- Email: [email protected]
00652 Thermo-hydrodynamic design and safety parameter studies of the TRIGA MARK II research reactor
05 Nuclear fuels (scientific, technical) 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 T R I M calculations and transmission electron microscopy (TEM) studies. Possible annihilation channels in CuCrZr and CuA125 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, and idenitfies the most probable types of positron trapping sites. Finally, the results are discussed in terms of microstructural changes of the studied materials upon irradiation and subsequent heat treatment.
05/00646 Power stabilization and temporal performance of a peaceful nuclear explosion reactor with a mixture of 90% flibe +10% UF4 (or ThF4) Onalan, S. Fusion Engineering and Design, 2004, 70, (3), 233-246. This work investigated the power stabilization and the temporal neutronic behaviour of a peaceful nuclear explosion reactor (PACER) with ThF4 and UF4 which produces an electrical energy of 1.2 GW ' from fusion explosions of 8.13 X 10 1 2 J to be repeated every 40 min during the operation period of 30 year. The use of ThF4 and UF4 is realized by a mixture zone consisted of flibe and fuel, instead of full flibe zone. The mixture compositions determined by volume fraction are 90% flibe +10% UF4, 90% flibe +10% ThF4 and 90% flibe +5% UF4+5% TbF4. The capacity factor of the reactor is 0.75. The cylindrical explosion chamber has a radius of 30 m and a height of 75 m. The mixture mass of 18 000 tonnes having a zone thickness of 5 m were circulated during the operation period. The mixture zone would be subdivided into jets so that the gas and the vapour bypasses the liquid as it vents and does not accelerate the liquid mixture to high velocities. The selected volume fraction is 75% void +25% mixture. The use of fuel materials in the P A C E R reactor resulted in high-energy production, sufficient tritium breeding and significant fissile fuel breeding. The averages of tritium breeding ratio (TBR) values over 30 years are between 1.1 and 1.17. Generally, the mixtures with UF4 show better performance than the mixture with ThF4. For the mixtures with ThF4, ThF4+UF4 and UF4, the energy production without the separation process reached from ~1430 MW (electric), ~t700 MW (electric) and ~2000 MW (electric) to ~1900 MW (electric), ~2150 MW (electric) and ~2320 MW (electric), respectively. The reached cumulative fissile fuel enrichments in the fuel (CFFE) in percentage are 1.8, 2.45 and 2.4%, respectively. The fuel obtained from the P A C E R could be used as a nuclear fuel only in the C A N D U and the advanced CANDU. In addition, the stabilization process is performed by means of the plutonium or uranium fuel separation from the mixture, after the energy output of the reactor reaches 1600 MW (electric), 1800 M W (electric) and 2000 MW (electric) at the operation periods of 11, 6 years and 2 months, respectively. At the end of the separation process, the separated fuel amounts are about 15, 374 and 11 tonnes, respectively. The CFFE values of the separated fuel at the end and at the start up of the separation process are 99.36 and 99.23%, 1.13 and 3.9%, and 99.99 and 99.2%, respectively. The CFFE values of the remained fuel at the end of the separation process are ~0.7, and 2.2%, ~0.7%, respectively. Consequently, in the evaluation in terms of sufficient tritium breeding, high energy production, significant fissile fuel production and the nuclear weapon hazard of the fuel, the mixture of 90% ftibe +5% ThF4+5% UF4 exhibited the highest performance.
05100647 Probabilistic analysis of accident precursors in the nuclear industry Hulsmans, M. and De Gelder, P. Journal of Hazardous Materials, 2004, 111, (1-3), 81-87. Feedback of operating experience has always been an important issue in the nuclear industry. A probabilistic safety analysis (PSA) can be used as a tool to analyse how an operational event might have developed adversely in order to obtain a quantitative assessment of the safety significance of the event. This process is called PSA-based event analysis (PSAEA). A comprehensive set of PSAEA guidelines was developed by an international project. The main characteristics of this methodology are summarized. This approach to analyse incidents can be used to meet different objectives of utilities or nuclear regulators. The paper describes the main objectives and the experiences of the Belgian nuclear regulatory organization AVN with the application of PSA-based event analysis. Some interesting aspects of the process of PSAEA are further developed and underlined. Several case studies are discussed and an overview of the obtained results is given. Finally, the interest of a broad and interactive forum on PSAEA is highlighted.
96
Fuel and Energy Abstracts March 2005
05/00648 Quantification of the transferability of reactivity effect investigations in large multiregion systems van Geemert, R. et al. Annals of Nuclear Energy, 2004, 31, (15), 17351763. A methodology has been developed for the accurate assessment of localized reactivity perturbations in a BWR lattice embedded in a larger multiplying system, based on a full-system, unperturbed calculation, and on perturbed calculations on reduced-geometry models with reflective boundary conditions (typically, reflectedassembly calculations). Reflective reduced-geometry calculations are to be followed by a fast transferability correction for making the results representative of what full system computations would have produced. In this way, one can avoid the problem of having insufficient accuracy in the results (in spite of extremely lengthy iterations), particularly for cases of small reactivity effects. Furthermore, the factorization of reactivity effect transferability, a key feature of the developed methodology, provides valuable insight into the different effects contributing to a particular integral transferability factor, along with a quantification of the relative importance of these effects for each individually considered case. The initial investment, needed for realizing the relatively low required computational effort involved in the post-correction procedure, is to obtain a limited number of adjoint equation solutions defined for the reference state at full system level. Application results are reported for the numerical analysis of fuel pin removal reactivity effects in LWR-PROTEUS. The latter is a programme of integral experiments, employing essentially a central LWR test zone driven critical by surrounding driver and buffer regions.
05/00649 Strength calculation of NPP equipment and pipelines during operation. Low- and high-cycle corrosion fatigue Filatov, V. M. and Evropin, S. V. International Journal of Pressure Vessels and Piping, 2004, 81, (8), 719 724. This paper presents empirical equations and design curves for structural steels employed in nuclear power facilities with light water reactors. These equations allow to take into account the effects of cycle asymmetry, water coolant and ductility decrease during operation. The fatigue curves cover the low-cycle and high-cycle regions (up to 1012 cycles). The equations include the mechanical characteristics of steels under static tension. The coolant effect on steel fatigue is allowed for using a model developed at the Argonne National Laboratory.
05/00650 Studies on the corrosion behavior of ceriumimplanted zirconium Peng, D. Q. et al. Journal of Nuclear Materials, 2004, 324, (1), 71-75. In order to study the influence of cerium ion implantation on the aqueous corrosion behaviour of zirconium, specimens were implanted with cerium ions with a fluence ranging from 1 × 1020 to 1 X 10-21 ions/ m 2 at about 150°C, using a M E V V A source at an extracted voltage of 40 kV. The valence and element penetration distribution of the surface layer were analysed by X-ray photoelectron spectroscopy (XPS) and auger electron spectroscopy (AES) respectively. The potentiodynamic polarization technique was employed to investigate the aqueous corrosion resistance of zirconium in a 1N H2SO4 solution. It was found that there was a remarkable improvement in the aqueous corrosion behaviour of zirconium implanted with cerium ions compared with that of the as-received zirconium. The corrosion resistance improvement of the cerium-implanted zirconium is probably due to the addition of cerium oxide dispersoid into the zirconium matrix and ozddization protection.
05/00651 The IAEA initiative on fast reactor data retrieval and knowledge preservation Stanculescu, A. International Journal of Nuclear Knowledge Management, 2004, 1, (1-2), 131-138. The paper introduces the International Atomic Energy Agency (IAEA) initiative on fast reactor data retrieval and knowledge preservation. Apart from giving the rationale for the initiative and describing it, the paper outlines the road map of the initiative and explains IAEA's role in its implementation. The status of the initiative is summarized, and a concrete work plan for the future activities proposed.
05/00652 Thermo-hydrodynamic design and safety parameter studies of the TRIGA MARK II research reactor Huda, M. Q. and Rahman, M. Annals of Nuclear Energy, 2004, 31, (10), 1101-1118. The P A R E T computer code was used to analyse important thermohydrodynamic design and safety parameters of the 3 M W T R I G A M A R K II research reactor at Atomic Energy Research Establishment (AERE), Savar, Dhaka, Bangladesh. The study involves the determination of the departure from nucleate boiling (DNB) value and studying its effect over the thermo-hydrodynamic design of the reactor. In the process the temperature profile, heat flux and pressure drop across the hottest channel of the T R I G A core were evaluated. The DNB ratio (DNBR), which is defined as the ratio of the critical heat
05 Nuclear fuels (economics, policy, supplies, forecasts) flux to the heat flux achieved in the core, was computed by means of a suitable correlation as defined in P A R E T code. Over the length 0.381 m of the hottest channel the DNBR varies, starting from 3.8951 to 5.4031, with a minimum of 2.7851. The peak heat flux occurs at the axial centre of the fuel elements; therefore the DNBR is minimum at this location. The reactor core should be designed so as to prevent the DNBR from dropping below a chosen value under a high heat flux transient condition for the most adverse set of mechanical and coolant conditions. The loss-of-flow accident (LOFA) scenario of the reactor has also been studied to ensure that the existing design and procedures are adequate to assure that the consequences from this anticipated occurrence does not lead to a significant accident. The loss-of-flow transient after a trip time of 4.08 s at 85% of loss of normal flow for the T R I G A core shows a peak temperature of 709.22°C in the fuel centreline and 131.94°C in the clad and 46.63°C in the coolant exit of the hottest channel. The transient was terminated at 15% of nominal flow after approximately 48.0 s and the time at which the reversal of coolant flow starts is approximately 67.0 s.
05/00653 Training of the position controller in SST-1 using TSC simulations Bandyopadhyay, I. et al. Fusion Engineering and Design, 2004, 70, (3), 209-220. For sustenance of the long duration discharges in the Steady State Tokamak (SST-1), it would be necessary to accurately measure and control the plasma shape, radial and vertical positions. While the plasma shape and position would be controlled by a set of superconducting coils placed outside the vessel and copper feedback control coils placed inside the vessel, their diagnostics would be carried out by an array of magnetic probes placed inside the vessel, just outside the first wall. The main objective of this paper is to demonstrate the reconstruction of plasma position and shape parameters from magnetic probe measurements in steady-state operation. Function parametrization method would be employed to infer the plasma position and shape parameters from the probe signals, which needs large volume of prior database for the probe signals for various plasma positions and shape parameters. This has been generated through SST-1 discharge simulations using the Tokamak Simulation Code (TSC), which generally has excellent agreements with experimental data. The discharge simulations using TSC over a wide range of SST-1 plasma parameters and use of the corresponding database in fitting of plasma position and shape parameters to probe signals are presented. It is also shown that the subsequent reconstruction of plasma position and shape from test data also generated from TSC simulations is well within the error limits.
05/00654 Transmutation of minor actinides discharged from LMFBR spent fuel in a high power density fusion reactor 12lbeyli, M. Energy Conversion and Management, 2004, 45, (20), 32193238. Significant amounts of nuclear wastes consisting of plutonium, minor actinides and long lived fission products are produced during the operation of commercial nuclear power plants. Therefore, the destruction of these wastes is very important with respect to public health, environment and also the future of nuclear energy. In this study, transmutation of minor actinides (MAs) discharged from LMFBR spent fuel in a high power density fusion reactor has been investigated under a neutron wall load of 10 MW/m z for an operation period of 10 years. Also, the effect of MA percentage on the transmutation has been examined. The fuel zone, containing MAs as spheres cladded with W-5Re, has been located behind the first wall to utilize the high neutron flux for transmutation effectively. Helium at 40 arm has been used as an energy carrier. At the end of the operation period, the total burning and transmutation are greater than the total buildups in all investigated cases, and very high burnups (420-470 GWd/tHM) are reached, depending on the MA content. The total transmutation rate values are 906 and 979 kg/GWth year at startup and decrease to 140 and 178 kg/GWth year at the end of the operation for fuel with 10% and 20% MA, respectively. Over an operation period of 10 years, the effective half lives decrease from 2.38, 2.21 and 3.08 years to 1.95, 1.80 and 2.59 years for 237 Np, 241 Am and 243 Am, respectively. Total atomic densities decrease exponentially during the operation period. The reductions in the total atomic densities with respect to the initial ones are 79%, 81%, 82%, 83%, 85% and 86% for 10%, 12%, 14%, 16%, 18% and 20% MAs, respectively.
05100655 Utilisation of research and training reactors in the study programme of students at the Slovak University of Technology Slugen, V. et al. International Journal of Nuclear Knowledge Management, 2004, 1, (1-2), 68-77. Preparing operating staff for the nuclear industry is and also will be one of the most serious education processes, mainly in the CentralEuropean countries where about 40-50% of the electricity is produced
in nuclear power plants. In the Central-European region exists a very extensive and also effective international collaboration in nuclear industry and education. Similarly, the level education in universities and technical high schools of this area is also good. Slovak university of technology Bratislava has established contacts with many universities abroad for utilization of research and training reactors.
05/00656 Validation of coupled neutronic/thermal-hydraulic code RELAP5-3D for RBMK-1500 reactor analysis application Uspuras, E. et al. Annals of Nuclear Energy, 2004, 31, (15), 1667-1708. This paper summarizes RELAP5-3D code validation activities carried out at the Lithuanian Energy Institute, which was performed through the modelling of RBMK-1500 specific transients taking place at Ignalina NPP. A best estimate RELAP5-3D model of the INPP RBMK-1500 reactor has been developed and validated against real plant data, as well as with the calculation results obtained using the Russian STEPAN/KOBRA code. The obtained calculation results demonstrate reasonable agreement with Ignalina NPP measured data. Behaviours of the separate MCC thermal-hydraulic parameters, as well as physical processes are predicted reasonably well to the real processes, occurring in the primary circuit of RBMK-1500 reactor. The calculated reactivity and the total reactor core power behaviour in time are also in reasonable agreement with the measured plant data, which demonstrates the correct modelling of the neutronic processes taking place in RBMK-1500 reactor core. The performed validation of RELAP5-3D model of Ignalina NPP RBMK-1500 reactor allowed to improve the model, which in the future would be used for the safety substantiation calculations of RBMK-1500 reactors. Future activities are discussed.
Economics, policy, supplies, forecasts 05•00657 Achievements of the IAEA technical working group on life management of nuclear power plants (TWGLMNPP) under the chairmanship of Acad. Myrddin Davies Kang, K.-S. and Tipping, P. International Journal of Pressure Vessels and Piping, 2004, 81, (8), 673-676. This meeting, organized by CRISM-PROMETEY in St Petersburg, Russia, is held to honour the memory of Academician Myrddin Davies, who passed away due to a tragic road accident on 11 March 2003 in Stretton, England. Academician Myrddin Davies started technical collaboration with the IAEA in the early 1980s, and in 1985 became chairman of the International Working Group on Reliability of Reactor Pressure Components (IWG-RRPC). Under his chairmanship this grew to become the Technical Working Group on Life Management of Nuclear Power Plants (TWG-LMNPP) covering broader issues and with world wide collaboration. An insight to the creation, working methods and achievements of the TWG-LMNPP is given in this paper. Acad. Myrddin Davies was a competent chairman at many specialist meetings, major conferences hosted by IAEA, other European organizations and Nuclear Engineering International activities. The direction given to the TWG-LMNPP by Acad. Myrddin Davies is shown to have made a significant contribution to the safe use of nuclear energy. Major contributions to nuclear technology of the TWGLMNPP, during the Chairmanship of Myrddin Davies, are thus cited.
05•00658 Methods and practices used in incident analysis in the Finnish nuclear power industry Suksi, S. Journal of Hazardous Materials, 2004, 111, (1-3), 73-79. According to the Finnish Nuclear Energy Act it is licensee's responsibility to ensure safe use of nuclear energy. Radiation and Nuclear Safety Authority (STUK) is the regulatory body responsible for the state supervision of the safe use of nuclear power in Finland. One essential prerequisite for the safe and reliable operation of nuclear power plants is that lessons are learned from the operational experience. It is utility's prime responsibility to assess the operational events and implement appropriate corrective actions. STUK controls licensees' operational experience feedback arrangements and implementation as part of its inspection activities. In addition to this in Finland, the regulatory body performs its own assessment of the operational experience. Review and investigation of operational events is a part of the regulatory oversight of operational safety. Review of operational events is clone by STUK basically at three different levels. First step is to perform a general review of all operational events, transients and reactor scram reports, which the licensees submit for information to STUK. The second level activities are related to the clarification of events at site and entering of events' specific data into the event register database of STUK. This is done for events that meet the set criteria for the operator to submit a special report to STUK for approval. Safety significance of operational events is determined using
Fuel and Energy Abstracts March 2005 97
05/00646 Power stabilization and temporal performance of a peaceful nuclear explosion reactor with a mixture of 90% flibe +10% UF4 (or ThF4) Onalan, S. Fusion Engineering and Design, 2004, 70, (3), 233-246. This work investigated the power stabilization and the temporal neutronic behaviour of a peaceful nuclear explosion reactor (PACER) with ThF4 and UF4 which produces an electrical energy of 1.2 GW ' from fusion explosions of 8.13 X 10 1 2 J to be repeated every 40 min during the operation period of 30 year. The use of ThF4 and UF4 is realized by a mixture zone consisted of flibe and fuel, instead of full flibe zone. The mixture compositions determined by volume fraction are 90% flibe +10% UF4, 90% flibe +10% ThF4 and 90% flibe +5% UF4+5% TbF4. The capacity factor of the reactor is 0.75. The cylindrical explosion chamber has a radius of 30 m and a height of 75 m. The mixture mass of 18 000 tonnes having a zone thickness of 5 m were circulated during the operation period. The mixture zone would be subdivided into jets so that the gas and the vapour bypasses the liquid as it vents and does not accelerate the liquid mixture to high velocities. The selected volume fraction is 75% void +25% mixture. The use of fuel materials in the P A C E R reactor resulted in high-energy production, sufficient tritium breeding and significant fissile fuel breeding. The averages of tritium breeding ratio (TBR) values over 30 years are between 1.1 and 1.17. Generally, the mixtures with UF4 show better performance than the mixture with ThF4. For the mixtures with ThF4, ThF4+UF4 and UF4, the energy production without the separation process reached from ~1430 MW (electric), ~t700 MW (electric) and ~2000 MW (electric) to ~1900 MW (electric), ~2150 MW (electric) and ~2320 MW (electric), respectively. The reached cumulative fissile fuel enrichments in the fuel (CFFE) in percentage are 1.8, 2.45 and 2.4%, respectively. The fuel obtained from the P A C E R could be used as a nuclear fuel only in the C A N D U and the advanced CANDU. In addition, the stabilization process is performed by means of the plutonium or uranium fuel separation from the mixture, after the energy output of the reactor reaches 1600 MW (electric), 1800 M W (electric) and 2000 MW (electric) at the operation periods of 11, 6 years and 2 months, respectively. At the end of the separation process, the separated fuel amounts are about 15, 374 and 11 tonnes, respectively. The CFFE values of the separated fuel at the end and at the start up of the separation process are 99.36 and 99.23%, 1.13 and 3.9%, and 99.99 and 99.2%, respectively. The CFFE values of the remained fuel at the end of the separation process are ~0.7, and 2.2%, ~0.7%, respectively. Consequently, in the evaluation in terms of sufficient tritium breeding, high energy production, significant fissile fuel production and the nuclear weapon hazard of the fuel, the mixture of 90% ftibe +5% ThF4+5% UF4 exhibited the highest performance.
05100647 Probabilistic analysis of accident precursors in the nuclear industry Hulsmans, M. and De Gelder, P. Journal of Hazardous Materials, 2004, 111, (1-3), 81-87. Feedback of operating experience has always been an important issue in the nuclear industry. A probabilistic safety analysis (PSA) can be used as a tool to analyse how an operational event might have developed adversely in order to obtain a quantitative assessment of the safety significance of the event. This process is called PSA-based event analysis (PSAEA). A comprehensive set of PSAEA guidelines was developed by an international project. The main characteristics of this methodology are summarized. This approach to analyse incidents can be used to meet different objectives of utilities or nuclear regulators. The paper describes the main objectives and the experiences of the Belgian nuclear regulatory organization AVN with the application of PSA-based event analysis. Some interesting aspects of the process of PSAEA are further developed and underlined. Several case studies are discussed and an overview of the obtained results is given. Finally, the interest of a broad and interactive forum on PSAEA is highlighted.
96
Fuel and Energy Abstracts March 2005
05/00648 Quantification of the transferability of reactivity effect investigations in large multiregion systems van Geemert, R. et al. Annals of Nuclear Energy, 2004, 31, (15), 17351763. A methodology has been developed for the accurate assessment of localized reactivity perturbations in a BWR lattice embedded in a larger multiplying system, based on a full-system, unperturbed calculation, and on perturbed calculations on reduced-geometry models with reflective boundary conditions (typically, reflectedassembly calculations). Reflective reduced-geometry calculations are to be followed by a fast transferability correction for making the results representative of what full system computations would have produced. In this way, one can avoid the problem of having insufficient accuracy in the results (in spite of extremely lengthy iterations), particularly for cases of small reactivity effects. Furthermore, the factorization of reactivity effect transferability, a key feature of the developed methodology, provides valuable insight into the different effects contributing to a particular integral transferability factor, along with a quantification of the relative importance of these effects for each individually considered case. The initial investment, needed for realizing the relatively low required computational effort involved in the post-correction procedure, is to obtain a limited number of adjoint equation solutions defined for the reference state at full system level. Application results are reported for the numerical analysis of fuel pin removal reactivity effects in LWR-PROTEUS. The latter is a programme of integral experiments, employing essentially a central LWR test zone driven critical by surrounding driver and buffer regions.
05/00649 Strength calculation of NPP equipment and pipelines during operation. Low- and high-cycle corrosion fatigue Filatov, V. M. and Evropin, S. V. International Journal of Pressure Vessels and Piping, 2004, 81, (8), 719 724. This paper presents empirical equations and design curves for structural steels employed in nuclear power facilities with light water reactors. These equations allow to take into account the effects of cycle asymmetry, water coolant and ductility decrease during operation. The fatigue curves cover the low-cycle and high-cycle regions (up to 1012 cycles). The equations include the mechanical characteristics of steels under static tension. The coolant effect on steel fatigue is allowed for using a model developed at the Argonne National Laboratory.
05/00650 Studies on the corrosion behavior of ceriumimplanted zirconium Peng, D. Q. et al. Journal of Nuclear Materials, 2004, 324, (1), 71-75. In order to study the influence of cerium ion implantation on the aqueous corrosion behaviour of zirconium, specimens were implanted with cerium ions with a fluence ranging from 1 × 1020 to 1 X 10-21 ions/ m 2 at about 150°C, using a M E V V A source at an extracted voltage of 40 kV. The valence and element penetration distribution of the surface layer were analysed by X-ray photoelectron spectroscopy (XPS) and auger electron spectroscopy (AES) respectively. The potentiodynamic polarization technique was employed to investigate the aqueous corrosion resistance of zirconium in a 1N H2SO4 solution. It was found that there was a remarkable improvement in the aqueous corrosion behaviour of zirconium implanted with cerium ions compared with that of the as-received zirconium. The corrosion resistance improvement of the cerium-implanted zirconium is probably due to the addition of cerium oxide dispersoid into the zirconium matrix and ozddization protection.
05/00651 The IAEA initiative on fast reactor data retrieval and knowledge preservation Stanculescu, A. International Journal of Nuclear Knowledge Management, 2004, 1, (1-2), 131-138. The paper introduces the International Atomic Energy Agency (IAEA) initiative on fast reactor data retrieval and knowledge preservation. Apart from giving the rationale for the initiative and describing it, the paper outlines the road map of the initiative and explains IAEA's role in its implementation. The status of the initiative is summarized, and a concrete work plan for the future activities proposed.
05/00652 Thermo-hydrodynamic design and safety parameter studies of the TRIGA MARK II research reactor Huda, M. Q. and Rahman, M. Annals of Nuclear Energy, 2004, 31, (10), 1101-1118. The P A R E T computer code was used to analyse important thermohydrodynamic design and safety parameters of the 3 M W T R I G A M A R K II research reactor at Atomic Energy Research Establishment (AERE), Savar, Dhaka, Bangladesh. The study involves the determination of the departure from nucleate boiling (DNB) value and studying its effect over the thermo-hydrodynamic design of the reactor. In the process the temperature profile, heat flux and pressure drop across the hottest channel of the T R I G A core were evaluated. The DNB ratio (DNBR), which is defined as the ratio of the critical heat
05 Nuclear fuels (economics, policy, supplies, forecasts) flux to the heat flux achieved in the core, was computed by means of a suitable correlation as defined in P A R E T code. Over the length 0.381 m of the hottest channel the DNBR varies, starting from 3.8951 to 5.4031, with a minimum of 2.7851. The peak heat flux occurs at the axial centre of the fuel elements; therefore the DNBR is minimum at this location. The reactor core should be designed so as to prevent the DNBR from dropping below a chosen value under a high heat flux transient condition for the most adverse set of mechanical and coolant conditions. The loss-of-flow accident (LOFA) scenario of the reactor has also been studied to ensure that the existing design and procedures are adequate to assure that the consequences from this anticipated occurrence does not lead to a significant accident. The loss-of-flow transient after a trip time of 4.08 s at 85% of loss of normal flow for the T R I G A core shows a peak temperature of 709.22°C in the fuel centreline and 131.94°C in the clad and 46.63°C in the coolant exit of the hottest channel. The transient was terminated at 15% of nominal flow after approximately 48.0 s and the time at which the reversal of coolant flow starts is approximately 67.0 s.
05/00653 Training of the position controller in SST-1 using TSC simulations Bandyopadhyay, I. et al. Fusion Engineering and Design, 2004, 70, (3), 209-220. For sustenance of the long duration discharges in the Steady State Tokamak (SST-1), it would be necessary to accurately measure and control the plasma shape, radial and vertical positions. While the plasma shape and position would be controlled by a set of superconducting coils placed outside the vessel and copper feedback control coils placed inside the vessel, their diagnostics would be carried out by an array of magnetic probes placed inside the vessel, just outside the first wall. The main objective of this paper is to demonstrate the reconstruction of plasma position and shape parameters from magnetic probe measurements in steady-state operation. Function parametrization method would be employed to infer the plasma position and shape parameters from the probe signals, which needs large volume of prior database for the probe signals for various plasma positions and shape parameters. This has been generated through SST-1 discharge simulations using the Tokamak Simulation Code (TSC), which generally has excellent agreements with experimental data. The discharge simulations using TSC over a wide range of SST-1 plasma parameters and use of the corresponding database in fitting of plasma position and shape parameters to probe signals are presented. It is also shown that the subsequent reconstruction of plasma position and shape from test data also generated from TSC simulations is well within the error limits.
05/00654 Transmutation of minor actinides discharged from LMFBR spent fuel in a high power density fusion reactor 12lbeyli, M. Energy Conversion and Management, 2004, 45, (20), 32193238. Significant amounts of nuclear wastes consisting of plutonium, minor actinides and long lived fission products are produced during the operation of commercial nuclear power plants. Therefore, the destruction of these wastes is very important with respect to public health, environment and also the future of nuclear energy. In this study, transmutation of minor actinides (MAs) discharged from LMFBR spent fuel in a high power density fusion reactor has been investigated under a neutron wall load of 10 MW/m z for an operation period of 10 years. Also, the effect of MA percentage on the transmutation has been examined. The fuel zone, containing MAs as spheres cladded with W-5Re, has been located behind the first wall to utilize the high neutron flux for transmutation effectively. Helium at 40 arm has been used as an energy carrier. At the end of the operation period, the total burning and transmutation are greater than the total buildups in all investigated cases, and very high burnups (420-470 GWd/tHM) are reached, depending on the MA content. The total transmutation rate values are 906 and 979 kg/GWth year at startup and decrease to 140 and 178 kg/GWth year at the end of the operation for fuel with 10% and 20% MA, respectively. Over an operation period of 10 years, the effective half lives decrease from 2.38, 2.21 and 3.08 years to 1.95, 1.80 and 2.59 years for 237 Np, 241 Am and 243 Am, respectively. Total atomic densities decrease exponentially during the operation period. The reductions in the total atomic densities with respect to the initial ones are 79%, 81%, 82%, 83%, 85% and 86% for 10%, 12%, 14%, 16%, 18% and 20% MAs, respectively.
05100655 Utilisation of research and training reactors in the study programme of students at the Slovak University of Technology Slugen, V. et al. International Journal of Nuclear Knowledge Management, 2004, 1, (1-2), 68-77. Preparing operating staff for the nuclear industry is and also will be one of the most serious education processes, mainly in the CentralEuropean countries where about 40-50% of the electricity is produced
in nuclear power plants. In the Central-European region exists a very extensive and also effective international collaboration in nuclear industry and education. Similarly, the level education in universities and technical high schools of this area is also good. Slovak university of technology Bratislava has established contacts with many universities abroad for utilization of research and training reactors.
05/00656 Validation of coupled neutronic/thermal-hydraulic code RELAP5-3D for RBMK-1500 reactor analysis application Uspuras, E. et al. Annals of Nuclear Energy, 2004, 31, (15), 1667-1708. This paper summarizes RELAP5-3D code validation activities carried out at the Lithuanian Energy Institute, which was performed through the modelling of RBMK-1500 specific transients taking place at Ignalina NPP. A best estimate RELAP5-3D model of the INPP RBMK-1500 reactor has been developed and validated against real plant data, as well as with the calculation results obtained using the Russian STEPAN/KOBRA code. The obtained calculation results demonstrate reasonable agreement with Ignalina NPP measured data. Behaviours of the separate MCC thermal-hydraulic parameters, as well as physical processes are predicted reasonably well to the real processes, occurring in the primary circuit of RBMK-1500 reactor. The calculated reactivity and the total reactor core power behaviour in time are also in reasonable agreement with the measured plant data, which demonstrates the correct modelling of the neutronic processes taking place in RBMK-1500 reactor core. The performed validation of RELAP5-3D model of Ignalina NPP RBMK-1500 reactor allowed to improve the model, which in the future would be used for the safety substantiation calculations of RBMK-1500 reactors. Future activities are discussed.
Economics, policy, supplies, forecasts 05•00657 Achievements of the IAEA technical working group on life management of nuclear power plants (TWGLMNPP) under the chairmanship of Acad. Myrddin Davies Kang, K.-S. and Tipping, P. International Journal of Pressure Vessels and Piping, 2004, 81, (8), 673-676. This meeting, organized by CRISM-PROMETEY in St Petersburg, Russia, is held to honour the memory of Academician Myrddin Davies, who passed away due to a tragic road accident on 11 March 2003 in Stretton, England. Academician Myrddin Davies started technical collaboration with the IAEA in the early 1980s, and in 1985 became chairman of the International Working Group on Reliability of Reactor Pressure Components (IWG-RRPC). Under his chairmanship this grew to become the Technical Working Group on Life Management of Nuclear Power Plants (TWG-LMNPP) covering broader issues and with world wide collaboration. An insight to the creation, working methods and achievements of the TWG-LMNPP is given in this paper. Acad. Myrddin Davies was a competent chairman at many specialist meetings, major conferences hosted by IAEA, other European organizations and Nuclear Engineering International activities. The direction given to the TWG-LMNPP by Acad. Myrddin Davies is shown to have made a significant contribution to the safe use of nuclear energy. Major contributions to nuclear technology of the TWGLMNPP, during the Chairmanship of Myrddin Davies, are thus cited.
05•00658 Methods and practices used in incident analysis in the Finnish nuclear power industry Suksi, S. Journal of Hazardous Materials, 2004, 111, (1-3), 73-79. According to the Finnish Nuclear Energy Act it is licensee's responsibility to ensure safe use of nuclear energy. Radiation and Nuclear Safety Authority (STUK) is the regulatory body responsible for the state supervision of the safe use of nuclear power in Finland. One essential prerequisite for the safe and reliable operation of nuclear power plants is that lessons are learned from the operational experience. It is utility's prime responsibility to assess the operational events and implement appropriate corrective actions. STUK controls licensees' operational experience feedback arrangements and implementation as part of its inspection activities. In addition to this in Finland, the regulatory body performs its own assessment of the operational experience. Review and investigation of operational events is a part of the regulatory oversight of operational safety. Review of operational events is clone by STUK basically at three different levels. First step is to perform a general review of all operational events, transients and reactor scram reports, which the licensees submit for information to STUK. The second level activities are related to the clarification of events at site and entering of events' specific data into the event register database of STUK. This is done for events that meet the set criteria for the operator to submit a special report to STUK for approval. Safety significance of operational events is determined using
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