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• Flow and rupture study of liquid film on corrugated plate wall are reviewed. • Chaotic analysis of film thickness series will be a prospect research direction. • Film rupture mechanism cannot be explained by surface wave disturbance model. • More study will be focused on rupture phenomenon at the corrugated plate corner. Steam generated by steam generator in a nuclear power plant must firstly be subjected to steam-water separation treatment before entering the turbine. The research on flow and rupture characteristics of the liquid film on the wall surface of corrugated plate has important and practical significance for improving the separation efficiency. Since the 1940s, researchers from the United States, France, Japan and other countries have carried out related researches. With the rapid development of steam-water separation technology, many scholars have carried out a large number of experiments, theoretical study and numerical calculations on this phenomenon. Based on the published papers, the research progress and significant achievements of the flow and rupture characteristics of the liquid film on corrugated plate wall are reviewed. Shortcomings of past research and prospects for future research are summarized and proposed, respectively, which provides a reference for improvement of separation efficiency of the corrugated plate dryer. [ABSTRACT FROM AUTHOR]

Abstract: The isotopic composition of spent nuclear fuels is vital data for studies on the nuclear fuel cycle and reactor physics. The Japan Atomic Energy Agency (JAEA) has been active in obtaining such data for pressurized water reactor (PWR) and boiling water reactor (BWR) fuels, and some data has already been published. These data have been registered with the international Spent Fuel Isotopic Composition Database (SFCOMPO) and widely used as international benchmarks for burnup calculation codes and libraries. In this paper, Assay Data of Spent Nuclear Fuel from two fuel assemblies irradiated in the Ohi-1 and Ohi-2 PWRs in Japan are shown. The destructive assay data from Ohi-2 have already been published. However, these data were not suitable for the benchmarking of calculation codes and libraries because several important specifications and data were not included. This paper summarizes the details of destructive assay data and specifications required for analyses of isotopic composition from Ohi-1 and Ohi-2. For precise burnup analyses, the burnup values of destructive assay samples were re-evaluated in this study. These destructive assay data were analyzed using the SWAT2.1 code, and the calculation results showed good agreement with experimental results. This indicates that the quality of destructive assay data from Ohi-1 and Ohi-2 PWRs is high, and that these destructive assay data are suitable for the benchmarking of burnup calculation code systems. [Copyright &y& Elsevier]

• Data scarcity of the Fukushima Unit 1 accident poses challenges for interpretation. • A postulated scenario has been proposed matching key accident footprints. • Fission products results are consistent with the measured dose rates. • Wetwell and cavity modeling have been shown to be highly influential. The accident occurred in Japan on 11 March 2011 has revived the interest for the analysis of severe accidents. The scarce and sometimes unreliable data concerning boundary conditions, effectiveness of accident management measures and equipment performance, pose a tough challenge in modelling the accident scenarios. Throughout an analysis of the challenges posed by the Fukushima Daiichi Unit 1 data recorded, this paper describes the major postulates proposed by CIEMAT concerning the equipment and component responses, the effectiveness of accident management actions and the MELCOR model applied. Among most influencing assumptions are those related to the reactor pressure vessel (RPV) leaking pathways and failure mode, the water flow rate entering the reactor, the potential leaking pathways and failure mode and location from the primary containment vessel (PCV) to the reactor building, the corium relocation from RPV to the cavity and its distribution in the PCV, the potential stratification of the suppression pool and the hypotheses made a priori concerning fission product release and transport. Based on the postulated scenario and model, a remarkable agreement of the thermal footprints in terms of RPV and PCV pressures during 500 h has been achieved, in which the RPV and PCV leaks/failures as well as venting played a determining role in the short run of the accident and water injection heavily conditioned the long one. As for the scarce data related to fission products (FP), a consistent agreement is found in the suppression chamber, but estimates in the Dry-Well are about an order of magnitude below measurements despite showing the observed trend. A number of factors might affect FP comparisons to data, from the approximate method to derive dose rates (measurements) from FP masses (MELCOR results) to the RPV and PCV postulated failures. Anyway, based on the data available the set of hypotheses and approximations made seem to make up a defensible scenario for Fukushima Daiichi Unit 1. The studies and results presented in this paper have been achieved under the frame of the OECD/BSAF projects through the CSN-CIEMAT collaboration agreement on severe accidents research. [ABSTRACT FROM AUTHOR]

Abstract: Although MOX fuel coming from reprocessed depleted uranium fuels has been used as a recycling strategy by countries like France and Japan it is not a common policy in the 30 countries that uses nuclear power, nowadays it seems to be a more direct alternative to reduce the depleted fuel interim storage. Previously, the spent fuel pools of Laguna Verde Nuclear Power plant were redesigned to host the total operating life depleted fuel under its original nominal power condition, however the plant has been up-rated to 120% of its original nominal power increasing the number of depleted fuel forecasted. This new situation makes necessary the analysis of alternatives, being one of them recycling. The current paper assesses the viability of using MOX fuel in the up-rated Power Plant; the design of the boiling water reactor MOX fuel addresses the two main constraints of its use: shutdown margin and reactor stability. Fuel design proposed sets the appropriate MOX enrichment and the maximum MOX fuel batch reload that does not imply any modification to the reactor control systems to avoid an extra economical cost due to its use. [Copyright &y& Elsevier]

Abstract: This paper presents an integrated multicriteria analysis method for the quantitative evaluation of a state’s nuclear nonproliferation credibility level. Underscoring the implications of policy on the sources of political demand for nuclear weapons rather than focusing on efforts to restrict the supply of specific weapons technology from the “haves” to the “have-nots”, the proposed methodology considers the political, social, and cultural dimensions of nuclear proliferation. This methodology comprises three steps: (1) identifying the factors that influence credibility formation and employing them to construct a criteria tree that will illustrate the relationships among these factors; (2) defining the weight coefficients of each criterion through pairwise comparisons of the Analytical Hierarchy Process (AHP); and (3) assigning numerical scores to a state under each criterion and combining them with the weight coefficients in order to provide an overall assessment of the state. The functionality of this methodology is examined by assessing the current level of nuclear nonproliferation credibility of four countries: Japan, North Korea, South Korea, and Switzerland. [Copyright &y& Elsevier]

• Experimental and analytical evaluation of the effects of nitrogen intrusion in an RPV bottom leak scenario. • The ASVAD safety valve, an innovative passive system to avoid the injection of nitrogen from the accumulators is presented. • Benefits of avoiding the injection of nitrogen thanks to the installation of the ASVAD valve. Decay heat removal can be seriously degraded by the presence of non-condensable gases in the cooling circuits. Nitrogen gas may be pushed into the primary system after a full discharge of the accumulators. This may produce various adverse effects: the interruption of natural circulation, the limitation of the primary to secondary heat transfer during the reflux cooling and prevent the startup of the active injection by stabilization of the pressure above the injection set point. State-of-the-art system codes have proven to be capable to simulate non-condensable gas effects in accident situations. The ASVAD valve, has been designed to avoid the inflow of nitrogen into the primary system by means of passive concepts. This paper addresses the complications derived from the nitrogen and evaluates the ASVAD valve performance through the simulation of a vessel bottom leak experiment at the LSTF facility in Japan. [ABSTRACT FROM AUTHOR]

• The complete CFD model of Monju reactor upper plenum was established. • The CFD steady state calculated results agreed well with experimental data. • The transient formation process of thermal stratification were obtained. • The main factors that affects the thermal stratification process were analyzed. In this paper, the IAEA benchmark, a trip transient scenario from 40% rated power of Monju reactor in Japan, was studied using CFD method. The reactor steady state and shutdown transient scenarios were simulated and the 3-D distributions of important thermal hydraulic parameters in the SFR upper plenum was achieved. The steady state calculated results agreed well with experimental data and acceptable deviations were obtained, which indicate the application and prediction accuracy of CFD method. The complex development of thermal stratification in the upper plenum of SFR under accident conditions were obtained. The detailed flow fields in the bottom of center measuring column and near with the through flow holes were revealed. The influence of different local structures on the thermal stratification in the SFR upper plenum were understood. This work could provide a certain basis for the SFR upper plenum structure design and operation strategy making. [ABSTRACT FROM AUTHOR]