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Nuclear Science Research Institute
JAEA-Review 2018-036, 216 Pages, 2019/03
JAEA-Review-2018-036.pdf:19.22MB
Nuclear Science Research Institute (NSRI) is composed of Planning and Coordination Office, Fukushima Project Team and six departments, namely Department of Operational Safety Administration, Department of Radiation Protection, Engineering Services Department, Department of Research Reactor and Tandem Accelerator, Department of Fukushima Technology Development and Department of Decommissioning and Waste Management, and each departments manage facilities and develop related technologies to achieve the "Middle-term Plan" successfully and effectively. In order to contribute the future research and development and to promote management business, this annual report summarizes information on the activities of NSRI of JFY 2013 and 2014 as well as the activity on research and development carried out by Nuclear Safety Research Center, Advanced Research Center, Nuclear Science and Engineering Center and Quantum Beam Science Center, and activity of Nuclear Human Resource Development Center, using facilities of NSRI.
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
Fuketa, Toyoshi*; Nagase, Fumihisa
Zirconium in the Nuclear Industry; 18th International Symposium (ASTM STP 1597), p.52 - 92, 2018/01
Extensive research programs have been performed for more than two decades in JAEA and a better understanding has been developed for fuel behavior under accident conditions. The program is comprised of: RIA studies including pulse-irradiation experiments in the NSRR, cladding mechanical tests, and development and verification of a computer code RANNS; LOCA tests including integral thermal shock tests, oxidation rate measurements, and cladding mechanical tests; development and verification of a computer code FEMAXI-6, etc. Data and findings from the research programs provided technical basis directly and indirectly for regulatory criteria in Japan and other countries. This paper reviews and summarizes the major outcome from the research programs and identifies further research needs, as the acceptance technical paper for the Kroll Medal award of ASTM.
Amaya, Masaki; Udagawa, Yutaka; Narukawa, Takafumi; Mihara, Takeshi; Taniguchi, Yoshinori
Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 10 Pages, 2017/09
Amaya, Masaki; Udagawa, Yutaka; Narukawa, Takafumi; Mihara, Takeshi; Taniguchi, Yoshinori
Proceedings of Annual Topical Meeting on LWR Fuels with Enhanced Safety and Performance (TopFuel 2016) (USB Flash Drive), p.53 - 62, 2016/09
In order to evaluate adequacy of present safety criteria and safety margins in terms of advanced fuels and provide a database for future regulation on them, JAEA started an extensive research program called ALPS-II program, which has been sponsored by NRA, Japan. This program is primarily composed of tests simulating a RIA and a LOCA on the high-burnup advanced fuels irradiated in commercial PWR or BWR. Recently, the failure limits of the high-burnup advanced fuels under RIA conditions were investigated at NSRR, and post-test examinations on the fuel rods after the pulse irradiation tests are being performed. In terms of the simulated LOCA test, integral thermal shock tests and high temperature oxidation tests were carried out at RFEF, and the fracture limits, high temperature oxidation rate, etc. of the high-burnup advanced fuel cladding were investigated. This paper mainly describes some recent experimental results obtained in this program with respect to RIA and LOCA.
Udagawa, Yutaka; Sugiyama, Tomoyuki*; Amaya, Masaki
Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.1183 - 1189, 2016/04
Kudo, Tamotsu; Onizawa, Kunio*; Nakamura, Takehiko
JAEA-Evaluation 2015-011, 209 Pages, 2015/11
JAEA-Evaluation-2015-011.pdf:10.36MB
Japan Atomic Energy Agency (JAEA) consulted an assessment committee, "Evaluation Committee of Research and Development (R&D) Activities for Nuclear Safety", for post- and pre-review assessment of R&D on nuclear safety research. In response to JAEA's request, the Committee assessed mainly the progress of the R&D project according to guidelines, which addressed the rationale behind the R&D project, the relevance of the project outcome and the efficiency of the project implementation during the period of the current and next plan. As a result, the Committee concluded that the progress of the R&D project is satisfactory. This report describes the results of evaluation by the Committee. In addition, the appendix of this report contains presentations used for the evaluation, and responses from JAEA on the comments from the member of the Committee.
Amaya, Masaki; Udagawa, Yutaka; Narukawa, Takafumi; Mihara, Takeshi; Sugiyama, Tomoyuki
Proceedings of Annual Topical Meeting on Reactor Fuel Performance (TopFuel 2015), Part.2 (Internet), p.10 - 18, 2015/09
Advanced fuels which consist of cladding materials with high corrosion resistance and pellets with lower fission gas release have been developed by utilities and fuel vendors to improve fuel performance even in the high burnup region and also raise the safety level of current nuclear power plants to a higher one. In order to evaluate the adequacy of present safety criteria and safety margins in terms of such advanced fuels and provide a database for future regulation on them, Japan Atomic Energy Agency (JAEA) has started a new extensive research program called ALPS-II program (Phase II of Advanced LWR Fuel Performance and Safety program). This program is primarily composed of tests simulating a reactivity-initiated accident (RIA) and a loss-of-coolant accident (LOCA) on high burnup advanced fuels shipped from European nuclear power plants. This paper describes an outline of this program and some experimental results with respect to RIA and LOCA which have been obtained in this program.
Udagawa, Yutaka; Suzuki, Motoe; Amaya, Masaki
JAEA-Data/Code 2014-025, 27 Pages, 2015/02
JAEA-Data-Code-2014-025.pdf:2.53MB
A light water reactor fuel analysis code RANNS has been developed to analyze thermal and mechanical behaviors of a single fuel rod in mainly reactivity-initiated accident (RIA) conditions. The recent model development for the RANNS code has been focused on improving predictability of stress, strain, and temperature inside a fuel rod during pellet cladding mechanical interaction (PCMI), which is one of the most important behaviors of high-burnup fuels under RIA conditions. This report provides descriptions of the models developed and/or validated recently via experimental analyses using the RANNS code on the RIA-simulating experiments conducted in the nuclear safety research reactor (NSRR): models for mechanical behaviors as relocation of fuel pellets, pellet yielding, pellet-cladding mechanical bonding, and PCMI failure limit of fuel cladding, and thermal behaviors as pellet-cladding gap conductance and heat transfer from fuel rod surface to coolant water.
Udagawa, Yutaka; Sugiyama, Tomoyuki; Suzuki, Motoe; Amaya, Masaki
IAEA-TECDOC-CD-1775; Proceedings of Modelling of Water Cooled Fuel Including Design Basis and Severe Accidents (CD-ROM), p.200 - 219, 2015/00
Udagawa, Yutaka; Sugiyama, Tomoyuki; Amaya, Masaki
Proceedings of 2014 Water Reactor Fuel Performance Meeting/ Top Fuel / LWR Fuel Performance Meeting (WRFPM 2014) (USB Flash Drive), 8 Pages, 2014/10
Udagawa, Yutaka; Sugiyama, Tomoyuki; Amaya, Masaki
JAEA-Data/Code 2013-021, 43 Pages, 2014/02
JAEA-Data-Code-2013-021.pdf:2.74MB
JAEA-Data-Code-2013-021-appendix(CD-ROM).zip:126.49MB
In order to study the effects of cooling conditions on the boiling heat transfer from the fuel rod surface to the coolant water, RIA-simulating experiments with fresh fuels had been conducted in the nuclear safety research reactor (NSRR) under cooling conditions with subcoolings of 
10 to 80 K, flow velocities of 0 to 3 m/s, pressures of 0.1 to 16 MPa. In addition, pre-irradiated fuels had been subjected to the NSRR experiments under cooling conditions with subcoolings of 80 K, stagnant water, and atmospheric pressure. Out of the NSRR experiments, this report presents the fuel specifications, the test conditions, and the transient records during the pulse operations for the cases that the cladding temperature had been successfully measured. Characteristic parameters such as cladding peak temperatures were extracted from the transient records for summarizing the effects of cooling conditions and pre-irradiation on the heat transfer from the cladding surface.
Udagawa, Yutaka; Mihara, Takeshi; Sugiyama, Tomoyuki; Suzuki, Motoe; Amaya, Masaki
Journal of Nuclear Science and Technology, 51(2), p.208 - 219, 2014/02
Times Cited Count：5 Percentile：43.1(Nuclear Science & Technology)Udagawa, Yutaka; Sugiyama, Tomoyuki; Suzuki, Motoe; Nagase, Fumihisa
Journal of Nuclear Science and Technology, 50(6), p.645 - 653, 2013/06
Times Cited Count：2 Percentile：72.96(Nuclear Science & Technology)Fuketa, Toyoshi
Comprehensive Nuclear Materials, 2, p.579 - 593, 2012/02
Sugiyama, Tomoyuki; Udagawa, Yutaka; Suzuki, Motoe; Nagase, Fumihisa
Proceedings of 2011 Water Reactor Fuel Performance Meeting (WRFPM 2011) (CD-ROM), 6 Pages, 2011/09
The Japan Atomic Energy Agency has performed pulse irradiation tests using the NSRR to investigate fuel behavior under Reactivity-Initiated Accident (RIA) conditions. The NSRR tests have provided data of the pellet-cladding mechanical interaction (PCMI) failure of high burnup fuels up to 77 GWd/t. In particular, the PCMI failure limit is the important information which is needed in the reactor safety review. However, there are some differences between the NSRR tests and RIAs supposed in power reactors, such as the coolant temperature and the width of power pulse. Influence of these differences should be quantitatively evaluated in order to estimate the PCMI failure limit anticipated under the power reactor conditions from the NSRR data. This paper presents experimental results from a set of room and high temperature RIA tests, and discusses the evaluation procedure of the influence of coolant temperature and power pulse width on the failure limit on the basis of the experimental data.
O
doped PWR fuels under power transientYanagisawa, Kazuaki
JAEA-Review 2010-054, 109 Pages, 2010/11
JAEA-Review-2010-054.pdf:36.09MB
(1) The niobia doped PWR fuel (test specimen) did not fail below 260 cal/g defined as the failure threshold for RIA. (2) The fuel failure occurred by the cladding melt-brittle mechanism irrespective to the doping. (3) The test specimen caused a significant axial PCMI to the magnitude of 18%. (4)Above 276 cal/g, the test specimen restructured and formed the bonding, the small radial cracks, the metallic agglomeration and inclusion at the very narrow ring area. The grain size of the test specimen at the ring was reduced from original 31 to 10-21 microns. Meanwhile, undoped fuel grew the grain size from original 9 to 60 microns at the similar ring because of the recrystallization.
Tr
gours, N.*; Sugiyama, Tomoyuki
DPAM-SEMCA-2010-288, 73 Pages, 2010/11
The JAEA and IRSN are planning an experimental program to clarify the dynamics in fission gas release from fuel pellets under reactivity-initiated accident (RIA) conditions. In the Fission Gas Dynamics (FGD) test, high burnup fuel pellets in a rigid chamber is pulse-irradiated in the NSRR and the history of fission gas release is evaluated by transient measurements of the pressure and temperature in the chamber. In order to validate the chamber design and test conditions, preparatory calculations were performed on the high burnup UO
and MOX fuels, which had been tested in the NSRR, using the transient fuel behavior analysis code SCANAIR. The calculations provided informative results for the chamber design; the newly developed pressure sensor, which uses a linear variable differential transformer (LVDT), has an enough quick response, the thermal insulation around the chamber can be omitted because of the large heat capacity of the chamber wall, and so on.
Umeda, Miki; Sugiyama, Tomoyuki; Nagase, Fumihisa; Fuketa, Toyoshi; Ueta, Shohei; Sawa, Kazuhiro
Journal of Nuclear Science and Technology, 47(11), p.991 - 997, 2010/11
Fuketa, Toyoshi; Nagase, Fumihisa; Sugiyama, Tomoyuki; Amaya, Masaki
Proceedings of 2010 LWR Fuel Performance Meeting/TopFuel/WRFPM (CD-ROM), p.244 - 253, 2010/09
In order to evaluate adequacy of present safety criteria and safety margins and to provide a database for future regulation on higher burnup UO and MOX fuels, new cladding and pellets, an extensive program has been performed in the Japan Atomic Energy Agency (JAEA). The research program "Advanced LWR Fuel Performance and Safety" (ALPS) is comprised primarily of tests simulating a reactivity-initiated accident (RIA) and a loss-of-coolant accident (LOCA) on high burnup fuels shipped from European nuclear power plants. The paper describes key observations in the first phase (FY2002 to FY2007) and an outline of the second phase (FY2008 to FY2013) of the program.
