05 Nuclear fuels (scientific, technical)

emission effects are taken into consideration were also performed for the investigated reaction between 13.0 and 15.0 MeV neutron energy range.

04101283 Density functional study of chemical erosion mechanisms in carbon and boron-doped carbon as plasma facing material in tokamaks Ferro, Y. et al. Journal of Nuclear Materials, 2003, 321, (2-3), 294-304. Quantum calculations (density functional theory) have been developed in order to propose a new insight into the chemical erosion of pure and doped graphite considered as models for the various carbon materials cladding the first wall of magnetically controlled fusion devices. The elementary processes considered are H, C and CHn adsorption and desorption, and C, C~, CH, CCH~, extraction from the surface. The quantum results are compared to experiments through tentative new interpretations of the high-resolution electron energy loss spectroscopy vibration spectra, of the thermal desorption spectroscopy experiments and of the analytical models proposed for chemical erosion.

04/01284 Durability test on irradiated rock-like oxide fuels Kuramoto, K. et al. Journal of Nuclear Materials, 2003, 319, (1), 180- 187. For a profitable use of Pu, the Japan Atomic Energy Research Institute has been promoting researches for once-through type fuels. The strategy consists of stable rock-like oxide fuel fabrication in conven- tional fuel facilities followed by almost complete Pu burning in LWR and disposal of chemically stable spent fuel without further processing. Because leach rates of hazardous nuclides, such as TRU and /3- emitters, that have long half-lives, are very important for the evaluation of geological safety, leaching tests in deionized water at 363 K were performed with reference to the MCC-1 method. Five irradiated fuel pellets, a single phase fuel of a yttria-stabilized zirconia (YSZ) containing UO2 (U-YSZ), two fuels of U-YSZ particle dispersed in MgAI204 (SPI) or A1203 (COR) matrix, two homogeneous-blended fuels of U-YSZ and SPI or COR powders, were submitted to the tests. Stainless steel containers with Au coating and ethylene propylene diene monomer were used as leaching vessels and packing, respect- ively. The evaluated normalized leach rates of Zr, U and Pu were obviously lower than those of the other important elements and nuclides. Americium, Np and especially Y showed unexpectedly high evaluated normalized leach rates. The volatile elements, Cs and I, showed enhanced leaching within particle-dispersed type fuels because of crack formation around the particle.

04•01285 Full scale electrical insulation coating development Moreschi, L. F. et al. Fusion Engineering and Design, 2003, 69, (1-4), 303-307. A de-mountable mechanical system has been designed by ITER-IT for the attachment and the remote handling of the blanket modules. The modules are attached to the supporting structure by bolting using four flexible radial supports (cartridges) which are electrically insulated from the vacuum vessel to avoid high-value circulating currents. The hydraulic coaxial connectors also electrically connect the blanket to the module and straps are used to provide an electrical path for halo currents. Coating the contact surfaces with alumina provides the required electrical insulation, The objective of this work was to test the performance of the electrical insulation coatings specified for the ITER blanket attachment design. The testing campaign required the manufacture of mock-ups and the application of alumina coating by plasma spray technique. The flexible attachment system was mechani- cally tested with a 500 kN push/pull load for 1000 cycles. Another assembly was thermally cycled between 150 and 300°C for 1000 times in inert gas atmosphere and then mechanically cycled. The activity included mechanical testing of two pairs of wedged coated keys under an impact load of 500 kN applied 1000 times. No failures and damages occurred on the alumina coatings, which preserved good insulating and mechanical properties confirming suitability for the ITER blanket system.

04•01286 High corrosion resistant Ti-5%Ta-1,8%Nb alloy for fuel reprocessing application Kapoor, K. et al. Journal of Nuclear Materials, 2003, 322, (1), 36-44. The conventional low carbon austenitic stainless steels display good corrosion resistance behaviour in nitric acid media. However, they are sensitive to intergranular corrosion in boiling nitric acid media in the presence of oxidizing ions like hexavalent chromium, tetravalent iron and hexavalent plutonium. The Ti -5%Ta- l .8%Nb alloy was evaluated as a candidate material for such applications of nuclear fuel reprocessing. Extensive tests were carried out to establish the superior corrosion properties in comparison to the conventional stainless steel or nitric acid grade stainless steel. The fabrieability of this new alloy to various shapes like rod, sheet, wire and its weldability, which is required for making vessels, was found to be good.

04/01287 Improved performances of the fast reactor calculational system ERANOS-ERALIB1 due to improved a priori nuclear data and consideration of additional specific integral data Fort, E. et al. Annals of Nuclear Energy, 2003, 30, (18), 1879-1898. A single consistent scheme of calculational methods and nuclear data called ERANOS-ERALIB1 was produced in 1996 to calculate fast reactor neutronic parameters. It represents a significant improvement on previous schemes such as CARNAVAL-IV, PROPANE and VASCO, each of which were required in order to calculate one specific application. The nuclear data library ERALIB1 has been obtained by a consistent statistical adjustment based on 355 integral data from 71 different systems. The performance of ERALIB1 is excellent, as demonstrated during its validation for which all the k~rf SUPER-PHENIX data were reproduced to within 70 pcm. The only restriction on this satisfactory performance is related to the rather poor prediction of the sodium void reactivity effect. This was due to very bad nuclear data for 23Na, and the unsatisfactory methods used to calculate the sensitivity coefficients for the sodium void reactivity variation ApN A. To improve the performance relative to this point and to enlarge the domain of validation several actions have been undertaken: (1) a revision of the formalism and algorithms used to calculate the derivatives of ApN A tO the sodium cross section data, (2) a significant enlargement of the integral database related to this aspect of the sodium void effect. Compared to the initial database established in support of ERALIB1, several additional (18) sodium void configur- ations corresponding to voids of different volumes at different core locations have been studied. In order to broaden the range of application of the improved library, which will be called ERALIB1.A, significant effort has been devoted to additional configurations which have firstly been evaluated, and then if judged suitable, included in the adjustment process. They are related to two specifically targeted experimental programmes: (1) a study of neutron deep penetration. Several configurations of the JANUS experimental programme (shielding constructed of separate steel, iron and sodium plates) have been analysed. With this complementary information ERALIB IA becomes applicable for accurate predictions of shielding configur- ations, (2) a study of steel reflectors for a fast reactor of the SUPER- PHENIX type. The measurements performed in the MASURCA (the CIRANO experimental programme) and FCA facilities include spectral indices (F25(r)/F25(0) . . . . ) at different positions in the reflector. As a consequence of these measurements, important information has been obtained for additional 'secondary' structural material isotopes, such as 57Fe, 60Ni and 53Cr. Significant effort has also been devoted to the analysis of 29 6eff experiments. The result of this is an improvement of the uncertainty on u~(E) which guarantees a prediction of ~ff with the required accuracy (3% for critical configurations, and 5% for power reactors). The consistent statistical adjustment method by Gandini et al. (1973) has been completed. Rigorous criteria have been introduced to identify any data which are suspect and/or inconsistent in the integral database. These data may introduce additional bias in the adjusted library, and for that reason they must be discarded before adjustment.

04/01288 Inert matrix fuel performance during the first two irradiation cycles in a test reactor: comparison with modelling results Hellwig, C. and Kasemeyer, U. Journal o f Nuclear Materials, 2003, 319, (1), 87-94. In the inert matrix fuel (IMF) type investigated at the Paul Seherrer Institute, plutonium is dissolved in the yttrium stabilized zirconium oxide (YSZ), a highly radiation-resistant cubic phase with additions of erbium as burnable poison for reactivity control. A first irradiation experiment of YSZ based IMF is ongoing in the OECD Material Test Reactor in Halden together with mixed oxide fuel. The results of the first two cycles for IMF to a burnup of some 105 kW d cm -3 are presented and the modelling results in comparison with the exper- imental results are shown. A first approximation for a simple swelling model for this YSZ based IMF can be given. Possible fission gas release mechanisms are briefly discussed. The implications of the modelling results are also discussed.

04101289 Introducing the migration mode method for the solution of the space and energy dependent diffusion equation Dall'Osso, A. Annals of Nuclear Energy, 2003, 30, (18), 1829-1845. The most usual method to take into account energy dependence in whole core spatial neutronics calculation is the multi-group method. In thermal spectrum reactors as PWR, two-group theory is sufficient to describe accurately the neutron spectrum variation among spatial regions. When the spectrum hardens the precision of two-group theory decreases and more groups are necessary to keep a good accuracy. The aim of the computation method presented here is to represent with a good accuracy the spectral transitions that appear in these situations, without increasing the number of unknowns (i.e. the number of energy

176 Fuel and Energy Abstracts May 2004

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.

Fuel and Energy Abstracts May 2004 177