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Audouin L.,Institute for Radiological Protection and Nuclear Safety | Chandra L.,Nuclear Research and Consultancy Group | Consalvi J.-L.,CNRS IUSTI – University Institute of Thermodynamic Industrial Systems) | Gay L.,Électricité de France | And 13 more authors.
Nuclear Engineering and Design | Year: 2011

The objective of this work was to quantify comparisons between several computational results and measurements performed during a pool fire scenario in a well-confined compartment. This collaborative work was initiated under the framework of the OECD fire research program and involves the most frequently used fire models in the fire community, including field and zone models. The experimental scenario was conducted at the French Institut de Radioprotection et de Sûreté Nucléaire (IRSN) and deals with a full-scale liquid pool fire in a confined and mechanically ventilated compartment representative for nuclear plants. The practical use of different metric operators and their ability to report the capabilities of fire models are presented. The quantitative comparisons between measurements and numerical results obtained from "open" calculations concern six important quantities from a safety viewpoint: gas temperature, oxygen concentration, wall temperature, total heat flux, compartment pressure and ventilation flow rate during the whole fire duration. The results indicate that it is important to use more than one metric for the validation process in order to get information on the uncertainties associated with different aspects of fire safety. © 2010 Elsevier B.V. All rights reserved.


Perez M.,University of Barcelona | Reventos F.,University of Barcelona | Batet L.,University of Barcelona | Guba A.,AEKI | And 20 more authors.
Nuclear Engineering and Design | Year: 2011

This paper presents the results and the main lessons learnt from Phase V of BEMUSE, an international programme promoted by the Working Group on Accident Management and Analysis (GAMA) of OECD to address the issue of the capabilities of best-estimate computational tools and uncertainty analysis. The scope of Phase V is the uncertainty analysis of a Large Break Loss-Of-Coolant-Accident (LBLOCA) in a Pressurized Water Reactor. Fourteen participants from twelve organizations and ten countries participated in the Phase V of BEMUSE. The paper starts with a general description of the BEMUSE programme including the objectives, structure, and the outline of the Phase V specification. Then it summarizes some general aspects on the uncertain model parameters and the results for the uncertainty analysis and for the sensitivity evaluation. To end with, general recommendations and conclusions are presented as practical guidance for uncertainty analysis performance. © 2011 Elsevier B.V. All rights reserved.


Noirot J.,CEA Cadarache Center | Lamontagne J.,CEA Cadarache Center | Nakae N.,JNES | Kitagawa T.,MNF | And 2 more authors.
Journal of Nuclear Materials | Year: 2013

A UO2 fuel with a heterogeneous distribution of 235U was irradiated up to a high burn-up in the Halden Boiling Water Reactor (HBWR). The last 100 days of irradiation were performed with an increased level of linear power. The effect of the heterogeneous fissile isotope distribution on the formation of the HBS was studied free of the possible influence of Pu which exists in heterogeneous MOX fuels. The HBS formed in 235U-rich agglomerates and its main characteristics were very similar to those of the HBS formed in Pu-rich agglomerates of heterogeneous MOX fuels. The maximum local contents of Nd and Xe before HBS formation were studied in this fuel. In addition to a Pu effect that promotes the HBS phenomenon, comparison with previous results for heterogeneous MOX fuels showed that the local fission product concentration was not the only parameter that has to be taken into consideration. It appears that the local actinide depletion by fission and/or the energy locally deposited through electronic interactions in the fission fragment recoils also have an effect on the HBS formation threshold. Moreover, a major release of fission gases from the peripheral 235U-rich agglomerates of HBS bubbles and a Cs radial movement are also evidenced in this heterogeneous UO2. Cs deposits on the peripheral grain boundaries, including the HBS grain boundaries, are considered to reveal the release paths. © 2013 Elsevier B.V. All rights reserved.


Adorni M.,UNIPI | Herranz L.E.,CIEMAT | Hollands T.,GRS Society for plants and Reactor Safety | Ahn K.-II.,Korea Atomic Energy Research Institute | And 11 more authors.
Nuclear Engineering and Design | Year: 2016

The OECD/NEA Sandia Fuel Project provided unique thermal-hydraulic experimental data associated with Spent Fuel Pool (SFP) complete drain down. The study conducted at Sandia National Laboratories (SNL) was successfully completed (July 2009 to February 2013). The accident conditions of interest for the SFP were simulated in a full scale prototypic fashion (electrically heated, prototypic assemblies in a prototypic SFP rack) so that the experimental results closely represent actual fuel assembly responses. A major impetus for this work was to facilitate severe accident code validation and to reduce modeling uncertainties within the codes. Phase I focused on axial heating and burn propagation in a single PWR 17 × 17 assembly (i.e. “hot neighbors” configuration). Phase II addressed axial and radial heating and zirconium fire propagation including effects of fuel rod ballooning in a 1 × 4 assembly configuration (i.e. single, hot center assembly and four, “cooler neighbors”). This paper summarizes the comparative analysis regarding the final destructive ignition test of the phase I of the project. The objective of the benchmark is to evaluate and compare the predictive capabilities of computer codes concerning the ignition testing of PWR fuel assemblies. Nine institutions from eight different countries were involved in the benchmark calculations. The time to ignition and the maximum temperature are adequately captured by the calculations. It is believed that the benchmark constitutes an enlargement of the validation range for the codes to the conditions tested, thus enhancing the code applicability to other fuel assembly designs and configurations. The comparison of lumped parameter and CFD computer codes represents a further valuable achievement. © 2016 Elsevier B.V.


Ebisawa K.,Japan Nuclear Energy Safety Organization | Fujita M.,JNES | Iwabuchi Y.,JNES | Sugino H.,JNES
Nuclear Engineering and Technology | Year: 2012

The Tohoku earthquake (Mw9.0) occurred on March 11, 2011 and caused a large tsunami. The Fukushima Dai-ichi NPP (F1-NPP) were overwhelmed by the tsunami and core damage occurred. This paper describes the overview of F1-NPP accident and the usability of tsunami PRA at Tohoku earthquake. The paper makes reference to the following current issues: influence on seismic hazard of gigantic aftershocks and triggered earthquakes, concepts for evaluating core damage frequency considering common cause failure with correlation coefficient against seismic event at multi units and sites, and concepts of "seismic-tsunami PSA" considering a combination of seismic motion and tsunami effects.

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