New seismic design spectra for nuclear power plants

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  • nucleus the variation of the average energy released versus the incident neutron energy is also taken into account. The calculated prompt fission neutron spectra and average prompt neutron multiplicity well represent the experimental data. proving a better predictive power of the improved Los Alamos model.

    02/01471 Investigation of the use of ceramic materials in innovative light water reactor - fuel rod concepts Lippmann, W. et al. Nuclear Engineering and Design, 2001, 205, (1-2), 13-22. The RWTH Aachen and the TU Dresden have undertaken a joint research effort, the goal of which is the development of innovative fuel rods which would constitute a high-melting reactor core. An additional sintered silicon carbide (SSiC)-encasement of the UO2 pellets within the zircaloy cladding was planned. Various designs for the construction of the absorber rods were developed in order to achieve a failure temperature in excess of 1200C. At the RWTH Aachen, a series of depleted UO2 pellets were enclosed in gastight SSiC capsules through reaction sintering. The capsules were checked for leaks, and their thermomechanical behaviour was analysed after thermal strain; the capsules were heated to 1800C (maximally 2070C) in oxidizing water vapour as well as in air. Further series of experiments were conducted in order to examine the chemical behaviour of the SSiC pellets in the presence of various reactor component materials at high temperatures. SSiC was heated to 1800C while in contact with the following substances: zircaloy, steel, corium material, UO2, Ag-ln-Cd-alloy, HfO2, Dy203, Gd203, Sm203, BN, and B4C. With the exception of steel and corium material containing steel, the substances proved relatively inert in reactions with SSiC, such that their use in combination with SSiC can be judged to be favourable.

    02/01472 New seismic design spectra for nuclear power plants McGuire, R.K. et al. Nuclear Engineering and Design, 2001, 203, (2-3), 249-257. Under a US Nuclear Regulatory Commission-sponsored project recommendations for seismic design ground motions for nuclear facilities are being developed. These recommendations will take several forms. Spectral shapes will be developed empirically and augmented as necessary by analytical models. Alternative methods of scaling the recommended shapes will be included that use a procedure that integrates over fragility curves to obtain approximately consistent risk at all sites. Site-specific soil effects will be taken into account by recommending site-specific analyses that can be used to modify rock hazard curves at a site. Also, a database of strong motion records will be archived for the project, along with recommendations on the development of artificial motions. This will aid the generation of motions for detailed soil- and structural-response studies.

    02/01473 Optimization layer by layer networks for in-core fuel management optimization computations in PWRs Jang, C.S. et al. Annals of Nuclear Energy, 2001, 28, (11), 1115-1132. The optimization layer by layer (OLL) learning algorithm is applied to prediction of assembly-wise power and burnup distribution, the critical soluble boron concentration, and the pin power peaking factor (PPPF) with core burnup in the pressurized water reactor (PWR) of the Korean Nuclear Unit (KNU) 11. It is shown that the OLL trained neural net-works can predict core depletion characteristics as accurately as the high-precision modern nodal method codes and that the OLL trained neural networks can compute core depletion characteristics about 40 times faster than the modern nodal method code. The OLL networks are then utilized for determining the optimum fuel assembly (FA) loading pattern (LP) of the equilibrium cycle KNU 11 PWR core by a simulated annealing (SA) scheme. By demonstrating that the FA LP optimization by the SA scheme can be carried out within 10 to 15 rain thanks to the speedy neutronies evaluation of the OLL networks, it is proposed that the OLL networks can make a satisfactory substitute for core evaluation codes based on modern nodal methods in in-core fuel management optimization computations where neutronics analysis for a large number of trial loading patterns has to be carried out.

    02/01474 Optimized gadolinia concepts for advanced in- core fuel management in PWRs Schlieck, M. et al. Nuclear Engineering and Design, 2001, 205, (1-2), 191-198. Utilities operating LWRs require fuel assemblies and in-core fuel management service, which ensure safe, flexible and cost-effective production of electricity. Because the reliability of the fuel has always been the most important requirement, advanced measures to minimize fuel cycle costs are receiving increasing attention in the light of the pressure on costs within the deregulated power generation markets. The role of in-core fuel management in supporting the goal to minimize fuel cycle costs consists in the development of more demanding core loading strategies, i.e. in the first place, more

    05 Nuclear fuels (scientific, technical)

    advanced low leakage loading patterns. A prerequisite for this type of loading pattern is the use of an optimized burnable absorber design. Gadolinia (Gd) as integrated burnable absorber is a very effective means for limiting the critical boron concentration and power peaking factors. Current development efforts for optimizing Gd-fuel are focused on the reduction of the inherent penalties of today's Gd-FA designs, i.e. reduced average fuel assembly (FA) enrichment and heavy metal content, as well as the residual reactivity binding. The most effective way to overcome these drawbacks is the reduction of the Gd203 concentration to values of ~2 w/o, for which, according to recent measurements of the heat conductivity of modern Gd-fuels, the reduction of the fissile content in the Gd-rods is no longer necessary. Various feasibility studies have been performed to evaluate the consequences of FA designs with low Gd-concentrations (low-Gd designs) for Siemens PWRs and non-Siemens PWRs, for which more restrictive boundary conditions with respect to critical boron concen- tration and peaking factors have to be fulfilled. These studies, as well as operation experience of reactor cycles using low Gd-FA reload designs, confirm that the in-core fuel management can handle the different Gd burnout characteristics without problems. The economical benefits of low-Gd designs compared to conventional Gd designs are comparable to those achievable by distinctly more costly and complex alternatives, like the use of enriched gadolinia.

    02/01475 Refioodingexperiments with LWR-type fuel rod simulators in the QUENCH facility Sepold, L. et al. Nuclear Engineering and Design, 2001, 204, (!-3), 205- 220. The QUENCH experiments at the Karlsruhe Research Center are carried out to investigate the hydrogen generated during reflooding of an uncovered Light Water Reactor (LWR) core. The QUENCH test bundle is made up of 21 fuel rod simulators approximately 2.5 m long. The Zircaloy-4 rod cladding is identical to that used in PWRs (Pressurized Water Reactors) with respect to material and dimensions. Pellets are made of zirconia to simulate UO2. After a transient phase with a heating rate of 0.5-1 K s -~ water of approx. 395 K is admitted from the bottom when the best bundle has reached its pre-determined temperature. Except for the flooding (quenching) phase, the QUENCH test phases are conducted in an argon/steam atmosphere at 3 g s -1 each. The results of the first two experiments, QUENCH-01 (with pre-oxidation of 300 /~m oxide layer thickness at the cladding outside surface) and QUENCH-02 (reference test without pre- oxidation), are compared in the paper. The pre-oxidized LWR fuel rod simulators of QUENCH-01 were quenched from a maximum temperature of ~1870 K. In the second bundle experiment, QUENCH- 02, quenching started at ~2.500 K. Pre-oxidation apparently.prevented a temperature escalation in the QUENCH-01 test bundle, while the QUENCH-02 test bundle experienced a temperature excursion which started in the transient phase and lasted throughout the flooding phase. The different behaviour of the two experiments is also reflected in hydrogen generation. While the bulk of H2 was produced during pre- oxidation of test QUENCH-01 (30 g), the largest amount, i.e. 170 g, of hydrogen was generated during the reflooding phase of test QUENCH- 02, at a maximum production rate of 2.5 g s- ras compared to 0.08 g s - l in test QUENCH-01. Similarities between the two experiments exist in the thermo-hydraulics during the quench phase, e.g. in the cooling behaviour, the quench temperatures, and quench velocities.

    02/01476 The concept and neutronic characteristics of a multipurpose fast reactor (MPFR) Netchaev, A. et al. Annals of Nuclear Energy, 2001, 28, (11), 1083-1099. The paper describes a concept and its neutronic characteristics of a long-life multipurpose nuclear reactor with self-sustained liquid metallic fuel, which is proposed to meet the requirements for the energy production in the future. The application of the liquid plutonium-uranium metallic alloys as a nuclear fuel demonstrates high potential to reach excellent conversion characteristics and high burn-up values with relatively small reactivity swings. Considerations about the capability of in-vessel separation of the main fission products to prolong the core lifetime without fuel reloading or shuffling are also discussed from the viewpoint of the influence on core burn-up characteristics.

    02/01477 Validation of a methodology for fuel management analysis of Laguna Verde nuclear power plant Francqois, J.L. et al. Annals of Nuclear Energy, 2001, 28, (5), 489-501. This paper shows the validation of the fuel management methodology based on the state of the art lattice physics code HELIOS and the CM- PRESTO code, for the fuel management analysis of the Laguna Verde nuclear power plant (LVNPP). The validation of these codes is performed with data from the first five operating cycles of LVNPP Unit 1. HELIOS calculations were performed for three different BWR standard type assemblies and compared with Monte Carlo, RECORD and fuel vendors' results. The CM-PRESTO results are compared against plant information, such as K-effective at hot and cold

    Fuel and Energy Abstracts May 2002 189


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