Kawada, Kenichi; Suzuki, Toru*
Nuclear Engineering and Technology, 53(12), p.3930 - 3943, 2021/12
To improve the capability of the SAS4A code, which simulates the initiating phase of core disruptive accidents for MOX-fueled Sodium-cooled Fast Reactors (SFRs), the authors have investigated in detail the physical phenomena under unprotected loss-of-flow (ULOF) conditions in a previous paper. As the conclusion of the last article, fuel stub motion, in which the residual fuel pellets would move toward the core central region after fuel pin disruption, was identified as one of the key phenomena to be appropriately simulated for the initiating phase of ULOF. In the present paper, a simple model describing fuel stub motion, which was not modeled in the previous SAS4A code, was newly proposed. The applicability of the proposed model was validated through a series of analyses for the CABRI experiments, by which the stub motion would be represented with reasonable conservativeness for the reactivity evaluation of disrupted core.
Takeda, Takeshi; Otsu, Iwao
Nuclear Engineering and Technology, 50(6), p.829 - 841, 2018/08
Takeda, Takeshi; Otsu, Iwao
Nuclear Engineering and Technology, 49(5), p.928 - 940, 2017/08
Zheng, X.; Ishikawa, Jun; Sugiyama, Tomoyuki; Maruyama, Yu
Nuclear Engineering and Technology, 49(2), p.434 - 441, 2017/03
Suzuki, Toru; Tobita, Yoshiharu; Kawada, Kenichi; Tagami, Hirotaka; Sogabe, Joji; Matsuba, Kenichi; Ito, Kei; Ohshima, Hiroyuki
Nuclear Engineering and Technology, 47(3), p.240 - 252, 2015/04
Cheng, S.; Yamano, Hidemasa; Suzuki, Toru; Tobita, Yoshiharu; Nakamura, Yuya*; Zhang, B.*; Matsumoto, Tatsuya*; Morita, Koji*
Nuclear Engineering and Technology, 45(3), p.323 - 334, 2013/06
Smith, G.*; Kato, Tomoko
Nuclear Engineering and Technology, 42(1), p.1 - 8, 2010/02
Geological disposal is designed to provide safe containment of radioactive waste for very long times, with the containment provided by a combination of engineered and geological barriers. In the extreme long term, after many thousands of years or longer, residual amounts of long-lived radionulides such as Cl-36, but also radionuclides in the natural decay chains, may be released into the environment normally accessed and used by humans, termed here, the biosphere. It is necessary to ensure that any such releases meet radiation protection objectives through the development of a safety case, which will include assessment of radiation doses to humans. The design of such dose calculations over such long timeframes is not straightforward, because of the range of potentially relevant assumptions which could be made, concerning environmental change and changes in human behavior. The issue has therefore been subject to international cooperation for many years. This paper summarizes the evolution and results of that collaboration leading up to the present day, taking account of developments in international recommendations on radiation protection objectives and the more recent greater focus on preparation of site specific safety cases.
Sawada, Atsushi; Sato, Hisashi
Nuclear Engineering and Technology, 42(1), p.9 - 16, 2010/02
Experimental studies for evaluating fracture were conducted by using transparent replica of a single fracture, in order to obtain data for contributing to the methodology improving how we define representative parameter values used for a parallel plate fracture model. Quantitative aperture distribution and tracer concentration data were obtained by measuring the attenuation of transmitted light through the fracture, in high spatial resolution. The representative aperture values evaluated from the multiple measurement methods, such as arithmetic mean of aperture distribution, transport aperture, average aperture evaluated from fracture void volume measurement, converged to a unique value, which indicates the accuracy of this experimental study. The aperture data was employed for studying the validity of numerical simulation under the assumption of local cubic law and showed that the calculated flow rate through the fracture is 10% - 100% larger than hydraulic test results.
Chuto, Toshinori; Nagase, Fumihisa; Fuketa, Toyoshi
Nuclear Engineering and Technology, 41(2), p.163 - 170, 2009/03
In order to evaluate high-temperature oxidation behavior of the advanced alloy cladding under LOCA conditions, isothermal oxidation tests in steam were performed with cladding specimens prepared from high burnup PWR fuel rods (79 MWd/kg). Cladding materials were M5 and ZIRLO which are Nb-containing alloys. Ring-shaped specimens were isothermally oxidized in flowing steam at temperatures from 1173 to 1473 K for the duration between 120 and 4000 s. Protective effect of the pre-formed corrosion layer is seen for the shorter time range at the lower temperatures. Influence of pre-hydriding is not significant for the examined range. Alloy composition change generally has small influence on the oxidation in the examined temperature range, though M5 shows obviously smaller oxidation constant at 1273 K. Consequently, the oxidation rates of the high burnup M5 and ZIRLO cladding are comparable or lower than that of unirradiated Zircaloy-4 cladding.
Nuclear Engineering and Technology, 39(6), p.683 - 690, 2007/12
To make nuclear energy truly sustainable, it is necessary to ensure not only the sustainability of the fuel supply but also the sustained availability of waste repositories, especially those for high-level radioactive waste (HLW). From this perspective, the effort to maximize the waste loading density in a given repository is important for easing repository capacity problems. In most cases, the loading of a repository is controlled by the decay heat of the emplaced waste. In this paper, a comparison of the decay heat characteristics of HLW is made among the various fuel cycle options. It is suggested that, for a future fast breeder reactor (FBR) cycle, the removal and burning of minor actinides (MA) would significantly reduce the heat load in waste and would allow for a reduction of repository size by half.
Ichimiya, Masakazu; Mizuno, Tomoyasu; Kotake, Shoji
Nuclear Engineering and Technology, 39(3), p.171 - 186, 2007/06
Critical issues in the development targets for the future fast reactor (FR) cycle system, including odium-cooled FR were to ensure safety assurance, efficient utilization of resources, reduction of environmental burden, assurance of nuclear non-proliferation, and economic competitiveness. A promising design concept of sodium-cooled fast reactor JSFR is proposed aiming at fully satisfaction of the development targets for the next generation nuclear energy system. A roadmap toward JSFR commercializationis described, to be followed up in a new framework of the Fast reactor Cycle Technology development (FaCT) Project launched in 2006.
Kunitomi, Kazuhiko; Yan, X.; Nishihara, Tetsuo; Sakaba, Nariaki; Mori, Tomoaki
Nuclear Engineering and Technology, 39(1), p.9 - 20, 2007/02
Design study on the Gas Turbine High Temperature Reactor 300-Cogeneration (GTHTR300C) aiming at producing both electricity by a gas turbine and hydrogen by a thermochemical water splitting method (IS process method) has been conducted. It is expected to be one of the most attractive systems to provide hydrogen for fuel cell vehicles after 2020s. The GTHTR300C employs a block type Very High Temperature Reactor (VHTR) with thermal power of 600MW and outlet coolant temperature of 950C. The maximum 370 MW to the secondary system is used for hydrogen production and the balance of the reactor thermal power is used for electricity generation. This paper describes the original design features focusing on the plant layout and plant cycle of the GTHTR300C together with present development status of the gas turbine, IHX, etc.
Miyamoto, Yoichi; Umeki, Hiroyuki; Osawa, Hideaki; Naito, Morimasa; Nakano, Katsushi; Makino, Hitoshi; Shimizu, Kazuhiko; Seo, Toshihiro
Nuclear Engineering and Technology, 38(6), p.505 - 534, 2006/08
Ensuring sufficient supplies of clean, economic and acceptable energy is a critical global challenge for the 21st century. There seems little alternative to a greatly expanded role for nuclear power, but implementation of this option will depend on ensuring that all resulting wastes can be disposed of safely. Although there is a consensus on the fundamental feasibility of such disposal by experts in the field, concepts have to be developed to make them more practical to implement and, in particular, more acceptable to key stakeholders. By considering global trends and using illustrative examples from Japan, key areas for future R&D are identified and potential areas where the synergies of international collaboration would be beneficial are highlighted.
Yoshida, Hiroyuki; Tamai, Hidesada; Onuki, Akira; Takase, Kazuyuki; Akimoto, Hajime
Nuclear Engineering and Technology, 38(2), p.119 - 128, 2006/04
The reduced-moderation water reactor core adopts a hexagonal tight-lattice arrangement. In the core, there is no sufficient information about the effects of the gap spacing and grid spacer configuration on the flow characteristics. Thus, we start to develop a predictable technology for thermal-hydraulic performance of the core using an advanced numerical simulation technology. As a part of this technology development, we are developing a two-phase flow simulation code TPFIT with an advanced interface tracking method. The vector and parallelization of the code was conducted to fit the large-scale simulations. The numerical results applied to large-scale water-vapor two-phase flow in tight lattice rod bundles are shown and compared with experimental results. In the results, a tendency of the predicted void fraction distribution in horizontal plane agreed with the measured values including the bridge formation of the liquid at the position of adjacent fuel rods where an interval is the narrowest.
Nuclear Engineering and Technology, 37(6), p.537 - 556, 2005/12
A primary-pipe rupture accident is one of the design-basis accidents of a High-Temperature Gas-cooled Reactor (HTGR). When the primary-pipe rupture accident occurs, air is expected to enter the reactor core from the breach and oxidize in-core graphite structures. This paper describes an experiment and analysis of the air ingress phenomena and the method for the prevention of air ingress into the reactor during the primary-pipe rupture accident. The numerical results are in good agreement with the experimental ones regarding the density of the gas mixture, the concentration of each gas species produced by the graphite oxidation reaction and the onset time of the natural circulation of air. A hydrogen production system connected to the High-Temperature Engineering Test Reactor (HTTR) is being designed to be able to produce hydrogen by themo-chemical Iodine-Sulfur process, using a nuclear heat of 10 MW supplied by the HTTR. The HTTR hydrogen production system is first connected to a nuclear reactor in the world; hence a permeation test of hydrogen isotopes through heat exchanger is carried out to obtain detailed data for safety review and development of analytical codes. This paper also describes an overview of the hydrogen permeation test and permeability of hydrogen and deuterium of Hastelloy XR.
Homma, Toshimitsu; Tomita, Kenichi*; Hato, Shinji*
Nuclear Engineering and Technology, 37(3), p.245 - 258, 2005/06
This paper addresses two types of uncertainty: stochastic uncertainty and subjective uncertainty in probabilistic accident consequence assessments. The off-site consequence assessment code OSCAAR has been applied to uncertainty and sensitivity analyses on the individual risks of early fatality and latent cancer fatality in the population due to a severe accident. A new stratified meteorological sampling scheme was successfully implemented into the trajectory model for atmospheric dispersion and the statistical variability of the probability distributions of the consequence was examined. A total of 65 uncertain input parameters was considered and 128 runs of OSCAAR were performed in the parameter uncertainty analysis. The study provided the range of uncertainty for the expected values of individual risks of early and latent cancer fatality close to the site. In the sensitivity analyses, the correlation/regression measures were useful for identifying those input parameters whose uncertainty makes an important contribution to the overall uncertainty for the consequence.