In the Best-Estimate Plus Uncertainty (BEPU) framework, the use of best-estimate code requires to go through a Verification, Validation and Uncertainty Quantification process (VVUQ). The relevance of the experimental data in relation to the physical phenomena of interest in the VVUQ process is crucial. Adequacy analysis of selected experimental databases addresses this problem. The outcomes of the analysis can be used to select a subset of relevant experimental data, to encourage designing new experiments or to drop some experiments from a database because of their substantial lack of adequacy. The development of a specific transparent and reproducible approach to analyze the relevance of experimental data for VVUQ still remains open and is the topic of this contribution.In this paper, the concept of adequacy initially introduced in the OECD/NEA SAPIUM (Systematic APproach for model Input Uncertainty quantification Methodology) activity is formalized. It is defined through two key properties, called representativeness and completeness, that allows considering the multifactorial dimension of the adequacy problem. A new systematic approach is then proposed to analyze the adequacy of a set of experimental databases. It relies on the introduction of two sets of criteria to characterize representativeness and completeness and on the use of multi-criteria decision analysis method to perform the analysis. Finally, the approach is applied in the framework of the new OECD/NEA ATRIUM activity which includes a set of practical IUQ exercises in thermal-hydraulics to test the SAPIUM guideline in determining input uncertainties and forward propagating them on an application case. It allows evaluating the adequacy of eight experimental databases coming from the Super Moby-dick, Sozzi-Sutherland and Marviken experiments and identifying the most adequate ones.

In boiling water reactors (BWRs), the fuel assembly incorporates nonuniform elements such as part-length fuel rods and water rods capable of modifying the moderator density distribution and axial pressure drop in order to improve the economic efficiency and the thermal margin. This study focuses on the void fraction distribution in a rod bundle with different nonheated rod arrangements in order to evaluate the effect of water rods as a non-heated section on the boiling flow dynamics in the rod bundle. A boiling flow experiment was conducted using the test facility for 3D thermal hydraulics in light water reactors (SIRIUS-3D) and a linear accelerator-driven high-energy X-ray computed tomography (CT) system under BWR rated pressure conditions. The test section was a 5 x 5 rod bundle of heated length 3708 mm that simulated a BWR rod bundle. The void fraction distri-bution in the 5 x 5 rod bundle was acquired at six heights via X-ray CT imaging. The experimental results show the effect of nonheated rod arrangements on the local void fraction and the evolution of boiling flow in the rod bundle.

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. (C) 2011 Elsevier B.V. All rights reserved.

Phase IV of BEMUSE Program is a necessary step for a subsequent uncertainty analysis. It includes the simulation of the reference scenario and a sensitivity study. The scenario is a LBLOCA and the reference plant is Zion 1 NPP, a 4 loop PWR unit. Thirteen participants coming from ten different countries have taken part in the exercise. The BEMUSE (Best Estimate Methods plus Uncertainty and Sensitivity Evaluation) Program has been promoted by the Working Group on Accident Management and Analysis (WGAMA) and endorsed by the Committee on the Safety of Nuclear Installations (CSNI). The paper presents the results of the calculations performed by participants and emphasizes its usefulness for future uncertainty evaluation, to be performed in next phase. The objectives of the activity are basically to simulate the LBLOCA reproducing the phenomena associated to the scenario and also to build a common, well-known, basis for the future comparison of uncertainty evaluation results among different methodologies and codes. The sensitivity calculations performed by participants are also presented. They allow studying the influence of different parameters such as material properties or initial and boundary conditions, upon the behaviour of the most relevant parameters related to the scenario.

Divertor surface of a magnetic confinement fusion reactor is exposed to strong radiative heating. According to standard design of the ITER, maximum heat flux on the divertor surface becomes locally near 30 MW m−2. To cool such high heat flux surface by water flow, it is necessary to establish a cooling method which enhanced the critical heat flux (CHF). We proposed a cooling by a planar impinging jet with free surface in the previous report. In the jet cooling on flat surface, high CHF was obtained in the limited region where the jet flow hits directly. As apart from the region, the CHF decreases abruptly with the distance from the center. To overcome this difficulty, it was proposed that the planar jet is applied to cool concave surface where the centrifugal force is efficiently used to enhance the CHF. In this study, the CHFs were investigated in the confined jet flow which was guarded by a wall on the other side of the heated wall, because the guard wall works to protect splash of water from liquid film by violent boiling and expects further enhancement of the CHF. In this study, the CHFs were investigated in the confined flow of two-dimensional jet on flat and concave surfaces in the various flow conditions and got a correlation for the CHF. Applicability of this cooling for divertor surface was assessed by using the experimental results.

Divertor surface of a fusion reactor is exposed to strong surface heating by high flux charged particles. According to typical design of ITER, the heat flux on divertor surface becomes locally near 20 MW m−2. Then, it is necessary to establish a cooling method to cool such high heat flux surface. A cooling by a planar impinging jet has been proposed. However, in the application of the impinging jet cooling to flat heated surface, high CHF is obtained in the limited region near center. As apart from the region, CHF decreases abruptly with a distance from the center. To overcome this difficulty, we proposed that an impinging planar jet was applied to a concave heated wall where a flow induced centrifugal force acts effectively along the curved surface. In a previous paper, we investigated the CHF for a free planer jet on the concave surface. In this study, CHFs in a confined planar jet, were investigated as a function of distance from the center in the various subcooling, flow velocity and curvature of cooling surface and compared with the CHF in the free planar jet cooling. To investigate CHF in the abnormal condition like mild plasma disruption, CHFs in the transient power input were investigated and discussed in comparison with the steady state CHF. Finally application by both impinging planar jets to the divertor surface are discussed.