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00106 Tritium management in the first-wall materials of A-DC and TAURO blankets
05 Nuclear fuels (economics, policy, supplies, forecasts) 06/00103 Source reactivity as an extra kinetic characteristic of coupled-source subcritical systems Bokov, P. V. Annals of'Nuclear Energy, 2005, 32, (8), 795 811. The objective of the present work is to study the kinetic (i.e. in absence of in-core feedbacks) response of the subcritical system to variation of the efficiency of the external neutron source. The particular class of the subcritical system with the intensity of the external source being intrinsically dependent on neutron production in the core (so-called coupled hybrid systems) is considered. The accelerator coupled system is taken as example, although this analysis may be expanded to other types of the coupled subcritieal systems. Within the framework of a simple mathematical model of coupled system, an interpretation of the external coupled source as supplementary group of delayed neutrons is given. An auxiliary quantity - source reactivity - is introduced for convenience and a modified inhour equation for coupled systems is deduced. Analytical solution of the modified inhour equation is obtained in approximation of one group of delayed neutrons. The principal conclusion resulting from this analysis is as follows: the response of the coupled system to 'source reactivity' variation is intrinsically different from the response to core reactivity variation. Namely, there is no equivalent of prompt criticality (accompanied by drastic decrease of the reactor period) in the case of 'source reactivity' variation.
06•00104 Sticking coefficient and surface loss probability of eroded species during bombardment of carbon with deuterium Roth, J. and Hopf, C. Journal o/'Nuclear Materials, 2004, 334, (2 3), 97 103. Investigations on the deposition behaviour of eroded carbon species under controlled ion beam conditions were performed to evaluate absolute sticking coefficients. Pyrolytic carbon was eroded by D ions at energies between 45 eV and 1 keV and temperatures between 20 and 475°C. The sticking coefficient of the eroded species was obtained from a quantitative comparison of the total deposition inside a cavity with the flux of carbon entering the cavity as well as from the distribution of deposited atoms inside the cavity. High sticking coefficients of the order of 0.6 were obtained at room temperature and high ion energy, as expected from physically eroded atoms. In contrast, hydrocarbon molecules chemically emitted at either elevated temperature or low ion energy have typical sticking coefficients around 0.02. At elevated temperatures, the quantitative comparison with weight-loss measurements indicate the dominance of saturated hydrocarbons with zero sticking to the cavity walls.
06•00105 The experimental accelerator driven system (XADS) designs in the EURATOM 5th framework programme Cinotti, L. et al. Journal of Nuclear Maferials, 2004, 335, (2), 148 155. The accelerator-driven system has the potential to safely fission Minor Actinides and transmute, on industrial scale, selected long-lived fission products into isotopes of lesser concern. European leading nuclear Industrial Companies and Research Centres, within the E U R A T O M 5th Framework Programme, have studied three solutions of experimental accelerator driven systems (XADS), different for the power (80 and 50 MW) or the primary coolant (Lead Bismuth Eutectic and Gas). Two main concepts of Target Unit are envisaged: the Window Target Unit which features a thin metallic sheet as a barrier between the LBE target and the Proton Beam Pipe and the Windowless Target Unit in which the proton beam impinges directly on the free surface of the liquid LBE target. These designs are under study in order to assess and compare them on a common basis, and to outline the main R&D needs.
06•00106 Tritium management in the first-wall materials of A-DC and TAURO blankets Esteban, G. A. ef a/. Journal of Nuclear Malerials, 2004, 335, (3), 353 358. The tritium management in the first wall of two European breeding blanket options, A-DC and TAURO, has been simulated numerically to analyse the influence of the material selected: O D S - R A F M steel for the Advanced Dual-Coolant (A-DC) and SiC(SiC composite for T A U R O options. The SRIM code has been used to simulate triton implantation and define the tritium source in each kind of material as a function of the depth. The TMAP4 code was used to analyse the posterior transitory gas transport process within the material, while taking into account the tritium transport properties of each material and the temperature variation through material thickness and operating time. Both the transient evolution and the final steady-state tritium transport behaviour have been characterized. The tritium transient flux to the coolant, the recycling flux and the absorbed tritium transient inventories have been simulated. Main conclusions have been drawn about the tritium performance of each first wall.
16
Fuel and Energy Abstracts
January 2006
Economics, policy, supplies, forecasts 06•00107 A decision supported system based on the combination of fuzzy expert estimates to assess the financial risks in high-level radioactive waste projects Fiordaliso, A. and Kunsch, P. Progress in Nut~ear Energy, 2005, 46, (3 4), 374 387. This paper describes the technical aspects of the implementation of a Decision Aid System developed at ONDRAF/NIRAS/NERAS, the Belgian radioactive waste management agency. The proposed system is intended to deal with the economic and financial uncertainties inherent to Radioactive Waste Management (RWM) and especially to HighLevel-Waste (HLW) repository projects. The economic calculus of such projects are particularly difficult due to many uncertainties that still exist regarding the final design, the eventual costs and the realization schedule. The general approach mainly relies on fuzzy logic rule-based systems to infer contingency or margin factors in such uncertain contexts. The parameters of the fuzzy models are identified based on preliminary rough estimates of the financial risks. This calibration enables to build a default system that can be further refined by the introduction of experts advices. The implemented software allows to introduce new information when available with the possibility to adjust the experts credibility.
06•00108 A historical overview of probabilistic risk assessment development and its use in the nuclear power industry: a tribute to the late Professor Norman Carl Rasmussen Keller, W. and Modarres, M. Reliability Engineering & System Saf~q)', 2005, 89,(3),271 285. This paper reviews the historical development of the probabilistic risk assessment (PRA) methods and applications in the nuclear industry. A review of nuclear safety and regulatory developments in the early days of nuclear power in the United States has been presented. It is argued that due to technical difficulties for measuring and characterizing uncertainties and concerns over legal challenges, safety design and regulation of nuclear power plants has primarily relied upon conservative safety assessment methods derived based on a set of design and safety principles. Further, it is noted that the conservatism adopted in safety and design assessments has allowed the use of deterministic performance assessment methods. This approach worked successfully in the early years of nuclear power epoch as the reactor design proved to be safe enough. However, it has been observed that as the conservative approach to design and safety criteria proved arbitrary, and yielded inconsistencies in the degree to which different safety measures in nuclear power plants protect safety and public heath, the urge for a more consistent assessment of safety became apparent in the late 1960s. In the early 1970s, as a result of public and political pressures, then the US Atomic Energy Commission initiated a new look at the safety of the nuclear power plants through a comprehensive study called 'Reactor Safety Study' (WASH-1400, or 'Rasmussen Study'-after its charismatic study leader Professor Norman Rasmussen of MIT) to demonstrate safety of the nuclear power plants. Completed in October 1975, this landmark study introduced a novel probabilistic, systematic and holistic approach to the assessment of safety, which ultimately resulted in a sweeping paradigm shift in safety design and regulation of nuclear power in the United States in the turn of the Century. Technical issues of historic significance and concerns raised by the subsequent reviews of the Rasmussen Study have been discussed. Effect of major events and developments such as the Three Mile Island accident and the Nuclear Regulatory Commission and the Nuclear Industry sponsored studies on the tools, techniques and applications of the PRA that culminated in the present day riskinformed initiatives has been discussed.
06•00109
A review of nuclear fuel performance codes
Aybar, H. S. and Ortego, P. Progress in Nuclear Energy, 2005, 46, (2), 127 141. The reliable prediction of nuclear fuel rod behaviour of nuclear power reactors constitutes a basic demand for safety-based calculations, for design purposes and for fuel performance assessments. There are many nuclear fuel performance codes, some of which are available in public domain. The recent fuel design and improvement activities are focused on to extend the burnup of fuel, and the use of new materials. Thus, the most important limitation of codes is the burnup validity limit. In this study, a review of the analytical capability of some of the existing computer codes for nuclear fuel performance calculations to highlight major strengths and weaknesses is performed.
06•00110 Assessment of the seismic resistance of a ventilation stack on a reactor building Makovicka, D. and Makovi~ka, D. N~tclear Engineering and Design, 2005, 235, (13), 1325 1334.
06•00104 Sticking coefficient and surface loss probability of eroded species during bombardment of carbon with deuterium Roth, J. and Hopf, C. Journal o/'Nuclear Materials, 2004, 334, (2 3), 97 103. Investigations on the deposition behaviour of eroded carbon species under controlled ion beam conditions were performed to evaluate absolute sticking coefficients. Pyrolytic carbon was eroded by D ions at energies between 45 eV and 1 keV and temperatures between 20 and 475°C. The sticking coefficient of the eroded species was obtained from a quantitative comparison of the total deposition inside a cavity with the flux of carbon entering the cavity as well as from the distribution of deposited atoms inside the cavity. High sticking coefficients of the order of 0.6 were obtained at room temperature and high ion energy, as expected from physically eroded atoms. In contrast, hydrocarbon molecules chemically emitted at either elevated temperature or low ion energy have typical sticking coefficients around 0.02. At elevated temperatures, the quantitative comparison with weight-loss measurements indicate the dominance of saturated hydrocarbons with zero sticking to the cavity walls.
06•00105 The experimental accelerator driven system (XADS) designs in the EURATOM 5th framework programme Cinotti, L. et al. Journal of Nuclear Maferials, 2004, 335, (2), 148 155. The accelerator-driven system has the potential to safely fission Minor Actinides and transmute, on industrial scale, selected long-lived fission products into isotopes of lesser concern. European leading nuclear Industrial Companies and Research Centres, within the E U R A T O M 5th Framework Programme, have studied three solutions of experimental accelerator driven systems (XADS), different for the power (80 and 50 MW) or the primary coolant (Lead Bismuth Eutectic and Gas). Two main concepts of Target Unit are envisaged: the Window Target Unit which features a thin metallic sheet as a barrier between the LBE target and the Proton Beam Pipe and the Windowless Target Unit in which the proton beam impinges directly on the free surface of the liquid LBE target. These designs are under study in order to assess and compare them on a common basis, and to outline the main R&D needs.
06•00106 Tritium management in the first-wall materials of A-DC and TAURO blankets Esteban, G. A. ef a/. Journal of Nuclear Malerials, 2004, 335, (3), 353 358. The tritium management in the first wall of two European breeding blanket options, A-DC and TAURO, has been simulated numerically to analyse the influence of the material selected: O D S - R A F M steel for the Advanced Dual-Coolant (A-DC) and SiC(SiC composite for T A U R O options. The SRIM code has been used to simulate triton implantation and define the tritium source in each kind of material as a function of the depth. The TMAP4 code was used to analyse the posterior transitory gas transport process within the material, while taking into account the tritium transport properties of each material and the temperature variation through material thickness and operating time. Both the transient evolution and the final steady-state tritium transport behaviour have been characterized. The tritium transient flux to the coolant, the recycling flux and the absorbed tritium transient inventories have been simulated. Main conclusions have been drawn about the tritium performance of each first wall.
16
Fuel and Energy Abstracts
January 2006
Economics, policy, supplies, forecasts 06•00107 A decision supported system based on the combination of fuzzy expert estimates to assess the financial risks in high-level radioactive waste projects Fiordaliso, A. and Kunsch, P. Progress in Nut~ear Energy, 2005, 46, (3 4), 374 387. This paper describes the technical aspects of the implementation of a Decision Aid System developed at ONDRAF/NIRAS/NERAS, the Belgian radioactive waste management agency. The proposed system is intended to deal with the economic and financial uncertainties inherent to Radioactive Waste Management (RWM) and especially to HighLevel-Waste (HLW) repository projects. The economic calculus of such projects are particularly difficult due to many uncertainties that still exist regarding the final design, the eventual costs and the realization schedule. The general approach mainly relies on fuzzy logic rule-based systems to infer contingency or margin factors in such uncertain contexts. The parameters of the fuzzy models are identified based on preliminary rough estimates of the financial risks. This calibration enables to build a default system that can be further refined by the introduction of experts advices. The implemented software allows to introduce new information when available with the possibility to adjust the experts credibility.
06•00108 A historical overview of probabilistic risk assessment development and its use in the nuclear power industry: a tribute to the late Professor Norman Carl Rasmussen Keller, W. and Modarres, M. Reliability Engineering & System Saf~q)', 2005, 89,(3),271 285. This paper reviews the historical development of the probabilistic risk assessment (PRA) methods and applications in the nuclear industry. A review of nuclear safety and regulatory developments in the early days of nuclear power in the United States has been presented. It is argued that due to technical difficulties for measuring and characterizing uncertainties and concerns over legal challenges, safety design and regulation of nuclear power plants has primarily relied upon conservative safety assessment methods derived based on a set of design and safety principles. Further, it is noted that the conservatism adopted in safety and design assessments has allowed the use of deterministic performance assessment methods. This approach worked successfully in the early years of nuclear power epoch as the reactor design proved to be safe enough. However, it has been observed that as the conservative approach to design and safety criteria proved arbitrary, and yielded inconsistencies in the degree to which different safety measures in nuclear power plants protect safety and public heath, the urge for a more consistent assessment of safety became apparent in the late 1960s. In the early 1970s, as a result of public and political pressures, then the US Atomic Energy Commission initiated a new look at the safety of the nuclear power plants through a comprehensive study called 'Reactor Safety Study' (WASH-1400, or 'Rasmussen Study'-after its charismatic study leader Professor Norman Rasmussen of MIT) to demonstrate safety of the nuclear power plants. Completed in October 1975, this landmark study introduced a novel probabilistic, systematic and holistic approach to the assessment of safety, which ultimately resulted in a sweeping paradigm shift in safety design and regulation of nuclear power in the United States in the turn of the Century. Technical issues of historic significance and concerns raised by the subsequent reviews of the Rasmussen Study have been discussed. Effect of major events and developments such as the Three Mile Island accident and the Nuclear Regulatory Commission and the Nuclear Industry sponsored studies on the tools, techniques and applications of the PRA that culminated in the present day riskinformed initiatives has been discussed.
06•00109
A review of nuclear fuel performance codes
Aybar, H. S. and Ortego, P. Progress in Nuclear Energy, 2005, 46, (2), 127 141. The reliable prediction of nuclear fuel rod behaviour of nuclear power reactors constitutes a basic demand for safety-based calculations, for design purposes and for fuel performance assessments. There are many nuclear fuel performance codes, some of which are available in public domain. The recent fuel design and improvement activities are focused on to extend the burnup of fuel, and the use of new materials. Thus, the most important limitation of codes is the burnup validity limit. In this study, a review of the analytical capability of some of the existing computer codes for nuclear fuel performance calculations to highlight major strengths and weaknesses is performed.
06•00110 Assessment of the seismic resistance of a ventilation stack on a reactor building Makovicka, D. and Makovi~ka, D. N~tclear Engineering and Design, 2005, 235, (13), 1325 1334.