05 Nuclear fuels (scientific, technical)

05/01015 Identification of reactor internals' vibration modes of a Korean standard PWR using structural modeling and neutron noise analysis Park, J. el al. Progress in Nuclear Energy, 2003, 43, (1-4), 177-186. The vibration characteristics of a Korean standard PWR reactor internals have been estimated through a three-dimensional finite element analyses and verified by using the mode separated power spectral density functions obtained from the ex-core neutron noise signals. Also the natural vibration modes of the fuel assembly have been identified measuring both the ex-core and the in-core neutron noise signals, which are close to each other. As a result, the fundamental bending mode frequency of the reactor internal structure is found to be around 8 Hz and the fundamental shell mode frequency 14.5 Hz, respectively. It is also shown that the fundamental bending mode frequency of the fuel assembly is 2.3 Hz and the 2nd bending mode frequency 5.8 Hz, respectively. These results can be used for the supplements of tile Korean standard PWR's CVAP (Comprehensive Vibration Assessment Program) data.

05101016 Lesson learnt from mathematical modeling of primary circuit of VVER 440/213 reactors Pecinka, L. et al. Progress in Nuclear Energy, 2003, 43, (1-4), 167-175. For the analysis of flow induced vibrations of a VVER 440/213 unit, a generalized model has been developed which consists of the reactor as a lumped mass model with 77 degrees of freedom. Linking of the loops to the reactor pressure vessel is modelled using translational and rotational stiffnesses. Main results obtained: eigenfrequencies and mode shapes of the whole system and dynamic response of the whole system generated by the main circulating pump pressure pulsations.

05•01017 Lessons learned with vibration monitoring systems in German nuclear power plants Kolbasseff, A. and Sunder, R. Progress in Nuclear Energy, 2003, 43, (1- 4), 159-165. Compared to international standards, vibration diagnosis of German pressurized water reactors ranks high. Based on the Condition Monitoring System COMOS a lot of operational experiences could be gathered meanwhile and led to knowledge of long-term experiences. Meanwhile a successor system called 'COMOSnt ' is developed and installed in several plants. After a brief description of COMOS system features, a case study based on PWR internal vibrations is presented. In German BWRs, until now no multi-sensorial vibration monitoring comparable to PWR standards is established. Reasons for this can be related to constructional characteristics, differences in neutron-flux instrumentation and variable speed-driven operating mode of reactor recirculation pumps. However, current issues with regard to core internals were reasons to investigate BWR vibration monitoring principles, taking into account for example vibration sensors at reactor recirculation pumps, self-powered neutron detectors and aceeler- ometers from loose parts monitoring. Noise analyses of incore-neutron flux signals showing specific modes of fuel assembly vibrations are presented in detail. The correlation analyses could be verified by means of accompanying structure model calculations. At present, specific systems for recirculation pump monitoring are running in three BWRs, an overall vibration monitoring concept including core internals is tested in a reference plant.

05/01018 Materials for high performance light water reactors Ehrlich, K. et al. Journal of Nuclear Materials, 2004, 327, (2-3), 140 147. A state-of-the-art study was performed to investigate the operational conditions for in-core and out-of-core materials in a high performance light water reactor (HPLWR) and to evaluate the potential of existing structural materials for application in fuel elements, core structures and out-of-core components. In the conventional parts of a HPLWR- plant the approved materials of supercritical fossil power plants (SCFPP) can be used for given temperatures (_<600°C) and pressures (~250 bar). These are either commercial ferritic/martensitic or austenitic stainless steels. Taking the conditions of existing light water reactors (LWR) into account an assessment of potential cladding materials was made, based on existing creep-rupture data, an extensive analysis of the corrosion in conventional steam power plants and available information on material behaviour under irradiation. As a major result it is shown that for an assumed maximum temperature of 650°C not only Ni-alloys, but also austenitic stainless steels can be used as cladding materials.

05/01019 On-line neuro-expert monitoring system for Borssele Nuclear Power Plant Nabeshima, K. et al. Progress in Nuclear Energy, 2003, 43, (1-4), 397- 404. A new method for an on-line monitoring system for the nuclear power plants has been developed utilizing the neural networks and the expert system. The integration of them is expected to enhance a substantial

potential of the functionality as operators support. The recurrent neural network and the feed-forward neural network with adaptive learning are selected for the plant modelling and anomaly detection because of the high capability of modelling for dynamic behaviour. The expert system is used as a decision agent, which works on the information space of both the neural networks and the human operators. The information of other sensory signals is also fed to the expert system, together with the outputs that the neural networks generate from the measured plant signals. The expert system can treat almost all known correlation between plant status patterns and operation modes as a priori set of rules. From the off-line test at Borssele Nuclear Power Plant (PWR 480 MWe) in the Netherlands, it was shown that the neuro-expert system successfully monitored the plant status. The expert system worked satisfactorily in diagnosing the system status by using the outputs of the neural networks and a priori knowledge base from the PWR simulator. The electric power coefficient is simultaneously monitored from the measured reactive and active electric power signals.

05101020 Qualitative analysis of the SN approximations of the transport equation and combined conduction-radiation heat transfer problem in a slab Vilhena, M. T. et al. Progress in Nuclear Energy, 2003, 42, (4), 427-437. A review is presented on the recent work carried out by the research group on mathematical aspects of the existence and uniqueness of solutions of the one-dimensional steady-state transport equation as well as an analytical study of the non-linear radiative transfer equation in a slab. Dependence on control parameters about the solutions of the one-dimensional SN, approximations of the transport equation is also addressed.

05101021 Reactivity monitoring in ADS, application to the MYRRHA ADS project Baeten, P. and Abderrahim, H. A. Progress in Nuclear Energy, 2003, 43, (1-4), 413-419. Monitoring of reactivity in an ADS should be performed on-line with a simple, accurate and robust technique. Within the range of exper- imental reactor techniques, no single technique can be selected which meet these requirements. Therefore a combination of different techniques has to be chosen in a way that various off-line techniques serve as a calibration method for the on-line measurement technique. As an on-line measurement technique, the current-to-flux reactivity indicator is the most simple and robust solution. The current-to-flux reactivity indicator is based on the fact that in a sub-critical multiplying medium with a driving source the flux level is proportional to the driving source intensity, hence the beam current, and the reactivity level. However, since the proportionality constant depends on a number of core-dependent parameters and detector characteristics, this current-to-flux indicator has to be calibrated on a regular basis. For this calibration, one could benefit from the occurrence of accelerator beam trips to determine the reactivity level in dollars by means of a prompt jump analysis of the flux level change. Hence, the prompt jump reactivity indicator could act as a first calibration tool of the current-to-flux indicator. Since the prompt jump indicator still relies on the value for the effective delayed neut ron fraction to determine reactivity level, complementary techniques have to be used to obtain a more accurate determination of the reactivity. Techniques based on reactor noise methods such as the RAPJA-technique, which i s a combination of the Rossi-Alpha method and a Prompt Jump Analysis can be used in this respect. In the future the bi-speetral ratio from the Cf-source driven noise analysis could be used for this purpose:

05/01022 Searching for full power control rod patterns in a boiling water reactor using genetic algorithms Montes, J. et al. Annals of Nuclear Energy, 2004, 31, (16), 1939-1954. One of the most important questions related to both safety and economic aspects in a nuclear power reactor operation, is without any doubt its reactivity control. During normal operation of a boiling water reactor, the reactivity control of its core is strongly determined by control rods patterns efficiency. In this paper, GACRP system is proposed based on the concepts of genetic algorithms for full power control rod patterns search. This system was carried out using LVNPP transition cycle characteristics, being applied too to an equilibrium cycle. Several operation scenarios, including core water flow variation throughout the cycle and different target axial power distributions, are considered. Genetic algorithm fitness function includes reactor security parameters, such as MLHGR, MCPR, reactor kerr and axial power density.

05/01023 Stochastic techniques for the control and surveillance of a modular pebble bed reactor Kemeny, L. G. Progress in Nuclear Energy, 2003, 43, (1-4), 445-452. This paper represents the first of a series of publications describing work in progress on the research, design and testing of a control and surveillance system for a Modular Pebble Bed Reactor. The scope of

156 Fuel and Energy Abstracts May 2005