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02320 A study on the systematic framework to developeffective diagnosis procedures of nuclear power plants
05 Nuclear fuels (economics, policy, supplies, forecasts) classified into three release processes, and each process was well described by a simple exponential decrease with time. It was found that Hz GDC showed the superior hydrogen isotope release characteristics to the He GDC, probably because of chemical processes, such as isotope exchanges assisted by the chemical sputtering process between discharged hydrogen and hydrogen isotopes in the plasma facing carbon tiles were enhanced by the H2 glow discharge. Based on the release kinetics observed in the present work, it is estimated that it takes several days to reduce tritium (deuterium) inventory within the removal depth by the H2 GDC in JT-60U to a half by continuous H2 GDC at 573 K.
04/02314 Creep surge effect under cyclic irradiation of structural materials for fusion reactors Tsepelev, A. B. Fusion Engineering and Design, 2004, 70, (2), 79-85. Experimental investigations were performed on the radiation-creep behaviour of austenitic stainless steels, c~-Fe and Fe-Cr alloys. The effect of an abrupt increase of creep rate early after either switching on and switching off irradiation (radiation-induced creep surge) was revealed. The effect of electron irradiation on the thermal-activation parameters of creep was examined. The radiation creep surge effect is discussed in terms of a dynamic preference of dislocations under nonstationary conditions typical of the formation of new radiationequilibrium flows of radiation induced point defects.
04•02315 Development of plasma stored energy feedback control and its application to high performance discharges on JT-60U Oikawa, T. et al. Fusion Engineering andDesign, 2004, 70, (2), 175 183. The real-time feedback control of the plasma stored energy has been developed for control of the plasma MHD stability in the JT-60U tokamak. The plasma stored energy can be detected with high accuracy in real-time by a function parameterization method for various plasmas available in JT-60U, such as Ohmic plasmas, the L-mode, the H-mode, the high poloidal beta mode and the reversed shear mode over a wide range of the plasma parameters. By manipulating the neutral bean] injection power, the plasma stored energy has been successfully controlled along the preprogrammed reference waveform. Especially in the reversed shear mode, this feedback control scheme has improved the reproducibility of the formation of the internal transport barrier, and MHD activities could be suppressed keeping the normalized beta in a stable region. A DT equivalent fusion amplification gain of 0.5 was sustained for 0.8 s in a reversed shear plasma by employing this feedback control scheme.
04/02316 Minimizing the gamma radiation effects to spent fuel pool walls Tippayakul, C. et al. Annals of Nuclear Energy, 2004, 31, (5), 459~480. This paper presents the study of the gamma exposure to the spent fuel pool walls from spent fuel assemblies. The Monte Carlo code, MCNP4C, was used to model and analyse the spent fuel pool walls. In addition, the gamma source from the spent fuel assembly was modelled by the isotope generation and depletion code, ORIGEN2.2. Both PWR (TMI-I) and BWR (PB-2) spent fuel pool models were investigated. The primary objective of the study was to determine the arrangement of the spent fuel assemblies to minimize the gamma radiation to the spent fuel pool walls. Different cases of spent fuel assembly configurations in the spent fuel pool were studied. The study results show the amount of gamma radiation deposited on each section of the spent fuel pool walls. Finally, the spent fuel assembly with axial peaking factor was also studied.
04/02317 Positron annihilation Investigations of defects in copper alloys selected for nuclear fusion technology Slugen, V. et al. Fusion Engineering and Design, 2004, 70, (2), 141-153. A positron annihilation spectroscopy (PAS) based on positron lifetime measurements, using the pulsed low energy positron system (PLEPS), is used to investigate defects create by hydrogen implantation and thermal treat of copper alloys. These alloys are designated for the use in the international thermonuclear experimental reactor (ITER). The results show that the changes in the microstructure of selected copper alloys (CuCrZr, CuA125) depend strongly on the preparing technology of alloys as well as on the implantation dose experimentally simulating the neutron treatment. The full recovering of the structure after isochronal annealing in vacuum is observed in all implanted specimens at the temperature of 450°C. Using the PLEPS technique, for the first time, depth profiling of the positron lifetime spectra in the near-surface region (20-500 nm) of hydrogen implanted copper alloys was performed and compared with the TRIM calculations and transmission electron microscopy (TEM) studies. Possible annihilation channels in CnCrZr and CuAl25 materials are discussed in details together with corresponding annihilation characteristics determined theoretically and using computer simulations. The paper discusses the results of positron lifetime measurements of the irradiated and non-irradiated CuZrCr and CuA125 specimens. The most probable types of positron
328
Fuel and Energy Abstracts September 2004
trapping sites were identified. Finally, the results are discussed in terms of microstructural changes of the studied materials upon irradiation and subsequent heat treatment.
04/02318 Temperature dynamics model of a pulsed fission material assembly Bondarchenko, E. A. et al. Annals of Nuclear Energy, 2004, 31, (6), 601 617. Heat exchange process differential equations are considered for a subcritical fuel assembly with an injector. The equations are obtained by means of the use of the Hermite polynomial. The model is created for modelling of temperature transitional processes. The parameters and dynamics are estimated for hypothetical fuel assembly consisting of real mountings: the powerful proton accelerator and the reactor IBR-2 core at its sub-critical state.
04/02319 Total cross-sections of crystalline uranium for thermal neutrons Habib, N. Annals of Nuclear Energy, 2004, 31, (6), 697-709. The total neutron cross-sections of c~-uranium (orthorhombic structure) have been calculated from 0.1 meV to 10 eV, using a generalized formula. Computer codes P O R U and SORU were developed to calculate the total neutron cross-section and neutron transmission through poly and mono-crystalline uranium respectively. The obtained agreement between the calculated total cross-sections of poly-crystalline uranium in the energy range from 4 meV to 1 eV and the available experimental values justifies the applicability of the used formula and the computer code. A feasibility study on using depleted uranium single crystal as a thermal neutron filter is also given.
Economics, policy, supplies, forecasts 04/02320 A study on the systematic framework to develop effective diagnosis procedures of nuclear power plants Park, J. and Jung, W. Reliability Engineering & System Safety, 2004, 84, (3), 319-335. In complex systems such as the nuclear and chemical industry, the importance of a diagnosis procedure has been well recognized, since identifying the nature of an on-going event should be preceded to determine successful countermeasures or remedial actions. Unfortunately, a systematic framework that can suggest a unified and consistent process for constructing useful diagnosis procedures seems to be scant. In this paper, the systematic framework that can provide a sound way in constructing a diagnosis procedure is suggested based on two kinds of technical bases, such as the decision-making strategies of h u m a n and the test sequencing technique. To demonstrate the appropriateness of suggested framework, the diagnosis procedure of the reference nuclear power plant is reformed based on it. Subjective ratings are conducted to compare reformed procedure with the original one, and results support that operators' performance in an event diagnosis could be improved. Thus, although well designed experiments are needed to draw a reliable conclusion, it is expected that suggested framework could be applied to provide a consistent process in constructing useful diagnosis procedures.
04/02321 Concept of a small-scale accelerator driven system for nuclear waste transmutation, part 1, Target optimization Brolly, A. and V6rtes, P. Annals' of Nuclear Energy, 2004, 31, (6), 585600. Recently, the transmutation of nuclear waste seems to be a solution of the most serious problem of nuclear energy. Since the past decade numerous conceptions of critical and sub-critical reactors dedicated to waste incineration were published. Mainly, because of safety problems, the proton accelerator driven sub-critical systems seem to be preferred. An alternative concept is based on electron-neutron (e-n) converter neutron source. In this paper after a short presentation of molten salt e - n converter (MSENC) transmuter concept, the target optimization for this device is given in details.
04/02322
On 'immortal' nuclear power plants
Weinberg, A. M. Technology in Society, 2004, 26~ (2-3), 447-453. US nuclear reactors have demonstrated that they can continue to operate well beyond their thirty or forty year operating licenses which, at the time they were issued, "~ere based on criteria for large fossil fuel power plants. Accordingly, a longevity criterion for 21st century reactors is proposed here, designed to last 100 years or more. Cheap electricity from a fully amortized long-life reactor is a gift from the generations that paid for the reactor to future generations that benefit from it, and they compensate (to a degree) for the burden of geologically sequestered wastes.
04/02314 Creep surge effect under cyclic irradiation of structural materials for fusion reactors Tsepelev, A. B. Fusion Engineering and Design, 2004, 70, (2), 79-85. Experimental investigations were performed on the radiation-creep behaviour of austenitic stainless steels, c~-Fe and Fe-Cr alloys. The effect of an abrupt increase of creep rate early after either switching on and switching off irradiation (radiation-induced creep surge) was revealed. The effect of electron irradiation on the thermal-activation parameters of creep was examined. The radiation creep surge effect is discussed in terms of a dynamic preference of dislocations under nonstationary conditions typical of the formation of new radiationequilibrium flows of radiation induced point defects.
04•02315 Development of plasma stored energy feedback control and its application to high performance discharges on JT-60U Oikawa, T. et al. Fusion Engineering andDesign, 2004, 70, (2), 175 183. The real-time feedback control of the plasma stored energy has been developed for control of the plasma MHD stability in the JT-60U tokamak. The plasma stored energy can be detected with high accuracy in real-time by a function parameterization method for various plasmas available in JT-60U, such as Ohmic plasmas, the L-mode, the H-mode, the high poloidal beta mode and the reversed shear mode over a wide range of the plasma parameters. By manipulating the neutral bean] injection power, the plasma stored energy has been successfully controlled along the preprogrammed reference waveform. Especially in the reversed shear mode, this feedback control scheme has improved the reproducibility of the formation of the internal transport barrier, and MHD activities could be suppressed keeping the normalized beta in a stable region. A DT equivalent fusion amplification gain of 0.5 was sustained for 0.8 s in a reversed shear plasma by employing this feedback control scheme.
04/02316 Minimizing the gamma radiation effects to spent fuel pool walls Tippayakul, C. et al. Annals of Nuclear Energy, 2004, 31, (5), 459~480. This paper presents the study of the gamma exposure to the spent fuel pool walls from spent fuel assemblies. The Monte Carlo code, MCNP4C, was used to model and analyse the spent fuel pool walls. In addition, the gamma source from the spent fuel assembly was modelled by the isotope generation and depletion code, ORIGEN2.2. Both PWR (TMI-I) and BWR (PB-2) spent fuel pool models were investigated. The primary objective of the study was to determine the arrangement of the spent fuel assemblies to minimize the gamma radiation to the spent fuel pool walls. Different cases of spent fuel assembly configurations in the spent fuel pool were studied. The study results show the amount of gamma radiation deposited on each section of the spent fuel pool walls. Finally, the spent fuel assembly with axial peaking factor was also studied.
04/02317 Positron annihilation Investigations of defects in copper alloys selected for nuclear fusion technology Slugen, V. et al. Fusion Engineering and Design, 2004, 70, (2), 141-153. A positron annihilation spectroscopy (PAS) based on positron lifetime measurements, using the pulsed low energy positron system (PLEPS), is used to investigate defects create by hydrogen implantation and thermal treat of copper alloys. These alloys are designated for the use in the international thermonuclear experimental reactor (ITER). The results show that the changes in the microstructure of selected copper alloys (CuCrZr, CuA125) depend strongly on the preparing technology of alloys as well as on the implantation dose experimentally simulating the neutron treatment. The full recovering of the structure after isochronal annealing in vacuum is observed in all implanted specimens at the temperature of 450°C. Using the PLEPS technique, for the first time, depth profiling of the positron lifetime spectra in the near-surface region (20-500 nm) of hydrogen implanted copper alloys was performed and compared with the TRIM calculations and transmission electron microscopy (TEM) studies. Possible annihilation channels in CnCrZr and CuAl25 materials are discussed in details together with corresponding annihilation characteristics determined theoretically and using computer simulations. The paper discusses the results of positron lifetime measurements of the irradiated and non-irradiated CuZrCr and CuA125 specimens. The most probable types of positron
328
Fuel and Energy Abstracts September 2004
trapping sites were identified. Finally, the results are discussed in terms of microstructural changes of the studied materials upon irradiation and subsequent heat treatment.
04/02318 Temperature dynamics model of a pulsed fission material assembly Bondarchenko, E. A. et al. Annals of Nuclear Energy, 2004, 31, (6), 601 617. Heat exchange process differential equations are considered for a subcritical fuel assembly with an injector. The equations are obtained by means of the use of the Hermite polynomial. The model is created for modelling of temperature transitional processes. The parameters and dynamics are estimated for hypothetical fuel assembly consisting of real mountings: the powerful proton accelerator and the reactor IBR-2 core at its sub-critical state.
04/02319 Total cross-sections of crystalline uranium for thermal neutrons Habib, N. Annals of Nuclear Energy, 2004, 31, (6), 697-709. The total neutron cross-sections of c~-uranium (orthorhombic structure) have been calculated from 0.1 meV to 10 eV, using a generalized formula. Computer codes P O R U and SORU were developed to calculate the total neutron cross-section and neutron transmission through poly and mono-crystalline uranium respectively. The obtained agreement between the calculated total cross-sections of poly-crystalline uranium in the energy range from 4 meV to 1 eV and the available experimental values justifies the applicability of the used formula and the computer code. A feasibility study on using depleted uranium single crystal as a thermal neutron filter is also given.
Economics, policy, supplies, forecasts 04/02320 A study on the systematic framework to develop effective diagnosis procedures of nuclear power plants Park, J. and Jung, W. Reliability Engineering & System Safety, 2004, 84, (3), 319-335. In complex systems such as the nuclear and chemical industry, the importance of a diagnosis procedure has been well recognized, since identifying the nature of an on-going event should be preceded to determine successful countermeasures or remedial actions. Unfortunately, a systematic framework that can suggest a unified and consistent process for constructing useful diagnosis procedures seems to be scant. In this paper, the systematic framework that can provide a sound way in constructing a diagnosis procedure is suggested based on two kinds of technical bases, such as the decision-making strategies of h u m a n and the test sequencing technique. To demonstrate the appropriateness of suggested framework, the diagnosis procedure of the reference nuclear power plant is reformed based on it. Subjective ratings are conducted to compare reformed procedure with the original one, and results support that operators' performance in an event diagnosis could be improved. Thus, although well designed experiments are needed to draw a reliable conclusion, it is expected that suggested framework could be applied to provide a consistent process in constructing useful diagnosis procedures.
04/02321 Concept of a small-scale accelerator driven system for nuclear waste transmutation, part 1, Target optimization Brolly, A. and V6rtes, P. Annals' of Nuclear Energy, 2004, 31, (6), 585600. Recently, the transmutation of nuclear waste seems to be a solution of the most serious problem of nuclear energy. Since the past decade numerous conceptions of critical and sub-critical reactors dedicated to waste incineration were published. Mainly, because of safety problems, the proton accelerator driven sub-critical systems seem to be preferred. An alternative concept is based on electron-neutron (e-n) converter neutron source. In this paper after a short presentation of molten salt e - n converter (MSENC) transmuter concept, the target optimization for this device is given in details.
04/02322
On 'immortal' nuclear power plants
Weinberg, A. M. Technology in Society, 2004, 26~ (2-3), 447-453. US nuclear reactors have demonstrated that they can continue to operate well beyond their thirty or forty year operating licenses which, at the time they were issued, "~ere based on criteria for large fossil fuel power plants. Accordingly, a longevity criterion for 21st century reactors is proposed here, designed to last 100 years or more. Cheap electricity from a fully amortized long-life reactor is a gift from the generations that paid for the reactor to future generations that benefit from it, and they compensate (to a degree) for the burden of geologically sequestered wastes.