Volume 75, Issue 1-2

Learning from operational experience and applying this knowledge promptly and intelligently is one of the ways to improve the safety of Nuclear Power Plant (NPP). Recent reviews of the effectiveness of Operational Experience Feedback (OEF) systems have pointed to the need for further improvement, with importance being placed on tailoring the information to the needs of the regulators. In 2007, at the request of a number of nuclear safety regulatory authorities in Europe, the Institute for Energy of the European Commission's Joint Research Centre (EC JRC) initiated a project on Nuclear Power Plant operational experience feedback, which adopts an integrated approach to the research needed to strengthen the European capabilities for assessment of NPP operational events and to promote the development of tools and mechanisms for the improved application of the lessons learned. Consequently, a so-called “European Clearinghouse” on NPP OEF was established, which includes scientific officers from the EC JRC, a number of European nuclear safety regulatory authorities and some of their Technical Support Organizations (TSOs). The paper discusses the activities implemented in 2008 within the framework of the European Clearinghouse on NPP OEF (hereinafter called the European NPP Clearinghouse) and provides an overview of the main conclusions drawn from the safety studies performed. Outlook of the activities carried out in 2009 are given.

As a consequence of a total loss of AC power supply (station blackout) at a VVER-1000 leading to unavailability of major active safety systems, the safety criteria ensuring the safe operation of the nuclear power plant would be violated and core heat-up with possible core degradation could occur. A dedicated accident management measure (primary side depressurization) can be applied to reduce the primary pressure and to activate the injection from the passive emergency core cooling systems (accumulators). The analyses presented in this paper are aiming at both a detailed investigation of the accident sequence, taking into account the depressurization of the primary circuit, and the possibilities to prevent or at least to mitigate a damage of the reactor core so as to gain additional time for taking necessary countermeasures. The analyses are performed using the codes ASTEC and ATHLET developed by IRSN (Institut de Radioprotection et de Sûreté Nucléaire) and GRS (Gesellschaft für Anlagen- und Reaktorsicherheit mbH).

The performed thermal-hydraulic analysis of the Accident Localisation System (ALS) has shown, that in the case of a Maximum Design Basis Accident (MDBA), in the long term, a significant amount of energy (∼40%) is transferred to the building structures. Therefore, special attention should be paid to the correct simulation of the building structures and heat transfer processes. Free convection, forced convection, condensation and conductivity are the phenomena, which are connected with the heat transfer to the building structures and should be investigated. In this paper, the heat transfer models available in the COCOSYS code, which is applied for the analysis, are discussed. The influence of the heat transfer models on the behaviour of thermal-hydraulic parameters in ALS compartments during MDBA is investigated.

The integrity of fuel rod claddings during a loss of coolant accident (LOCA) is usually assessed by comparing calculated stresses or strains with the corresponding failure thresholds. Alternatively, the cladding integrity could also be evaluated on the basis of the time elapsed since the LOCA onset. To apply this assessment criterion the relationship between failure time and cladding loading must be available. The relationship can be established computationally by monitoring the evaluation of cladding inner and outer radius during the high temperature creep process. The cladding is classified as defective when the predicted inner radius equals or exceeds the anticipated outer radius. Using a suitable creep rate correlation it is shown which results are yielded by this procedure.

A description of the selected decommissioning strategy for the research reactor MR at the site of the Kurchatov Institute in Moscow is presented. The MR reactor hall is planned to be used as a temporary fuel store for the other research reactors on the site. On the basis of the site-specific conditions and overall decommissioning goals, it was identified that phased immediate decommissioning is the preferable option. The current status of the reactor, expected final conditions and the sequence of decommissioning works are shown.

This paper describes and investigates upgrading of the silicon doping facility at the ETRR-2 research reactor to accommodate larger size silicon ingots in the two positions available for irradiation in the thermal column of the reactor. Two methods of irradiation are proposed. Both are based on irradiating two ingots simultaneously in one position. After a certain irradiation time, the silicon ingots are reshuffled inside the irradiation channel. The results obtained show that the proposed irradiation strategies of doping silicon ingots are simple to handle and result in acceptable values of resistivity variation.

The Particle Swarm Optimization (PSO) algorithm is used to optimize the design of shell-and-tube heat exchangers and determine the optimal feasible solutions so as to eliminate trial-and-error during the design process. The design formulation takes into account the area and the total annual cost of heat exchangers as two objective functions together with operating as well as geometrical constraints. The Nonlinear Constrained Single Objective Particle Swarm Optimization (NCSOPSO) algorithm is used to minimize and find the optimal feasible solution for each of the nonlinear constrained objective functions alone, respectively. Then, a novel Nonlinear Constrained Multobjective Particle Swarm Optimization (NCMOPSO) algorithm is used to minimize and find the Pareto optimal solutions for both of the nonlinear constrained objective functions together. The experimental results show that the two algorithms are very efficient, fast and can find the accurate optimal feasible solutions of the shell and tube heat exchangers design optimization problem.

Using certain well-known properties of Chebyshev polynomials, an efficient analytical approach to evaluate the Boltzmann equation is presented in one-dimensional spherical geometry. The procedure is based on the expansion of the angular flux in terms of a series of Chebyshev polynomials.

In the present work we propose a heuristic construction of a transport equation for neutrons with anisotropic scattering considering only the radial cylinder dimension. The eigenvalues of the solutions of the equation correspond to the positive values for the one dimensional case. The central idea of the procedure is the application of the S N method for the discretisation of the angular variable followed by the application of the zero order Hankel transformation. The basis the construction of the scattering terms in form of an integro-differential equation for stationary transport resides in the hypothesis that the eigenvalues that compose the elementary solutions are independent of geometry for a homogeneous medium. We compare the solutions for the cartesian one dimensional problem for an infinite cylinder with azimuthal symmetry and linear anisotropic scattering for two cases.

In this work a general analytical solution is reported, expressed in integral form for the time-dependent, one-dimensional S N transport equation in cartesian geometry valid for bounded and unbounded domains (0 < x <), using the double Laplace transform technique. The main idea consists in the application of the Laplace transform technique in time variable and the solution of the resulting equation by the LTS N method using appropriate boundary conditions for bounded and unbounded domain problems. We also report about the numerical simulations carried out.

In order to check new calculation results on critical slabs, numerical calculations have been done with a discrete ordinates method. The critical thickness of reflected slabs with various degrees of anisotropic scattering has been determined. In general, the agreement with the new calculation results is good. However, for thin slabs the deviations may be larger than one per cent, so some of these cases were studied in more detail.

We have analytically solved the boundary problem for a solitary CANDLE-Burn Up wave in a slab reactor configuration using the affine approach which can be considered as the analogon to the Buckling-Concept for spatially/stationary critical zones in linear reactor physics. It sheds some light on the problems, we are confronted with, when we have to deal with nonlinear mathematics in this new field of reactor physics.