groups). The neutron spectrum is considered as a combination of base shapes corresponding to the different modes of migration of the neutrons in the energy dimension. The resulting energy flux distri- bution is a continuous function that fits the real one. The spatial discretization leads to matrices having the same structure of the ones obtained with multi-group theory. Then the method can be easily applied to existing codes solving the diffusion equation on the whole core in 3D. A methodological comparison between the migration mode method and the multi-group (few-group) method as well as a numerical comparison is presented.

04/01290 IteraUve computation of time-eigenvalues of the neutron transport equation Lathouwers, D. Annals of Nuclear Energy, 2003, 30, (17), 1793-1806. The Implicitly Restarted Arnoldi Method (IRAM) is applied to the computation of prompt time-eigenvalues of the neutron transport equation. Derivation of the eigenvalue problem is based on a least- squares functional combined with a spherical harmonics angular discrefization and spatial finite elements. The method is applied to a mono-energetic homogeneous slab and compared to semi-analytical results. The results are found to be accurate if the angular discretiza- tion is sufficiently refined. The scheme is also applied to a model ADS geometry in both one and three energy groups. The IRAM is found to be very efficient but the solution of fixed source problems that are part of the algorithm need to be accelerated in the multigroup case to obtain an overall efficient method.

04/01291 Markov models for evaluating risk-informed in- service inspection strategies for nuclear power plant piping systems Fleming, K. N. Reliability Engineering & System Safety, 2004, 83, (1), 27-45. As part of an EPRI sponsored research project to develop technology for risk informed in-service inspection evaluations, new methods and databases were developed to predict piping system reliability. The methods include a Markov modelling technique for predicting the influence of alternative inspection strategies on piping system reliability, and Bayes' uncertainty analysis methods for quantifying uncertainties in piping system reliability parameters. This article describes these methods and associated databases needed for their quantification with particular emphasis on the application of the Markov piping reliability model. Insights are developed regarding reliability metrics that should be used in Probabilistic Risk Assessments for estimating time dependent frequencies of loss of coolant accidents and internal flooding events. The methodology for developing estimates of all the input parameters of the piping reliability models is described including the quantitative treatment of uncertainties in risk informed applications. Examples are presented to demonstrate the practical aspects of applying the Markov model and developing the inputs needed for its quantification.

04•01292 Mobilization and mechanisms of retardation in the Oklo natural reactor zone 2 (Gabon) - inferences from U, REE, Zr, Mo and Se isotopes Bros, R. et al. Applied Geochemistry, 2003, 18~ (12), 1807-1824• Mineralogical and isotopic studies were carried out on the natural nuclear reaction zone 2 from the Oklo deposit to evaluate the mobility of several nuclear reaction products in response to the alteration of the reaction zone and to identify the mechanisms which could retard the transport of released radio-nuclides. To address these issues, in situ isotopic analyses by SHRIMP and a selective extraction procedure were performed to constrain the structural location of nuclear reaction products (exchangeable and non-exchangeable) and their association with mineral phases. The distribution patterns of U, REE, Zr and Mo isotopes reveal that substantial amounts were released from the core and migrated through the hydrothermal alteration halo over metric distances, owing to uraninite dissolution and advective transport by hydrothermal solutions during and soon after criticality. The results emphasize the mobility of Zr at Oklo, this element being often considered as ' immobile' during water-rock interactions. The main output is the demonstration of the net effects of sorpfion and co- precipitation processes. Chlorite and to a lesser extent illite were found to have adsorbed significant amounts of U, REE, Zr (and probably Th) and less sorbing elements such as Mo. Co-precipitation of secondary UO2 and P-rich coffinite within the alteration halo is also an important means of retardation. The concentration of radio-nuclides released from the reactor were probably high and they display solubility limited transport behaviour. No retention effect was found for Se in the immediate vicinity of the reactor and this element may have moved farther from its source of production. These results have interesting implications for the evaluation of long-term containment of radio- nuclides. They provide a simple illustration of the performance of a clay barrier in the uptake of radio-nuclides by sorption onto clays and

05 Nuclear fuels (scientific, technical)

re-incorporation in secondary U-minerals. This study also demon- strates the robustness of these retention processes over extremely long periods of time.

04/01293 Modelling of HTRs with Monte Carlo: from a homogeneous to an exact heterogeneous core with microparticles Plukiene, R. and Ridikas, D. Annals of Nuclear Energy, 2003, 30, (15), 1573-1585. Recently a worldwide interest in HTR technology has been experi- enced. In this context the gas turbine modular helium-cooled reactor (GT-MHR) is a potential candidate for the maximum 239pu destruction in a once-through cycle. A particular feature of GT-MHR is that its coated fuel (TRISO particles) is supposed to provide an impermeable barrier to the release of fission products and, at the same time, to resist very deep burn-up rates (more than 90% for 239pu). In this paper a Monte Carlo approach is employed to characterize the neutron fluxes and the fuel evolution inside the tiny 200-rtm diameter fuel kernels with an exact and finite geometry description. The major goal is to obtain a quantitative comparison of different geometry sets, namely homo- geneous versus single-heterogeneous and double-heterogeneous, in terms of kerr eigenvalues, the length of the fuel cycle, neutron characteristics and the evolution of fuel composition in particular. In all cases the same Monteburns ( M C N P + O R I G E N ) code system is used. The study shows that the performance of GT-MHR is considerably influenced by the way its geometry is modelled within the Monte Carlo approach. The spatial and energy shielding of the neutron flux even in such small particles cannot be neglected for

• 2 4 0 important isotopes which have high resonance cross-sections as Pu 24~Pu and 167Er. Namely, the formation of 24 PH and burn-up of 67Er are responsible for the different length of the fuel cycle, being the shortest for a double-heterogeneous geometry. On the other hand, the evolution of 239pu at a constant reactor power and comparable neutron fluence is very similar for all three geometry configurations.

04•01294 Neutronics and thermal effects of graphite foams in the performance of nuclear energy systems Difilippo, F. C. Annals of Nuclear Energy, 2004, 31, (2). 135-149. Procedures to produce light graphite foam (,-~0.5 g/cm ~) that exhibits heat conductivifies similar to full density graphite have been developed at ORNL. The consequent substantial reduction in the thermal inertia might have a significant impact in standard designs of graphite system and it could make possible new concepts. Two possible applications are explored: (a) a modular, zero burnup reactivity swing, reactor and (b) the pebble bed accelerator-driven transmutator.

04/01295 Nuclear equipment parts classification: a functional modeling approach Kim, I. S. et al. Annals of Nuclear Energy, 2003, 30, (16), 1677-1690. Nuclear power plants comprise a large number of components designated as safety-related. These safety-related components are subject to a special quality assurance programmeto assure their availability when needed, and as a result, substantiation of the product quality and suitability of these items incurs considerable costs to utilities. Furthermore, the number of vendors capable of or willing to meet nuclear-unique requirements has dwindled, forcing utilities to procure commercial grade items and dedicate them for safety-related use. Functional classification of parts, sometimes called part classifi- cation, seeks to identify the particular subset of parts associated with a safety-related component that is not relied upon for the parent component to accomplish its safety-related functions. This paper presents a functional modeling approach to part classification, which is based on Goal Tree Success Tree and Master Logic Diagram (GTST- MLD). A case study for a pressure relief valve installed in an emergency diesel generator shows that the GTST-MLD provides a functional decomposition framework within which the various func- tions of the component can be modeled in a transparent, systematic and logical fashion in connection with a hierarchical topology of the structural elements.

04/01296 Nuclear reaction analysis of helium migration in zirconia Costantini, J. M. et al. Journal of Nuclear Materials, 2003, 321, (2 3), 281-287. This paper studied helium migration in monoclinic (a-ZrO2) and cubic yttria-stabilized zirconia, 3 YSZ (ZrOz:Y) from non-destructive He

3 4 depth profiling using the resonant He(d, p) He nuclear reaction. Results have been obtained on polycrystalline a-ZrO2 ceramics and YSZ single crystals implanted with 3-MeV 3He ions at a depth around 7 ~tm then isochronously annealed in air at temperatures between 200 and IIO0°C. In a-Zr02, no change of the depth profile is found up to 800°C. In contrast, two regimes are found in YSZ: (i) below 800°C, diffusion is controlled by helium trapping at native oxygen vacancies (~10 at.%), (ii) above 800°C, helium escapes out of the profile, with almost complete outgassing at llO0°C.

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