crack of the coatings caused by the explosion of CO produced by the graphite oxidation and failure of the coatings by melting or sublimation caused by core heat up due to the reaction heat. The CO concentration and the heat generated by graphite oxidation inside the circular tube were numerically evaluated. It was confirmed that the CO concentration at the outlet of coolant channel can be maintained below the explosion limit due to the reaction with oxygen in the air, and the reaction heat can be removed by physical phenomena under certain conditions of the coolant channel geometry without any engineered safety features.

Journal Articles

Concept of an inherently-safe high temperature gas-cooled reactor

Ohashi, Hirofumi; Sato, Hiroyuki; Tachibana, Yukio; Kunitomi, Kazuhiko; Ogawa, Masuro

AIP Conference Proceedings 1448, p.50 - 58, 2012/06

 Times Cited Count:7 Percentile:91.9

As the challenge to ensure no harmful release of radioactive materials at the accidents by deterministic approach instead to satisfy the safety goal of the risk by probabilistic approach, new concept of advanced reactor, an inherently-safe high temperature gas-cooled reactor, is proposed based on the experience of the operation of the actual HTGR in Japan, HTTR, and the design of the commercial plant (GTHTR300), utilizing the inherent safety features of the HTGR (i.e., safety features based on physical phenomena). The safety design philosophy of the inherently-safe HTGR for the safety analysis of the radiological consequences is determined as the confinement of radioactive materials is assured by only inherent safety features without engineered safety features, AC power or prompt actions by plant personnel if the design extension conditions occur. The concept and future R&D items for the inherently-safe HTGR are described in this paper.

Journal Articles

Evaluation of nuclear characteristics of light-water-moderated heterogeneous cores in modified STACY

Izawa, Kazuhiko; Aoyama, Yasuo; Sono, Hiroki; Ogawa, Kazuhiko; Yanagisawa, Hiroshi; Miyoshi, Yoshinori

Proceedings of 9th International Conference on Nuclear Criticality (ICNC 2011) (CD-ROM), 11 Pages, 2012/02

For reactor physics and criticality safety researches, the Static Experiment Critical Facility (STACY) will be modified. In the modification, the present STACY, solution-fuel-type homogeneous cores, will be converted to fuel-pin-type heterogeneous cores moderated by light water. For nuclear safety design of the modified STACY, computational analyses have been carried out by using a Monte Carlo code MVP and a transport code system DANTSYS with cross-section data based on the JENDL-3.3. In the analyses, basic nuclear characteristics have been evaluated, such as criticality, water-level worth and reactor shutdown margin. By the results of these analyses, the feasibility of reactivity control mechanism and the sufficiency of reactor shutdown margin of the modified STACY were confirmed. In addition, temperature and void coefficients of reactivity and kinetic parameters were obtained to comprehend nuclear characteristics of the modified STACY.

Journal Articles

High-temperature continuous operation of the HTTR

Takamatsu, Kuniyoshi; Sawa, Kazuhiro; Kunitomi, Kazuhiko; Hino, Ryutaro; Ogawa, Masuro; Komori, Yoshihiro; Nakazawa, Toshio*; Iyoku, Tatsuo; Fujimoto, Nozomu; Nishihara, Tetsuo; et al.

Nihon Genshiryoku Gakkai Wabun Rombunshi, 10(4), p.290 - 300, 2011/12

A high temperature (950$$^{circ}$$C) continuous operation has been performed for 50 days on the HTTR from January to March in 2010, and the potential to supply stable heat of high temperature for hydrogen production for a long time was demonstrated for the first time in the world. This successful operation could establish technological basis of HTGRs and show potential of nuclear energy as heat source for innovative thermo-chemical-based hydrogen production, emitting greenhouse gases on a "low-carbon path" for the first time in the world.

JAEA Reports

Experiments of sodium nitrate liquid waste treatment by biological method

Takahashi, Kuniaki; Meguro, Yoshihiro; Kawato, Yoshimi; Kuroda, Kazuhiko*; Ogawa, Naoki*

JAEA-Technology 2008-084, 12 Pages, 2009/02

JAEA-Technology-2008-084.pdf:1.06MB

Low level liquid waste discharged from a Reprocessing Facility includes sodium nitrate. In the case that it is directly solidified with cement and so on and then the solidified waste are disposed under the ground, sodium nitrate soaks into the environment through underground water layer. We planned to apply the biological treatment system that many ordinary industrial plants are running in the field of waste water treatment to reduce nitrate. We carried out degradation experiments of nitrate for 4wt% sodium nitrate solution by biological method. To solve the assignments that biological treatment technology has, we tested and obtained the results as shown below; (1) The amount of sludge ash could be cut down a tenth as much as usual. The disposal cost reduction of secondary waste is just in sight. (2) Treatment performance could be improved up to 7 kg-N/m$$^{3}$$/d from 4 kg-N/m$$^{3}$$/d. It could be expected the more compact system by improvement of the membrane set into the biological treatment tanks.

Journal Articles

New estimation method for void reactivity coefficient using the TRACY transient data

Ogawa, Kazuhiko; Kaminaga, Fumito*

Journal of Nuclear Science and Technology, 46(1), p.1 - 5, 2009/01

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

In the safety research regarding a criticality accident in nuclear reprocessing plants, prediction of nuclear fission yields is essential to evaluate its influence on a radiation of neutrons/$$gamma$$-rays to the public. For numerical calculations of the number of fissions in a criticality accident, a void reactivity coefficient due to the formation of radiolytic gas voids in the solution is required to evaluate the negative feedback reactivity. Small oscillations of both the power and the core pressure are measured after the first power pulse in a slow transient operation in TRACY. The small power oscillations are due to the reactivity feedback effect of the change of the volume of gas voids. In this paper, we propose a new estimation method of the void reactivity coefficient and estimate the coefficient using the TRACY transient data. The present estimated void reactivity coefficient is less than the calculated values in the uniform distribution of gas voids.

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