Volume 79, Issue 2

The operational experience (OE) feedback provides improvements in all industrial activities. Identification of the most important and valuable groups of events within accumulated experience is important in order to focus on a detailed investigation of events. The paper describes the new ranking method and compares it with three others. Methods have been described and applied to OE events utilised by nuclear power plants in France and Germany for twenty years. The results show that different ranking methods only roughly agree on which of the event groups are the most important ones. In the new ranking method the analytical hierarchy process is applied in order to assure consistent and comprehensive weighting determination for ranking indexes. The proposed method allows a transparent and flexible event groups ranking and identification of the most important OE for further more detailed investigation in order to complete the feedback.

In the previously published part 1 of the paper the uncertainty analysis of the large break loss-of-coolant accident (LBLOCA) for German PWR Konvoi was performed using a statistical method, which is based on the Wilks' theory. The evaluated output parameter is the peak cladding temperature (PCT). The primary goal of this (second) part of the paper is a ranking of the input uncertainties, according to their contributions to the PCT uncertainty in the ATHLET simulation of PWR LBLOCA, by performing a sensitivity analysis. It was shown, that the first extended set of varied parameters used in part 1 can be considerably reduced without any statistically significant influence on the uncertainty analysis results. Thus, it can be shown that the input uncertainty vector based on the LBLOCA PIRT of AREVA GmbH was complete. To minimize the number of varied parameters the statistical t-test was used and, thus, a set of uncertainty parameters with significant impact on the uncertainty of the PCT was identified. The main contribution to the uncertainty of the first cladding temperature maximum during the blowdown phase of the accident is produced by the core parameters that affect the fuel's stored energy at the beginning of the accident. However, the major contributors to the uncertainty of the second PCT maximum are the uncertainties in the code models, and first of all the uncertainties in the heat transfer coefficients for dispersed and pure steam flows.

A statistical analysis methodology based on the code scaling, applicability and uncertainty (CSAU) evaluation approach for predicting the safety margin in case of a postulated large-break loss-of-coolant accident (LBLOCA) in a pressurized-water reactor (PWR) was developed by AREVA. All expected LBLOCA phenomena are listed in the Phenomena Identification and Ranking Table (PIRT) and are prioritized according to their importance on the figure of merit, here the fuel rod peak cladding temperature (PCT). For the high-ranked phenomena parameters are identified, which allow a quantification of the analysis uncertainty. AREVA has updated the PIRT to the state of the art and extended it to the application to pressurized-water reactors with combined emergency core cooling injection of German-type PWRs. This paper describes how the uncertainty distributions, required for a statistical analysis, have been derived and presents the result of an exemplary statistical analysis for a German-type 4-loop plant compared to that of a conservative deterministic analysis.

An experimental facility called Containment Studies Facility (CSF) has been constructed at the Bhabha Atomic Research Centre (BARC) for the purpose of research and development in the area of nuclear reactor containment thermal hydraulics. The facility consists of reinforced concrete containment structural model (CM) and a Primary Heat Transport Model (PHTM) vessel. The containment model of CSF is divided into high enthalpy V1 volume (dry well) and low enthalpy V2 volume (wet well). The PHTM and associated pump and piping system is designed for simulating the Loss of Coolant Accident (LOCA) conditions within the containment model. Experiments were carried out in CSF for 30 bar and 50 bar blowdown conditions. Pressure, temperature and other transient parameters were recorded. The experimental results were compared with outputs of modeling carried out using RELAP code and in-house containment thermal hydraulic code CONTRAN.

Several numeric tools have been developed and reported in literature for simulation of two phase flow conditions. Most of the tools make use of the classical homogeneous equilibrium model of two phase flow which is reported to perform better under high pressure conditions with relatively small temperature gradients. System codes such as RELAP make use of two fluid model of two phase flow. Such system codes model the flow of two phases in one dimension in the axial direction. The provision of modeling flow in the direction perpendicular to the axial flow direction exists with the approximation of not considering the flux terms in the momentum equations in the perpendicular direction. This approach is reported to work well for situations with predominant single directional flows. To analyze certain situations with predominant flow in axial as well as radial directions a numeric tool (FLOW) has been developed. The approach employs the same momentum equation for axial as well as radial direction incorporating the flux terms in both directions. The pressure and the enthalpy of the fluid are used as independent variable in the formulation of the tool. The results are found to be in good agreement with 3D CFD simulations.

The Kiev's research reactor WWR-M has been in operation for more than 50 years. Decommissioning plan should to be prepared and approved before the beginning of the decommissioning activities. A key activity during reactor operation is implementing the pre-decommissioning complex engineering and radiation inspection. It should be done with objective to collect, arrange and analyze the data related to the engineering and radiation conditions of the reactor systems and equipment. Recently, such an inspection has been completed. The analysis of available documentation has consisted in the assessment of design, construction, technological, assembling, operation, maintenance and repair documentation for each system. The radiation survey was performed in two different ways, namely, by doing the experimental measurements and by performing calculations. The collected data are provided a comprehensive technical basis for the development of decommissioning documentation which is required for the planning and implementation of the reactor decommissioning.

CANDU type reactors have some peculiarities in initiation and progression of severe accidents. In the present paper one of the specificities – the End Fitting Failure accident – is analysed from the point of view of source term formation. The accident is initiated by a failure of the re-fueling machine. Fuel bundles are ejected in the re-fuelling machine room and fuel elements suffer a significant fragmentation by mechanical impact and by the rapid increase of the temperature. A direct transfer of the fission products occurs directly to the containment. The source term in the containment and also the source term to the environment is calculated supposing an open venting communication to the external atmosphere. The simulation is performed by using the ASTEC code in coupled calculation CPA-IODE-ISODOPE-DOSE option. The evolution of the distributions for the most important released fission products is presented for different regions and for different hosts. The most important factors of influence on the source term formation are identified and discussed.

In this study, the molten salt-heavy metal mixtures 93 – 85 % Li 20 Sn 80 + 5 % SFG-PuO 2 and 2 – 10 % UO 2 , 93 – 85 % Li 20 Sn 80 + 5 % SFG-PuO 2 and 2 – 10 % NpO 2 , 93 – 85 % Li 20 Sn 80 + 5 % SFG-PuO 2 and 2 – 10 % UCO were used as fluids. The fluids were used in the liquid first wall, blanket and shield zones of the designed hybrid reactor system. Four centimeter thick 9Cr2WVTa ferritic steel was used as the structural material. In this study, the effect of mixture components on the neutron flux was investigated in a designed fusion–fission hybrid reactor system. The neutron flux was investigated according to the mixture components, radial flux distribution and energy spectrum in the designed system. Three-dimensional analyses were performed using the most recent MCNPX-2.7.0 Monte Carlo radiation transport code and the ENDF/B-VII.0 nuclear data library.

Lignite, taken from basin in Afşin-Elbistan region, and ash samples were analyzed according to the qualitative, quantitative and radioactivity properties. An elemental analysis was made by using the Energy Dispersive X-Ray Fluorescence (EDXRF) technique. 59.5 keV photons emitted from a 241 Am source and 5.9 keV photons emitted from a 55 Fe radioactive source were used for excitation. The characteristic K X-rays of the elements were counted with a Si(Li) detector. For the same samples gross alpha, gross beta and radionuclide activities were also measured.

In this work a study of two-dimensional fixed-source neutron transport problems, in Cartesian geometry, is reported. The approach reduces the complexity of the multidimensional problem using a combination of nodal schemes and the Analytical Discrete Ordinates Method (ADO). The unknown leakage terms on the boundaries that appear from the use of the derivation of the nodal scheme are incorporated to the problem source term, such as to couple the one-dimensional integrated solutions, made explicit in terms of the x and y spatial variables. The formulation leads to a considerable reduction of the order of the associated eigenvalue problems when combined with the usual symmetric quadratures, thereby providing solutions that have a higher degree of computational efficiency. Reflective-type boundary conditions are introduced to represent the domain on a simpler form than that previously considered in connection with the ADO method. Numerical results obtained with the technique are provided and compared to those present in the literature.

The diffusion lengths for one-speed neutrons are calculated by solving transport equation in one-dimensional spherical geometry. The pseudo-slab transport equation is obtained using a reasonable approach. The neutron angular flux is expanded in a series of both the Chebyshev polynomials of first and second kinds. Then analytic and numerical results for asymptotic relaxation length are obtained by applying first order approximations to the transport equation and they are compared with each other.