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02038 Investigations of the SAD design parameters for optimum experimental performance
05 Nuclear fuels (scientific, technical) lation. The implemented numerical method provides more reliable safety margins and maintainability, exhibiting a more uniform structural safety level than the linear elastic analysis. A simplified non-linear heat transfer model, developed for symmetrical crosssections, is used to determine the steel temperature gradient and to establish a link between the thermo and the mechanical analysis. The load resulting from pressure and temperature thermodynamic calculations, obtained for the accident scenario, are considered in the structural quasi static analysis, so that the structural response can be tracked for the entire duration of the simulated accident.
05•02037
Instrumentation and control system design
Saito, K. et al. Nuclear Engineering and Design, 2004, 233, (1 3), 125 133. The instrumentation and control system of the high temperature engineering test reactor consists of the instrumentation, control equipments and safety protection systems. There are not many differences in the instrumentation and control equipments design between the high temperature engineering test reactor (HTTR) and light water reactors except for some features. Various kinds of R&D of reactor instrumentation were performed taking into account the HTTR operational conditions, and a plant dynamic analysis was carried out considering the operational conditions of the HTTR in order to design the control system. These systems are required to have a high reliability in respect to safety. In the rise-to-power test it was confirmed that the instrumentation has a high reliability and the control system has a high stability and reasonable damped characteristics for various disturbances.
05•02038 Investigations of the SAD design parameters for optimum experimental performance Domangska, G. et al. Annals of Nuclear Energy, 2004, 31, (18), 2127 2138. A project of fast experimental accelerator-driven system called SAD Subcritical Assembly in Dubna containing MOX fuel and driven by 660 MeV proton beam is described and analysed. It is shown by design calculations that the necessity exists for certain modifications, allowing for better reliability of measurements of system time characteristics. Different solutions such as: cadmium separation of the biological concrete shield, admixtures of B203 to the concrete and certain slowing down of neutrons were analysed. Experiments on a bare spallation targets were conducted and the production of radionuclides in the lead target were measured and compared with calculations.
05•02039
Nuclear design
Fujimoto, N. et al. Nuclear Engineering and Design, 2004, 233, (1 3), 23 36. The high-temperature engineering test reactor (HTTR) has been designed for an outlet temperature of 950~'C. That is the highest temperature in the world for a block-type high-temperature gas-cooled reactor (HTGR). The functions of the reactivity control system are determined considering the operational conditions, and the reactivity balance is planned so that the design requirements are fully satisfied. Moreover, the reactivity coefficients are evaluated to confirm the safety characteristics of the reactor. The power distribution in the core was optimized by changing the uranium enrichment to maintain the fuel temperature at less than the limit (1600~'C). Deviation from the optimized distribution due to the burnup of fissile materials was avoided by flattening time-dependent changes in local reactivities. Flattening was achieved by optimizing the specifications of the burnable poisons. The original nuclear design model had to be modified based on the first critical experiments. The Monte Carlo code MVP was also used to predict criticality of the initial core. The predicted excess reactivities are now in good agreement with the experimental results.
05•02040
Overview of HTTR design features
Shiozawa, S. et al. Nuclear Engineering and Design, 2004, 233, (1 3), 11 21. The Japan Atomic Energy Research Institute (JAERI) designed and constructed the high temperature engineering test reactor (HTTR) in order to establish and upgrade the technology basis for the high temperature gas-cooled reactor (HTGR) and develop the technology for high temperature heat applications. The HTTR is a helium-cooled and graphite-moderated HTGR with a thermal power of 30 MW and a maximum reactor outlet coolant temperature of 950c'C. The first criticality was attained on November 10, 1998 and the rated power operation with the reactor outlet coolant temperature of 850~'C was achieved on December 7, 2001. Since 2002, safety demonstration tests simulating anticipated operational occurrences such as decrease of primary coolant flow-rate and reactivity insertion have been carried out to ensure safety during off-normal conditions. From 2005, various irradiation tests for fuels and materials will start. In addition, a hydrogen production test facility will be coupled with the HTTR by
2015 to produce hydrogen by nuclear. The history and future plan, major design features and R&D programs of the HTTR are summarized in this paper.
05•02041 Parameter estimation during a transient application to BWR stability Tambouratzis, T. and Antonopoulos-Domis, M. Annals o f Nuclear Energy, 2004, 31, (18), 2077 2092. The estimation of system parameters is of obvious practical interest. During transient operation, these parameters are expected to change, whereby the system is rendered time-varying and classical signal processing techniques are not applicable. A novel methodology is proposed here, which combines wavelet multi-resolution analysis and selective wavelet coefficient removal with classical signal processing techniques in order to provide short-term estimates of the system parameters of interest. The use of highly overlapping time-windows further monitors the gradual changes in system parameter values. The potential of the proposed methodology is demonstrated with numerical experiments for the problem of stability evaluation of boiling water reactors during a transient.
05•02042
Performance test of HTTR
Nakagawa, S. et al. Nuclear Engineering and Design, 2004, 233, (1 3), 291 300. The high temperature gas-cooled reactor (HTGR) is particularly attractive due to its capability of producing high temperature helium gas and due to its inherent safety characteristics. The high temperature engineering test reactor (HTTR), which is the first HTGR in Japan, was successfully constructed at the Oarai Research Establishment of the Japan Atomic Energy Research Institute. The HTTR achieved full power of 30 MW at a reactor outlet coolant temperature of about 850c'C on 7 December 2001 during the 'rise-to-power tests'. Two kinds of tests were carried out during the 'rise-to-power tests'. One is commissioning test to get operation permit by the government and another is test to confirm a performance of the reactor, heat exchanger, and control system. From the test results of the 'rise-to-power tests' up to 30 MW, the functionality of the reactor and the cooling system were confirmed, and it was also confirmed that an operation of reactor facility could be performed safely.
05•02043 Potential of thorium molten salt reactorsdetailed calculations and concept evolution with a view to large scale energy production Nuttin, A. eta[. Progress in Nuclear Energy, 2005, 46, (1), 77 99. The paper discusses the concept of a thorium molten salt reactor dedicated to future nuclear energy production. The fuel of such reactors being liquid, it can be easily reprocessed to overcome neutronic limits. In the late sixties, the MSBR project showed that breeding is possible with thorium in a thermal spectrum, provided that an efficient pyrochemical reprocessing is added. With tools developed around the Monte Carlo MCNP code, the performance of a MSBR-like reference system with 232Th / 232U fuel was re-evaluated. An important reduction of inventories and induced radiotoxicities was found at equilibrium compared to other fuel cycles, with a doubling time of about 30 years. The study then considered how to start this interesting reference system with the plutonium from PWR spent fuel. Such a transition appears slow and difficult, since it is very sensitive to the fissile quality of the plutonium used. Deployment scenarios of Z3ZTh/ 232U MSBR-like systems from the existing French PWRs demonstrate the advantage of an upstream 232U production in other reactors, allowing a direct start of the MSBR-like systems with 232U. This finally leads to the exploration of alternatives to some MSBR features, for energy production with 232Th/ 232U fuel from the start. Different options were then tested, especially in terms of core neutronics optimization and reprocessing unit adaptation.
05•02044 sheet
Purex co-processing of spent LWR fuels: flow
Zabuno~lu, O. H. and Ozdemir, L. Annals of Nuclear Energy, 2005, 32, (2), 151 162. Purex co-processing of spent LWR fuel is investigated. In purex coprocessing, uranium and plutonium in spent fuel are processed and recovered together as a single stream, while in standard purex reprocessing uranium and plutonium are obtained as separate streams. A two-step (co-decontamination and co-stripping) flow sheet for purex co-processing is devised; concentrations, recoveries and decontamination factors are calculated; and methods to co-convert uranium plutonium nitrate to mixed oxide are reviewed. A closed nuclear fuel cycle in which at no point uranium and plutonium are separated from each other is reached.
05•02045
R&D on core seismic design
Iyoku, T. et al. Nuclear Engineering and Design, 2004, 233, (1 3), 225 234.
Fuel and Energy Abstracts
September 2005
303
05•02037
Instrumentation and control system design
Saito, K. et al. Nuclear Engineering and Design, 2004, 233, (1 3), 125 133. The instrumentation and control system of the high temperature engineering test reactor consists of the instrumentation, control equipments and safety protection systems. There are not many differences in the instrumentation and control equipments design between the high temperature engineering test reactor (HTTR) and light water reactors except for some features. Various kinds of R&D of reactor instrumentation were performed taking into account the HTTR operational conditions, and a plant dynamic analysis was carried out considering the operational conditions of the HTTR in order to design the control system. These systems are required to have a high reliability in respect to safety. In the rise-to-power test it was confirmed that the instrumentation has a high reliability and the control system has a high stability and reasonable damped characteristics for various disturbances.
05•02038 Investigations of the SAD design parameters for optimum experimental performance Domangska, G. et al. Annals of Nuclear Energy, 2004, 31, (18), 2127 2138. A project of fast experimental accelerator-driven system called SAD Subcritical Assembly in Dubna containing MOX fuel and driven by 660 MeV proton beam is described and analysed. It is shown by design calculations that the necessity exists for certain modifications, allowing for better reliability of measurements of system time characteristics. Different solutions such as: cadmium separation of the biological concrete shield, admixtures of B203 to the concrete and certain slowing down of neutrons were analysed. Experiments on a bare spallation targets were conducted and the production of radionuclides in the lead target were measured and compared with calculations.
05•02039
Nuclear design
Fujimoto, N. et al. Nuclear Engineering and Design, 2004, 233, (1 3), 23 36. The high-temperature engineering test reactor (HTTR) has been designed for an outlet temperature of 950~'C. That is the highest temperature in the world for a block-type high-temperature gas-cooled reactor (HTGR). The functions of the reactivity control system are determined considering the operational conditions, and the reactivity balance is planned so that the design requirements are fully satisfied. Moreover, the reactivity coefficients are evaluated to confirm the safety characteristics of the reactor. The power distribution in the core was optimized by changing the uranium enrichment to maintain the fuel temperature at less than the limit (1600~'C). Deviation from the optimized distribution due to the burnup of fissile materials was avoided by flattening time-dependent changes in local reactivities. Flattening was achieved by optimizing the specifications of the burnable poisons. The original nuclear design model had to be modified based on the first critical experiments. The Monte Carlo code MVP was also used to predict criticality of the initial core. The predicted excess reactivities are now in good agreement with the experimental results.
05•02040
Overview of HTTR design features
Shiozawa, S. et al. Nuclear Engineering and Design, 2004, 233, (1 3), 11 21. The Japan Atomic Energy Research Institute (JAERI) designed and constructed the high temperature engineering test reactor (HTTR) in order to establish and upgrade the technology basis for the high temperature gas-cooled reactor (HTGR) and develop the technology for high temperature heat applications. The HTTR is a helium-cooled and graphite-moderated HTGR with a thermal power of 30 MW and a maximum reactor outlet coolant temperature of 950c'C. The first criticality was attained on November 10, 1998 and the rated power operation with the reactor outlet coolant temperature of 850~'C was achieved on December 7, 2001. Since 2002, safety demonstration tests simulating anticipated operational occurrences such as decrease of primary coolant flow-rate and reactivity insertion have been carried out to ensure safety during off-normal conditions. From 2005, various irradiation tests for fuels and materials will start. In addition, a hydrogen production test facility will be coupled with the HTTR by
2015 to produce hydrogen by nuclear. The history and future plan, major design features and R&D programs of the HTTR are summarized in this paper.
05•02041 Parameter estimation during a transient application to BWR stability Tambouratzis, T. and Antonopoulos-Domis, M. Annals o f Nuclear Energy, 2004, 31, (18), 2077 2092. The estimation of system parameters is of obvious practical interest. During transient operation, these parameters are expected to change, whereby the system is rendered time-varying and classical signal processing techniques are not applicable. A novel methodology is proposed here, which combines wavelet multi-resolution analysis and selective wavelet coefficient removal with classical signal processing techniques in order to provide short-term estimates of the system parameters of interest. The use of highly overlapping time-windows further monitors the gradual changes in system parameter values. The potential of the proposed methodology is demonstrated with numerical experiments for the problem of stability evaluation of boiling water reactors during a transient.
05•02042
Performance test of HTTR
Nakagawa, S. et al. Nuclear Engineering and Design, 2004, 233, (1 3), 291 300. The high temperature gas-cooled reactor (HTGR) is particularly attractive due to its capability of producing high temperature helium gas and due to its inherent safety characteristics. The high temperature engineering test reactor (HTTR), which is the first HTGR in Japan, was successfully constructed at the Oarai Research Establishment of the Japan Atomic Energy Research Institute. The HTTR achieved full power of 30 MW at a reactor outlet coolant temperature of about 850c'C on 7 December 2001 during the 'rise-to-power tests'. Two kinds of tests were carried out during the 'rise-to-power tests'. One is commissioning test to get operation permit by the government and another is test to confirm a performance of the reactor, heat exchanger, and control system. From the test results of the 'rise-to-power tests' up to 30 MW, the functionality of the reactor and the cooling system were confirmed, and it was also confirmed that an operation of reactor facility could be performed safely.
05•02043 Potential of thorium molten salt reactorsdetailed calculations and concept evolution with a view to large scale energy production Nuttin, A. eta[. Progress in Nuclear Energy, 2005, 46, (1), 77 99. The paper discusses the concept of a thorium molten salt reactor dedicated to future nuclear energy production. The fuel of such reactors being liquid, it can be easily reprocessed to overcome neutronic limits. In the late sixties, the MSBR project showed that breeding is possible with thorium in a thermal spectrum, provided that an efficient pyrochemical reprocessing is added. With tools developed around the Monte Carlo MCNP code, the performance of a MSBR-like reference system with 232Th / 232U fuel was re-evaluated. An important reduction of inventories and induced radiotoxicities was found at equilibrium compared to other fuel cycles, with a doubling time of about 30 years. The study then considered how to start this interesting reference system with the plutonium from PWR spent fuel. Such a transition appears slow and difficult, since it is very sensitive to the fissile quality of the plutonium used. Deployment scenarios of Z3ZTh/ 232U MSBR-like systems from the existing French PWRs demonstrate the advantage of an upstream 232U production in other reactors, allowing a direct start of the MSBR-like systems with 232U. This finally leads to the exploration of alternatives to some MSBR features, for energy production with 232Th/ 232U fuel from the start. Different options were then tested, especially in terms of core neutronics optimization and reprocessing unit adaptation.
05•02044 sheet
Purex co-processing of spent LWR fuels: flow
Zabuno~lu, O. H. and Ozdemir, L. Annals of Nuclear Energy, 2005, 32, (2), 151 162. Purex co-processing of spent LWR fuel is investigated. In purex coprocessing, uranium and plutonium in spent fuel are processed and recovered together as a single stream, while in standard purex reprocessing uranium and plutonium are obtained as separate streams. A two-step (co-decontamination and co-stripping) flow sheet for purex co-processing is devised; concentrations, recoveries and decontamination factors are calculated; and methods to co-convert uranium plutonium nitrate to mixed oxide are reviewed. A closed nuclear fuel cycle in which at no point uranium and plutonium are separated from each other is reached.
05•02045
R&D on core seismic design
Iyoku, T. et al. Nuclear Engineering and Design, 2004, 233, (1 3), 225 234.
Fuel and Energy Abstracts
September 2005
303