Narukawa, Takafumi; Kondo, Keietsu; Fujimura, Yuki; Kakiuchi, Kazuo; Udagawa, Yutaka; Nemoto, Yoshiyuki
Journal of Nuclear Materials, 582, p.154467_1 - 154467_12, 2023/08
Uwaba, Tomoyuki; Nemoto, Junichi*; Ito, Masahiro*; Ishitani, Ikuo*; Doda, Norihiro; Tanaka, Masaaki; Otsuka, Satoshi
Nuclear Technology, 207(8), p.1280 - 1289, 2021/08
Computer codes for irradiation behavior analysis of a fuel pin and a fuel pin bundle and for coolant thermal hydraulics analysis were coupled into an integrated code system. In the system, each code provides data required by other codes and the analyzed results are shared among them. The system allows for the synthesizing of analyses of thermal, chemical and mechanical behaviors in a fuel subassembly under irradiation. A test analysis was made for a fuel subassembly containing a mixed oxide fuel pin bundle irradiated in a fast reactor. The results of the analysis were presented with transverse cross-sectional images of the fuel subassembly and three-dimensional images of a fuel pin and fuel pin bundle models. For detailed evaluation, various irradiation behaviors of all fuel pins in the subassembly were analyzed and correlated with irradiation conditions.
Onishi, Takashi; Maeda, Koji; Katsuyama, Kozo
Journal of Nuclear Science and Technology, 58(4), p.383 - 398, 2021/04
Kitamura, Akira; Akahori, Kuniaki; Nagata, Masanobu*
Genshiryoku Bakkuendo Kenkyu (CD-ROM), 27(2), p.83 - 93, 2020/12
Direct disposal of spent nuclear fuel (SNF) in deep underground repositories (hereafter "direct disposal") is a concept that disposal canisters stored fuel assemblies dispose without reprocessing. Behavior of radionuclide release from SNF must be different from that from vitrified glass. The present study established a methodology on determination of instant release fraction (IRF) of radionuclides from SNF, which is the one of the parameters on radionuclide release based on the latest safety assessment reports in other countries, especially for IRF values proportional to a fission gas release ratio (FGR). Recommended and maximum values of FGR have been estimated using the fuel performance code FEMAXI-7 after collecting FGR values on Japanese SNFs. Furthermore, recommended and maximum values of IRF for Japanese SNFs used in a pressurized water reactor (PWR) have been estimated using the presently obtained FGR values and experimentally obtained IRF values on foreign SNFs. The recommended and maximum IRF values obtained in the present study have been compared with those of the latest safety assessment reports in other countries.
Collaborative Laboratories for Advanced Decommissioning Science; The University of Tokyo*
JAEA-Review 2019-037, 90 Pages, 2020/03
JAEA/CLADS, had been conducting the Center of World Intelligence Project for Nuclear Science/Technology and Human Resource Development (hereafter referred to "the Project") in FY2018. The Project aims to contribute to solving problems in nuclear energy field represented by the decommissioning of the Fukushima Daiichi Nuclear Power Station, Tokyo Electric Power Company Holdings, Inc. For this purpose, intelligence was collected from all over the world, and basic research and human resource development were promoted by closely integrating/collaborating knowledge and experiences in various fields beyond the barrier of conventional organizations and research fields. The sponsor of the Project was moved from the Ministry of Education, Culture, Sports, Science and Technology to JAEA since the newly adopted proposals in FY2018. On this occasion, JAEA constructed a new research system where JAEA-academia collaboration is reinforced and medium-to-long term research/development and human resource development contributing to the decommissioning are stably and consecutively implemented. Among the adopted proposals in FY2018, this report summarizes the research results of the "Development of Technology to Prevent Scattering of Radioactive Materials in Fuel Debris Retrieval". The objective of the present study is to clarify the behavior of microparticles in gas and liquid phases in order to steadily confine radioactive microparticles at the time of debris retrieval in Fukushima Daiichi Nuclear Power Station. In addition, as measures to prevent scattering, we will evaluate and develop methods by experiments and simulation as to; (1) a method to suppress the scattering with minimum amount of water utilizing water spray etc., and (2) a method to suppress the scattering by solidifying fuel debris.
Narukawa, Takafumi; Amaya, Masaki
Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.912 - 921, 2019/09
Nakayoshi, Akira; Suzuki, Seiya; Okamura, Nobuo; Watanabe, Masayuki; Koizumi, Kenji
Journal of Nuclear Science and Technology, 55(10), p.1119 - 1129, 2018/10
Amaya, Masaki; Udagawa, Yutaka; Narukawa, Takafumi; Mihara, Takeshi; Taniguchi, Yoshinori
Proceedings of Annual Topical Meeting on Reactor Fuel Performance (TopFuel 2018) (Internet), 10 Pages, 2018/10
Narukawa, Takafumi; Yamaguchi, Akira*; Jang, S.*; Amaya, Masaki
Nuclear Engineering and Design, 331, p.147 - 152, 2018/05
Narukawa, Takafumi; Yamaguchi, Akira*; Jang, S.*; Amaya, Masaki
Journal of Nuclear Materials, 499, p.528 - 538, 2018/02
Arai, Yasuo; Minato, Kazuo
Journal of Nuclear Materials, 344(1-3), p.180 - 185, 2005/09
no abstracts in English
Yamaji, Akifumi*; Oka, Yoshiaki*; Ishiwatari, Yuki*; Liu, J.*; Koshizuka, Seiichi*; Suzuki, Motoe
Proceedings of 2005 International Congress on Advances in Nuclear Power Plants (ICAPP '05) (CD-ROM), 7 Pages, 2005/05
Ensuring the fuel integrities is one of the most fundamental parts in the High Temperature Supercritical-Pressure Light Water Reactor. Most abnormal transient events of SCLWR-H last for a short period of time and the fuel rods are replaced after being irradiated in the core. In this study, the fuel integrity criteria are rationalized based on the fact that the fuel rod mechanical failures can be represented by the strain of the fuel rod cladding. A new fuel rod is designed with a Stainless Steel cladding. It is internally pressurized to reduce the stress on the cladding and also to increase the gap conductance between the pellet and the cladding. The fuel integrities both at normal operation and abnormal transient conditions are evaluated using the fuel analysis code FEMAXI-6 of JAERI.
Fuketa, Toyoshi; Nagase, Fumihisa; Sasahara, Akihiro*
Nihon Genshiryoku Gakkai-Shi, 47(2), p.112 - 119, 2005/02
Behavior of LWR fuel during reactivity-initiated accident (RIA) and loss-of-coolant accident (LOCA) is described.
Sasajima, Hideo; Sugiyama, Tomoyuki; Nakamura, Takehiko*; Fuketa, Toyoshi
JAERI-Research 2004-022, 113 Pages, 2004/12
Results from power burst tests, GK-1 and GK-2, conducted at the NSRR, are summarized. The tests were performed on a 1414 PWR fuel rod irradiated to a burnup of 42 MWd/kgU in the Genkai unit #1 of Kyushu Electric Power Co., Inc. The instrumented test fuel rod in a double-container-type capsule was subjected to the pulse-irradiation with stagnant water cooling condition at 0.1 MPa and 293 K. Deposited energy and peak fuel enthalpy were 505 J/g and 389 J/g in the Test GK-1, and 490 J/g and 377 J/g in the Test GK-2, respectively. During the pulse-irradiations, DNB occurred and the cladding surface temperature reached 581 K and 569 K in the Tests GK-1 and -2, respectively. The maximum cladding hoop strain was 2.7% in the Test GK-1 and 1.2% in the Test GK-2. However, the test fuel rods did not fail. Estimated fission gas releases during the pulse-irradiations were 11.7% and 7.0% in the Tests GK-1 and -2, respectively.
Fuel Safety Research Laboratory
JAERI-Review 2004-021, 226 Pages, 2004/10
Fuel Safety Research Meeting 2004, which was organized by Japan Atomic Energy Research Institute, was held on March 1-2, 2004 at Toranomon Pastoral, Tokyo. Purposes of the meeting are to present and discuss results of experiments and analyses on reactor fuel safety and to exchange views and experiences among the participants. Technical topics of the meeting covered status of fuel safety research activities, fuel behavior under RIA and LOCA conditions, high burnup fuel behavior, and radionuclides release under severe accident conditions. This proceeding contains all the papers presented in the meeting.
Harada, Katsuya; Nakata, Masahito; Harada, Akio; Nihei, Yasuo; Yasuda, Ryo; Nishino, Yasuharu
JAERI-Tech 2004-034, 13 Pages, 2004/03
The Department of Hot Laboratories has been aiming the establishment of the melting temperature measuring technique for small samples obtained from the micro-region of irradiated fuel pellet. Due to the modification of the shape of tungsten capsule contained sample and the improvement of the detection method for melting temperature from indistinct thermal arrest point owing to small sample, it is possible to determine the melting temperature of small sample and to utilize effectively for the irradiated fuel pellet by using the existing apparatus. This paper describes the technique of the melting temperature measurement for small sample and the experimental results by using tantalum, molybdenum, hafnium oxide and un-irradiated UO pellet.
Proceedings of 2nd Japan-Korea-China (5th Japan-Korea) Seminar on Nuclear Reactor Fuel and Materials, p.4 - 10, 2004/03
In designing a fuel performance code which describes complicated interactions working in high burnup fuel, the code will inevitably become a complex structure of inter-dependent models. In normal operation conditions, PCMI occurs and the pellet-clad firm bonding layer makes the cladding to be subjected to a bi-axial stress state, i.e. under tough mechanical loading. In contrast, the bonding layer enhances thermal conductance, decreases the pellet temperature and keeps the pellet-clad contact, resulting in increased resistance against the Lift-Off. For pellet behaviors, the fission gas bubble growth is strongly dependent on temperature, so that a reliable prediction of fuel temperature is required by pellet radial meshing which can fully accommodate the burning analysis results and the rim structure growth. The presentation deals with modeling method in terms of specific aspects such as meshing.
Nagase, Fumihisa; Fuketa, Toyoshi
NUREG/CP-0185, p.321 - 331, 2004/00
With a view to obtaining basic data to evaluate high burnup fuel behavior under loss of coolant accident (LOCA) conditions, a research program is being conducted at the Japan Atomic Energy Research Institute (JAERI). The program consists of integral thermal shock tests and other separate tests for oxidation rate and mechanical property of fuel claddings. Prior to the tests on irradiated claddings, the tests have been conducted on non-irradiated claddings to examine separate effects of corrosion and hydrogen absorption during reactor operation. Hydrogen effects have been especially examined because hydrogen absorption has the great impact on cladding embrittlement. The tests on irradiated claddings have recently been started and preliminary results have been obtained. The present paper summarizes recent results from those studies.
Nakamura, Jinichi; Nakamura, Takehiko; Sasajima, Hideo; Suzuki, Motoe; Uetsuka, Hiroshi
HPR-359, Vol.2, p.34_1 - 34_16, 2002/09
In BWR, power oscillations can occur due to the void fraction fluctuation. To investigate the fuel behavior during power oscillation of BWRs, two types of irradiated fuel rods were tested under simulated power oscillation conditions in the Nuclear Safety Research Reactor(NSRR). One is high burnup BWR fuel (56GWd/t) test, with 4 power oscillation cycles, to clarify the behavior of high burnup fuel. The second one is high enriched fuel(20%,25GWd/t) test, with 7 power cycles, to perform the test under high power conditions. The fuel behavior data, such as cladding elongation, fuel stack elongation, cladding temperature, etc. were obtained in these tests. The DNB did not occur in these tests. The PCI was observed through cladding elongation and fuel stack elongation during the power oscillations, but the residual strain of cladding was very small. Fuel behavior under simulated power oscillations is discussed based on in-pile data and PIE data and is compared with FEMAXI-6 and FRAP-T6 calculation.
Fuel Safety Research Laboratory
JAERI-Conf 2001-010, 303 Pages, 2001/09
The 24th NSRR Technical Review Meeting was held at Tranomon Pastoral, Tokyo, on November 13 and 14, 2000. The purpose of the meeting was to present and discuss the recent progress of the NSRR program and other LWR fuel safety researches at JAERI. Twenty-one papers, including five by foreign institutes, were presented and discussed regarding fuel behavior during normal operation, reactivity initiated accident (RIA) and loss-of-coolant accident (LOCA) and FP release behavior during severe accident. The meeting was a great help in planning future research and promoting research cooperation. This proceeding contains the papers presented in the meeting.