Japan Nuclear Fuel Ltd. is a nuclear energy company based in Rokkasho, Aomori Prefecture, Japan involved in the production of nuclear fuel, as well as the reprocessing, storage and disposal of nuclear waste. The mission of Japan Nuclear Fuel Limited is to establish a nuclear fuel cycle infrastructure in Japan.In uranium enrichment JNFL plans to provide for an ultimate capacity of 1,500 ton-SWUyear, which is equivalent to the nuclear fuel used by 8 or 9 reactors at 1,000 MW-class nuclear plants.Rokkasho Reprocessing Plant, Japan's first commercial reprocessing plant, began reprocessing in 2007, however complications have delayed full commercial operation until 2012. The plant has a design reprocessing capacity of 800 tonnes-Uyear J-MOX fuel fabrication plant, which is located on the same site as the Rokkasho Reprocessing Plant.JNFL also operates low and high level nuclear waste long-term storage facilities which will accommodate 2,880 canisters of vitrified high level waste and the ultimate capacity of the Low-Level Radioactive Waste Disposal Center now under construction will be 600,000 m3. Wikipedia.
Ebina T.,Japan Nuclear Fuel Limited |
Yamazaki S.,Japan Nuclear Fuel Limited |
Yamazaki S.,Kobelco Research Institute |
Watanabe Y.,Tohoku University
Zairyo to Kankyo/ Corrosion Engineering | Year: 2016
Corrosion kinetics of a stainless steel in nitric acid at elevated temperature has been investigated with special emphasis placed on its correlation with reduction rate of Cr (VI) in the solution. Extra High Purity (EHP) stainless steel, which is immune to intergranular corrosion in trans-passive potential range in nitric acid solution, has been employed to examine the dominant factor of corrosion rate. The results indicate that dissolution amount of the alloy agrees well with the stoichiometric value of dissolution based on the reduction amount of Cr (VI). This fact means that the dissolution kinetics of the stainless steel is simply dominated by reduction rate of Cr (VI) in the boiling nitric acid including Cr(VI). Reduction rate constant of Cr (VI) follows the identical Arrhenius equation including all the data with and without boiling, although oxidation rate constant of Cr (III) to Cr (VI) is strongly affected by boiling. A dominant factor of corrosion rate has been cleared based on the reduction rate constant of Cr (VI) and the oxidation rate constant of Cr (III). © 2016, Japan Society of Corrosion Engineering. All rights reserved.
Shimizu J.,Japan Nuclear Fuel Limited |
Ohtsubo S.,MHI Nuclear Systems and Solution Engineering Co.
Transactions of the Atomic Energy Society of Japan | Year: 2017
Evaluation of the atmospheric-pressure change (APC) in a tornado is necessary to assess the integrity of nuclear-related facilities. The Rankine model has been most frequently used to theoretically calculate the APC in a tornado. The result, however, is considered to be overly conservative because the Rankine model wind speed at the ground is larger than that in reality. On the other hand, the wind speed of the Fujita model is closer to that of actual tornadoes but is expressed by more complicated algebraic equations than that in the Rankine model. Also, because it is impossible to analytically derive the APC equation using the Fujita model, numerical computation is required. A previous study employed the finite element method (FEM) for such a purpose. However, a general-purpose FEM code often requires complicated input parameters. In order to conduct parametric studies to evaluate the integrity of facilities in various cases of tornadoes, the finite-difference method code "TORPEC", which is specialized to analyze the APC, was developed as a convenient design tool. TORPEC is based on Poisson's equation derived from the Navier-Stokes equation. It also runs on widely available technical calculation software such as Microsoft® Excel VBA or MATLAB®. Taking advantage of such convenience, various calculations have been conducted to reveal the characteristics of APC as functions of the maximum tangential wind speed, axial position and tornado radius. TORPEC is used as a benchmark in the existing paper. The case study results obtained by TORPEC show a constant ratio of the pressure drop of the Fujita model against the Rankine model. This factor can be used to derive the Fujita model result from the Rankine model result without FEM analysis. © 2017 Atomic Energy Society of Japan, All Rights Reserved.
Kitashiba N.,Japan Nuclear Fuel Limited |
Azuma S.,Mitsubishi Group
LWR Fuel Performance Meeting, Top Fuel 2013 | Year: 2013
Mitsubishi Nuclear Fuel Co., Ltd. (MNF) has recently developed the ZDP-1 fuel assembly. ZDP stands for "Zero Defect Performance" and ZDP-1 is the first design in the ZDP fuel series. The ZDP-1 fuel assembly is designed to make more margins against grid-to-rod fretting (GTRF) wear at the bottom grid. ZDP-1 has three major improvements so as to reduce the flow exciting force on the bottom part of the fuel rod. One is the modified bottom nozzle which has small flow holes and unified flow-hole pattern. Second is shortened protrusion of the fuel rods below bottom grid. Third is the large tapered bottom end plug. The flow exciting forces of the bottom part of the fuel rod below the bottom grid are measured in the flow test facility at Takasago laboratory of Mitsubishi Heavy Industries, Ltd. The test results show each improvement has the significant effect to reduce the exciting force. In this paper, these test results and evaluations of the effects on the bottom grid GTRF are described. This technology will be also applied to the ZDP-2 fuel, which will be the next 17x17 fuel assembly from MNF.
Yasuda Y.,Japan Nuclear Fuel Limited |
Takahashi M.,Tohoku University
International Conference on Nuclear Engineering, Proceedings, ICONE | Year: 2015
In the present study, the concept of Generalized Failure Mechanism Knowledge (GFMK) has been proposed and applied practically to Rokkasho reprocessing plant. GFMK is a knowledge scheme that describes failure mechanism independent of particular subject. In this scheme, sets of conditions leading to specific failure are generalized, which allows us to apply to different equipment. Once GFMK have been constructed, it is possible to estimate the likelihood of the failure without expert judgments. Following two issues have been studied in this study. 1) Building a GFMK knowledgebase based on previous troubles cases 2) Application of GFMK to specific plant sub-systems to derive failure mechanism For issue 1), the GFMK knowledgebase has been constructed based on the previous failures cases in Rokkasho reprocessing plant with emphasis on the events related to aging. For issue 2), it has been confirmed that the possible failure mechanism can be obtained by applying GFMK to the realistic scale sub-system of Rokkasho reprocessing plant. The derived failure mechanism has been compared with the ones on the design specification and has been confirmed that most of the derived failure mechanism are feasible. Copyright © 2015 by JSME.
Tanaka K.-I.,Japan Atomic Power Company |
Ichige H.,Japan Atomic Power Company |
Tanabe H.,Japan Nuclear Fuel Limited
Proceedings of the International Conference on Radioactive Waste Management and Environmental Remediation, ICEM | Year: 2010
Preparatory tasks for decommissioning of nuclear power plant start with radiological characterization. Residual radioactivity inventory evaluation is a main part of the characterization. Reliable information on the inventory is important for specification for decommissioning plan. Japan Atomic Power Company (JAPC) has already started these tasks for Tsuruga Nuclear Power Plant Unit 1 (TS-1). We can optimize decommissioning plan using the information. To obtain the reliable information, we improved an evaluation procedure. The procedure is divided into two main steps. First step is neutron flux distribution calculation and second one is radioactivity distribution calculation. Radioactivity distribution is calculated using neutron flux distribution. In this work, we improved the evaluation procedure to obtain the reliable information on the inventory Because of the limitation of computer resource, two-dimension (2D) approximation model was applied to radioactivity distribution around Reactor Pressure Vessel (RPV). We can calculate reliable 2D neutron flux distribution by having better understanding of neutron transport phenomena. Neutron flux was measured at 30 locations in TS-1 Primary Containment Vessel (PCV) using activation foils. And in order to understand the neutron transport phenomenon inside the PCV, we also calculated neutron flux distribution with the three-dimensional (3D) discrete ordinates method calculation (Sn) code. By consideration about the result of the measurement and 3D calculation, we could understand the characteristics of the neutron flux distribution inside the PCV. To simulate the neutron flux distribution well with 2D Sn code, neutron flux behaviors inside the PCV had been investigated with referencing the measurement values and with observing calculated 3D neutron flux distribution. 2D calculation model had been modified repeatedly until reliable calculation result was provided. After several model modifications, the reliable 2D calculation was accomplished and important neutron transport phenomena that are necessary to simulate the neutron flux distribution well was understood. Network-parallel-computing technique was applied to radioactivity distribution calculation. Using this technique, we could calculate radioactivity at all space mesh points that were used with 2D Sn code and we obtained the radioactivity distribution. By using this distribution, we can estimate a quantity of radioactivity around RPV more accurately and optimize dismantling designs. © 2010 by ASME.
Nakajima K.,Kyoto University |
Itahara K.,Japan Nuclear Fuel Limited |
Okuno H.,Japan Atomic Energy Agency
ICNC 2015 - International Conference on Nuclear Criticality Safety | Year: 2015
In this paper an outline of the standard published in April 2015 by the Atomic Energy Society of Japan (AESJ) is presented. The standard gives basis of nuclear safety management of a spent fuel reprocessing plant that assumed burnup credit in nuclear criticality safety evaluation. Ten years ago the AESJ published Basic Items of Criticality Safety Control: 2004, AESJ-SC-F004:2004, which prescribed basic ideas, requirements and methods on nuclear criticality safety controls of facilities handling with nuclear fuel materials in general for preventing a nuclear criticality accident. However, it did not include any specific procedures for adopting burnup credit. Therefore, a new standard was envisaged as the first Standard for fuel reprocessing plants, which clarified the specific procedures to apply burnup credit to designers, operators, maintenance persons and administrators. The Standard, Procedures for Applying Burnup Credit to Criticality Safety Control of a Reprocessing Facility: 2014, AESJ-SC-F025:2014, was developed by the AESJ after vital discussions of the Working Group on Nuclear Criticality Safety Control, which were followed by critical reviews of the Special Committee on Nuclear Fuel Cycle and those of the Standard Committee, where both committees were established under the AESJ. The followings are some features of the standard: 1. In many countries, application of burnup credit has been considered especially in transportation and storage of spent fuel. The standard applies to a reprocessing plant, especially to the Rokkasho Reprocessing Plant in Japan. The present edition of the Standard assumes the first phase of burnup credit, i.e., burnup credit considering reduction in 235U, production of 239Pu and changes in other actinides, but not considering neutron absorption effect of fission products. 2. This standard covers nuclear criticality safety management including safety design and safety control stages. It prescribes each procedure of nuclear safety management with emphasis on its necessity rather in detail. Attached to the standard are seven appendices, which give reference to it, and explanations, which provide supplemental information to the readers. Future needs of AESJ standards in the field of nuclear criticality safety will be discussed in the presentation.
Sugawara T.,Akita University |
Shiono T.,University of Shiga Prefecture |
Yoshida S.,University of Shiga Prefecture |
Matsuoka J.,University of Shiga Prefecture |
And 2 more authors.
Physics and Chemistry of Glasses: European Journal of Glass Science and Technology Part B | Year: 2013
The densities of two simulated radioactive waste glasses and melts were determined between 298 and 1278 K by the Archimedean immersion method for annealed glasses at room temperature, dilatometric measurements for glasses in the temperature range from 373 K to the glass transition region and Archimedean densitometry using a molten chloride salt for liquids above 825 K. At temperatures above 1079 K the density of a stable liquid can be measured with a precision of 0·3%. The molar volume of the waste glass melt between 1079 and 1278 K is consistent with both the high temperature extrapolation of the molar volume at the glass transition temperature and the value estimated using partial molar volumes reported previously, which has been obtained using double-bob Archimedean densitometry. The temperature dependence of the molar volume (dV/dT) between the glass transition temperature and 1278 K is significantly larger than that calculated using partial molar thermal expansivities in the previous model. This suggests that the thermal expansivity of borosilicate melts is temperature dependent.
Swinhoe M.T.,Los Alamos National Laboratory |
Iwamoto T.,Japan Nuclear Fuel Limited |
Tamura T.,Japan Nuclear Fuel Limited
Nuclear Instruments and Methods in Physics Research, Section A: Accelerators, Spectrometers, Detectors and Associated Equipment | Year: 2010
Correlations are used to determine the 242Pu content of material using high resolution gamma measurements on the other plutonium isotopes because 242Pu itself has no practically detectable gamma emission lines. This paper presents an improved correlation that is particularly useful because, unlike some previous correlations, no prior knowledge of the reactor type or initial enrichment is required. This correlation has been shown to perform well over a range of plutonium from commercial BWR and PWR reactors. The agreement of the calculated 242Pu values with IDMS values is within 1% for 239Pu content of less than 70% and within 4% for 239Pu content of less than 80%. This simple form of the correlation is somewhat surprising given the complex behavior of 239/240 and 242/240 ratios. © 2010 Elsevier B.V.
Kitashiba N.,Japan Nuclear Fuel Limited
International Conference on Nuclear Engineering, Proceedings, ICONE | Year: 2015
In recent years, high burn-up fuel (∼55GWd/t (Assembly)) experienced some fuel leakages in Japanese PWRs. Grid-to-Rod fretting (GTRF) wear at the bottom grid position in some fuel assemblies was observed through fiberscope in onsite inspection and GTRF was considered to be cause of leakage issues. As a countermeasure to these fuel leakages, Mitsubishi Nuclear Fuel (MNF) has developed ZDP-1 fuel assembly. ZDP stands for "Zero Defect Performance". ZDP-1 is MNF's first design of ZDP fuel series. ZDP-1 fuel assembly is designed to make more margins against GTRF wear at the bottom grid. In this paper, the improvements of ZDP-1 and the wear evaluation results of ZDP-1 are explained, together with the situation of the wear issue and results of root cause analysis on the fuel leakages. Copyright © 2015 by JSME.
Kamaya M.,Japan Institute of Nuclear Safety System |
Kitsunai Y.,Japan Nuclear Fuel Limited |
Koshiishi M.,Japan Nuclear Fuel Limited
Journal of Nuclear Materials | Year: 2015
True stress-strain curves were obtained for irradiated 316L stainless steel by a tensile test and by a curve estimation procedure. In the tensile test, the digital image correlation technique together with iterative finite element analysis was applied in order to identify curves for strain larger than the necking strain. The true stress-strain curves were successfully obtained for the strain of more than 0.4 whereas the necking strain was about 0.2 in the minimum case. The obtained true stress-strain curves were approximated well with the Swift-type equation including the post-necking strain even if the exponential constant n was fixed to 0.5. Then, the true stress-strain curves were estimated by a curve estimation procedure, which was referred to as the K-fit method. Material properties required for the K-fit method were the yield and ultimate strengths or only the yield strength. Some modifications were made for the K-fit method in order to improve estimation accuracy for irradiated stainless steels. © 2015 Elsevier B.V. All rights reserved.