[en] The Tecnatom Hydraulic Loop is a dynamic training platform. It has been designed with the purpose of improving the work in teams. With this system, the student can obtain a full scope vision of a system. The hydraulic Loop is a part of the Tecnatom Maintenance Centre. The first objective of the hydraulic Loop is the instruction in components, process and process control using open control system. All the personal of an electric power plant can be trained in the Hydraulic Loop with specific courses. The development of a dynamic tool for tests previous to plant installations has been an additional objective of the Hydraulic Loop. The use of this platform is complementary to the use of full-scope simulators in order to debug and to analyse advanced control strategies. (Author)

[en] Enhancements of the TEMPO system for monitoring and optimisation of thermal performance of nuclear steam cycles are described. As described previously, TEMPO is based on first-principles models and work in an optimisation mode, a data reconciliation mode, or a simulation mode. For the data reconciliation (and optimisation) mode the most important enhancement is a parameter-sensitivity analysis feature. The results obtained with TEMPO during an extended test period at Swedish NPP Forsmark are included. The calculated values follow the measured ones closely, also during initial phases of coast-down. Feedwater flow is usually predicted to within 2 kg/s. The goodness of fit usually centered around 0.9, and dropped to zero in cases of known faults. The results are also analysed in terms of redundancy and model configuration, among other things. Monte Carlo simulations indicate that the analytically derived value for the goodness of fit is lower than the true value. The report also discusses briefly how to use the various calculational modes of TEMPO in fault detection, isolation and identification. (Author)

[en] This report presents an early progress of a feasibility study of a computational intelligence approach to the enhancement of the accuracy of feedwater flow measurements in the framework of an ongoing cooperation between Tecnatom s.a. in Madrid and the OECD Halden Reactor Project (HRP) in Halden. The aim of this research project is to contribute to the development and validation of a flow sensor in a nuclear power plant (NPP). The basic idea is to combine the use of applied computational intelligence approaches (noise analysis, neural networks, fuzzy systems, wavelets etc.) with existing traditional flow measurements, and in particular with cross correlation flow meter concepts. In this report, Section 2 outlines relevant aspects of thermal power calculations on electrical power plants. Section 3 reviews from the available literature possible approaches and solutions for feedwater flow measurement, including ultrasonic flow meters, cross-correlation flow meters, and 'Virtural' flow meters with artificial neural networks. Section 4 reports typical experimental measurements at the Tecnatom's facility. Section 5 presents an integration approach and preliminary experimental tests. Section 6 discusses the role of soft computing techniques in the context of feedwater flow measurements related nuclear fields, and Section 7 highlights the future research direction. (Author)

[en] The paper presents the results of the implementation of SFAT for verification of spent fuel at various types of nuclear reactors. The device has been used in those cases where the improved Cherenkov Viewing Device (ICVD) could not provide conclusive results. The paper summarises the experience gained in using SFAT for verification of spent fuel at facilities in Europe. The measurements were performed at Boiling and Pressurised Light Water Reactors of Western design, at WWER-440 and WWER-1000 reactors, and at Research Reactors of MTR type. The experience proved that SFAT can be successfully used for verification of spent fuel. In order to achieve best verification results, the SFAT must be carefully optimised taking into account the fuel assemblies' design features, storing arrangements and irradiation history. (author)

[en] Measurement of gamma and neutron emission from freshly discharged spent fuel assemblies was performed at a Boiling Water Reactor (BWR) to verify irradiated Uranium-Plutonium Mixed Oxide (MOX) assemblies and to distinguish them clearly from Low Enriched Uranium (LEU) spent fuel assemblies of different irradiation cycles. A Grand Fork Detector was used for these measurements. The measurements were performed two weeks after the assemblies were discharged from the reactor. The neutron and gamma rays ratios were used to verify irradiation history of MOX assembly after two cycles of irradiation, and to differentiate them from LEU assemblies with similar or different irradiation histories. The Fork detector was modified in order to operate in the intense gamma ray field emitted from short cooling time assemblies. 14 irradiated MOX assemblies with two cycles of irradiation, two LEU assemblies with two cycles of irradiation and one with four cycles of irradiation were measured for comparison. It was demonstrated that irradiated MOX assemblies can be successfully distinguished from irradiated LEU with a similar irradiation history. LEU irradiated assemblies with a longer irradiation history (four cycles) could also be easily distinguished from the irradiated MOX assemblies. The method can be used to re-establish the knowledge on MOX fuel inventory at LWRs when the continuity of knowledge, usually maintained through C/S measures, is lost. (author)