JAEA-Technology 2021-023, 190 Pages, 2021/11
Computational analyses on nuclear criticality characteristics were carried out for heterogeneous lattice systems composed of water moderator and fuel rods utilized in low-power research and test reactors, in which the depletion of fuel due to burnup is relatively small, by using the continuous-energy Monte Carlo code MVP Version 2 with the evaluated nuclear data library JENDL-4.0. In the analyses, the minimum critical number of fuel rods was evaluated using calculated neutron multiplication factors for the heterogeneous systems of the uranium dioxide fuel rod in the Static Experiments Critical Facility (STACY) and the Tank-type Critical Assembly (TCA), and the uranium-zirconium hydride fuel rod in the Nuclear Safety Research Reactor (NSRR). In addition, six sorts of the ratio of reaction rates, which are components of neutron multiplication factors, were calculated in the analyses to explain the variation of neutron multiplication factors with the ratio of water moderator to fuel volume in a unit fuel rod cell. Those results of analyses are considered to be useful for the confirmation of reasonableness and validity of criticality safety measures as data showing criticality characteristics for water-moderated heterogeneous lattice systems composed of the existing fuel rods in research and test reactors, of which criticality data are not sufficiently provided by the Criticality Safety Handbook.
Takeda, Tetsuaki*; Inagaki, Yoshiyuki; Aihara, Jun; Aoki, Takeshi; Fujiwara, Yusuke; Fukaya, Yuji; Goto, Minoru; Ho, H. Q.; Iigaki, Kazuhiko; Imai, Yoshiyuki; et al.
High Temperature Gas-Cooled Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.5, 464 Pages, 2021/02
As a general overview of the research and development of a High Temperature Gas-cooled Reactor (HTGR) in JAEA, this book describes the achievements by the High Temperature Engineering Test Reactor (HTTR) on the designs, key component technologies such as fuel, reactor internals, high temperature components, etc., and operational experience such as rise-to-power tests, high temperature operation at 950C, safety demonstration tests, etc. In addition, based on the knowledge of the HTTR, the development of designs and component technologies such as high performance fuel, helium gas turbine and hydrogen production by IS process for commercial HTGRs are described. These results are very useful for the future development of HTGRs. This book is published as one of a series of technical books on fossil fuel and nuclear energy systems by the Power Energy Systems Division of the Japan Society of Mechanical Engineers.
Collaborative Laboratories for Advanced Decommissioning Science; Tohoku University*
JAEA-Review 2020-032, 97 Pages, 2021/01
The Collaborative Laboratories for Advanced Decommissioning Science (CLADS), Japan Atomic Energy Agency (JAEA), had been conducting the Nuclear Energy Science & Technology and Human Resource Development Project (hereafter referred to "the Project") in FY2019. The Project aims to contribute to solving problems in the nuclear energy field represented by the decommissioning of the Fukushima Daiichi Nuclear Power Station, Tokyo Electric Power Company Holdings, Inc. (TEPCO). 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 "Basic Research on the Stability of Fuel Debris Including Alloy Phase" conducted in FY2019. In the present study, we focus on fuel debris consisting of oxide phase and alloy phase generated by the high-temperature chemical reaction between structure materials (SUS pipes, pressure vessels, etc.) and fuels (melted fuels, claddings components, etc.). We synthesize the simulated debris of UO-SUS system and UO-Zr(ZrO)-SUS system by high-temperature heat treatment, and measure their chemical property and dissolution behavior in water. Also, we will conduct research and development to spectroscopically analyze secular changes of oxide phase and alloy phase in the simulated debris.
Udagawa, Yutaka; Fuketa, Toyoshi*
Comprehensive Nuclear Materials, 2nd Edition, Vol.2, p.322 - 338, 2020/08
Collaborative Laboratories for Advanced Decommissioning Science; Tohoku University*
JAEA-Review 2019-035, 61 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 "Basic Research on the Stability of Fuel Debris Including Alloy Phase". In the present study, we focus on fuel debris consisting of oxide phase and alloy phase generated by the high-temperature chemical reaction between structure materials (SUS pipes, pressure vessels, etc.) and fuels (melted fuels, claddings components, etc.). We synthesize the simulated debris of UO-SUS system and UO-Zr(ZrO)-SUS system by high-temperature heat treatment, and measure their chemical property and dissolution behavior in water. Also, we will conduct research and development to spectroscopically analyze secular changes of oxide phase and alloy phase in the simulated debris.
Kurata, Masaki; Osaka, Masahiko; Jacquemain, D.*; Barrachin, M.*; Haste, T.*
Advances in Nuclear Fuel Chemistry, p.555 - 625, 2020/00
The importance of fuel chemistry has been revivaled since Fukushima-Daiichi Nuclear Power Station (FDNPS) accident. The inspection and analysis of damaged three units, which had been operated in March 11, 2011, showed large differences in the accident progression sequence for these units, because operators attempted to prevent or mitigate the accident progression of each unit by all means possible. Characteristics of fuel debris are considered to be largely influenced by the difference in the sequence and, hence, deviated from those predicted from prototypic accident scenarios, which had been mainly identified from the analysis of Three Mile Island-2 (TMI-2) accident and the following sim-tests. For the proper improvement of our knowledge on severe accident (SA), including non-prototypic conditions, one has to start improving the phenomenology of fuel/core degradation and fission product (FP) behavior. Advances in the chemistry is the most essential approach. The present review attempts to focus on the recent updates and remaining concerns after the FDNPS accident.
Takamatsu, Yuki*; Ishii, Hiroto*; Oishi, Yuji*; Muta, Hiroaki*; Yamanaka, Shinsuke*; Suzuki, Eriko; Nakajima, Kunihisa; Miwa, Shuhei; Osaka, Masahiko; Kurosaki, Ken*
Nihon Genshiryoku Gakkai Wabun Rombunshi, 17(3/4), p.106 - 110, 2018/12
In order to establish the synthesis method of simulated fuel contacting Cesium (Cs) which is required for the evaluation of physical/chemical characteristics in fuel and release behavior of Cs, sintering tests of the cerium dioxide (CeO) based simulated fuels containing Cesium iodide (CsI) are performed by using spark plasma sintering (SPS) method. The sintered CeO pellets with homogeneous distribution of several micro meter of CsI spherical precipitates were successfully obtained by optimizing SPS conditions.
Aihara, Jun; Ueta, Shohei; Nishihara, Tetsuo
JAEA-Technology 2015-040, 32 Pages, 2016/02
Original FORNAX-A is a calculation code for amount of fission product (FP) released from fuel rods of pin-in-type high temperature gas-cooled reactors (HTGRs). This report is for explanation what calculations become possible with minor changed FORNAX-A.
Tonoike, Kotaro; Okubo, Kiyoshi; Takada, Tomoyuki*
Proceedings of International Conference on Nuclear Criticality Safety (ICNC 2015) (DVD-ROM), p.292 - 300, 2015/09
The damaged Unit 1-3 reactors of the Fukushima Daiichi Nuclear Power Station may contain fuel debris of a significant amount that is in a form of molten-core-concrete-interaction (MCCI) product with porous structure. Such low density MCCI product including fissile material is a great concern for its criticality control, especially under submerged condition, due to its fairly good neutron moderation. This report shows computation results of basic criticality characteristics of the MCCI product, which will facilitate criticality risk assessments during decommissioning of the reactors. The results imply that water bound in concrete may raise the risk from the viewpoints of possibility of criticality events and of effectiveness of mitigation measures such as neutron poison injection into coolant water.
Nomoto, Yasunobu; Aihara, Jun; Nakagawa, Shigeaki; Isaka, Kazuyoshi; Ohashi, Hirofumi
JAEA-Data/Code 2015-008, 39 Pages, 2015/06
HTFP is a calculation code for amount of additionally released fission product (FP) from fuel rods of pin-in-type according to transient of core temperature at the accident of high temperature gas-cooled reactors (HTGRs). This code analyzes FP release inventory from core according to the transient of core temperature at the accident as an input data and considering FP release rate from a fuel compact and a graphite sleeve and radioactive decay of FP. This report describes the outline of HTFP code and its input data. The computed solutions using the HTFP code were compared to those of HTCORE code, which was used for the design of the High Temperature Engineering Test Reactor (HTTR) to validate the analysis models of the HTFP code. The comparison of HTFP code results with HTCORE code results showed the good agreement.
Shirasu, Noriko; Kurata, Masaki; Ogawa, Toru*
Proceedings of 2014 Water Reactor Fuel Performance Meeting/ Top Fuel / LWR Fuel Performance Meeting (WRFPM 2014) (USB Flash Drive), 6 Pages, 2014/09
In the accident of Fukushima-Daiichi Nuclear Power Plant, degraded fuels containing Zircaloy probably reacted with BC control blades containing stainless steel cladding or blade sheath. Since light elements like B and C are able to react easily with various elements and form various chemical species, several concerns are pointed out, such as variation in volatility and heat generation by oxidation of B and C. The chemical states of degraded fuel were evaluated on the assumption of thermodynamic equilibrium under various conditions of oxygen potential and temperature. The chemical behavior of B affects significantly the variation in oxygen potential with progressing severe accident, and many kinds of volatile compounds are formed by oxidation. The behavior of B causes the changes of volatility of FPs, such as Sr, Cs and Mo.
Oigawa, Hiroyuki; Yokoo, Takeshi*; Nishihara, Kenji; Morita, Yasuji; Ikeda, Takao*; Takaki, Naoyuki*
Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 6 Pages, 2005/10
The benefit of implementing Partitioning and Transmutation (P&T) of high-level wastes was parametrically surveyed. The possible reduction of the geological repository area was estimated. By recycling minor actinides (MA), the repository area required for unit spent fuel was reduced significantly in the case of MOX-LWR. This effect was caused by removal of Am which is a long-term heat source. By partitioning the fission products, in addition to MA recycling, further 70-80% reduction from the MA-recovery case can be expected for both UO and MOX. This significant reduction was independent of the cooling time before the partitioning process.
Nishihara, Tetsuo; Takeda, Tetsuaki
JAERI-Tech 2005-038, 40 Pages, 2005/07
R&D of the IS process technology and the integration technology of a high-temperature gas reactor and a heat utilization system is advanced by Japan Atomic Energy Research Institute. Hydrogen suuply system to fuel cell cars consists of off-site system and on-site system. The cost evaluation of on-site system must include the cost of storage, transportation, and refueling as well as that of production. Moreover, the hydrogen produced with the fossil fuel must consider the cost of carbon dioxide disposal. The cost evaluation data related to storage, transportaion and refueling in Japan and the USA are investigated. The cost evaluation of several off-site and on-siste systems are carried out.
Kaneko, Tetsuji; Tsukatani, Ichiro; Kiuchi, Kiyoshi
JAERI-Research 2005-005, 23 Pages, 2005/03
Fuel elements used in The Reduced-Moderation Water Reactor (RMWR) have the lamellar structure consisting of MOX pellets and UO blankets in order to attain the high breeding ratio and high burn-up simultaneously. It is a characteristic of the fuel elements that there is high thermal stress caused by inhomogeneous linear power density along the longitudinal direction of the fuel rod. Therefore, it is important to evaluate the local deformation behavior due to the transient temperature distribution. To estimate the thermal deformation behavior, the temperature and stress distribution of the fuel cladding tube assumed in the designed reactor were analyzed. Moreover, basic physical properties and mechanical properties for analyzing the deformation behavior were obtained by experiment using fuel cladding tubes made of candidate alloys. In addition, the appropriate experimental conditions for realizing the practical thermal deformation behavior of the fuel cladding tube was selected by adjusting the testing temperature distribution based on data obtained with thermal analysis.
Kaneko, Tetsuji; Tsukatani, Ichiro; Kiuchi, Kiyoshi
JAERI-Tech 2004-035, 18 Pages, 2004/03
Fuel elements used in the Reduced-Moderation Water Reactor (RMWR) have the stacking structure consisting of MOX pellets and UO blankets in a fuel rod in order to attain the high breeding ratio and high burn-up simultaneously. It is a characteristic of the fuel elements that there is high thermal stress caused by inhomogeneous linear power density along the longitudinal direction of the fuel rod in comparison with the present LWR fuels. For this reason, it is important to estimate local deformation behavior of the fuel cladding tube with temperature difference caused by MOX pellet and UO blanket. The testing machine was designed to investigate thermal-fatigue behavior under biaxial stress condition. The testing machine consists of the temperature distribution control unit, low cycle fatigue testing unit and internal pressure loading unit, it is also possible to conduct the simulation tests to investigate effects of pressure change with burn-up and longitudinal load change due to operation modes and restriction of fuel rods.
Mineo, Hideaki; Isogai, Hikaru; Morita, Yasuji; Uchiyama, Gunzo*
Journal of Nuclear Science and Technology, 41(2), p.126 - 134, 2004/02
A simple equation was proposed for the dissolution rate of spent LWR fuel, of which the change in the dissolution area was estimated by taking into account of the area of the cracks occurring due to thermal shrinkage of the pellets during irradiation. The applicability of proposed equation was examined using LWR fuel dissolution test results in the present study as well as the results obtained by other workers. The equation showed good agreements with the dissolution test results obtained from spent fuel pellets and pulverized spent fuel. It was indicated that the proposed equation was simple and would be useful for the prediction of dissolution of spent LWR fuels. However, the initial effective dissolution area, the parameter of the equation, was found to depend on the temperature, which could not be explained by the proposed equation. Further studies on the role of other factors affecting dissolution rate, such as nitrous acid, in the dissolution of spent fuel was required.
Minato, Kazuo; Hayashi, Hirokazu; Mizuguchi, Koji*; Sato, Takeyuki*; Amano, Osamu*; Miyamoto, Satoshi*
Proceedings of GLOBAL2003 Atoms for Prosperity; Updating Eisenhower's Global Vision for Nuclear Energy (CD-ROM), p.778 - 781, 2003/11
The simulation technology for the pyrochemical reprocessing of oxide fuels was developed to analyze experimental data, to predict experimental results, and to propose adequate conditions and processes. The simulation method was based on calculations of chemical equilibrium and electrochemical reactions. Some model calculations to simulate the experimental results were made on the process of electro-codeposition of UO and PuO. Although it was difficult to trace the experiments and compare the calculated results with the experimental results quantitatively due to the limitation of available data on the experimental conditions, the calculated results were consistent with the experimental results. The phenomena of the repeated oxidation-reduction reactions between Pu and Pu ions and those between Fe and Fe ions were theoretically analyzed,which caused the low current efficiency in the electro-codeposition process.
Sasa, Toshinobu; Oigawa, Hiroyuki; Tsujimoto, Kazufumi; Nishihara, Kenji; Kikuchi, Kenji; Kurata, Yuji; Saito, Shigeru; Futakawa, Masatoshi; Umeno, Makoto*; Ouchi, Nobuo; et al.
Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), 9 Pages, 2003/04
JAERI carries out research and development on accelerator-driven system (ADS) to transmute minor actinides and long-lived fission products in high-level radioactive waste. The system is composed of high intensity proton accelerator, lead-bismuth spallation target and lead-bismuth cooled subcritical core with nitride fuel. About 2500 kg of minor actinide is loaded into the subcritical core. Annual transmutation amount using this system is 250 kg with 800MW of thermal output. A superconducting linear accelerator with the beam power of 20 - 30MW is connected to drive the subcritical core. The nitride fuel without uranium, such as (Np, Am, Pu)N, is selected. The fuel irradiated in the ADS is reprocessed by pyrochemical process followed by the re-fabrication of the fuel. Many research and development activities are under way. Especially, to study and evaluate the feasibility of the ADS from physics and engineering aspects, the Transmutation Experimental Facility (TEF) is proposed under a framework of the High-Intensity Proton Accelerator Project.
Nakamura, Jinichi; Sugiyama, Tomoyuki; Nakamura, Takehiko; Kanazawa, Toru; Sasajima, Hideo
JAERI-Tech 2003-008, 32 Pages, 2003/03
no abstracts in English
Kusagaya, Kazuyuki*; Sugiyama, Tomoyuki; Nakamura, Takehiko; Uetsuka, Hiroshi
JAERI-Tech 2002-105, 24 Pages, 2003/01
High-temperature and high-pressure influence on the destructive force at the fuel rod failure in reactivity-initiated-accident (RIA) simulating experiment using the NSRR (Nuclear Safety Research Reactor) is estimated, for the purpose of mechanical designing of a new experimental capsule for simulating the temperature and pressure condition of typical commercial BWR. When knowledge on pressure impulse and water hammer, which are the cause of the destructive force, and steam property dependence on temperature and pressure are taken into account, one can qualitatively estimate that the destructive force in the BWR operation condition is smaller than that in the room temperature and atmospheric pressure condition. The water column velocity, which determines the impact by water hammer, is further investigated quantitatively by modeling the experimental system and the water hammer phenomenon. As a result, the maximum velocity of water column in the BWR operation condition is calculated to be only about 10% of that in the room temperature and atmospheric pressure condition.