- Email: [email protected]
02958 Investigation of fuel rod behaviour under extended 1 burnup conditions with ROFEM fuel performance code
05
Nuclear
fuels (scientific, technical)
Zeolite synthesis from fly ash and cement kiln dust 98io2949 Grutzeck, M. W. Ceram. Trans., 1996, 72, 353-364. When added to portland cement paste, zeolites normally undergo a pozzolanic reaction. However, if the composition of the cement is modified by blending it with fly ash, the calcium silicate hydrate (C-S-H) that forms has a low CaO/SiOz ratio which allows it to coexist with a zeolite. If alkali is added to the system, it then becomes possible to nucleate and grow a zeolitic phase with C-S-H. When the fly ash and NaOH are mixed with cement kiln dust, cancrinite-like phases and tobermorite form instead of the usual Nap-1 and analcime. This implies that a zeolite-containing monolith could be produced that would exhibit both the cation-exchange and adsorptive properties of zeolites while retaining the characteristic strength and ease of use attributable to cement based materials. The potential of these composites as a new class of inexpensive cation exchange and/or chemical adsorbents for use in large-scale applications is promising.
05
NUCLEAR Scientific,
FUELS
Technical
Analysis of the Rowlands uranium oxide pin-cell 98102950 benchmark with an updated WIMS-D library Trkov, A. Ann. Nucl. Energy, 1998, 25, (lo), 747-752. In order to obtain a self consistent set which can be used as reference, the Rowlands uranium oxide light water reactor pin-cell numerical benchmark results from the literature were analysed. The materials relevant to the benchmark from the JEF-2.2 evaluated nuclear data file were processed with the NJOY code and the WIMS-D multi-group library was updated. An input for WIMSD-5A was prepared. Integral parameters, which include reaction rates and multiplication factors for the pin cell at different temperatures, moderator density and leakage were calculated. A comparison with the previously defined reference values was completed. 98102951 ;u$rd
Development of a computer code to estimate the failure using primary coolant activities of operatmg
Chun, M.-H. et ui. Ann. Nucl. Energy, 1998, 25, (IO), 753-763. Entitled as ‘CAAP (Coolant Activity Analysis Program)‘, a Windows computer code has been developed to evaluate the number, the degree of failures and the location of failed fuel rods using primary coolant radioactivity data obtained from operating PWRS. New models are developed and incorporated into the CAAP program to improve some of the drawbacks of existing computer codes. The iodine and noble gas activities obtained by grab sampling and that obtained from the primary coolant radioactivity monitoring system can be used to estimate fuel rod failures with CAAP. CAAP has been combined with the primary coolant radioactivity monitoring system at Kori 3 and 4 operating nuclear power plants for an on-line evaluation of the number of failed fuel rods and the degree of failures. The validity of the computational models of CAAP has been examined using nuclear power plant data collected. The number of failed fuel rods has been estimated with CAAP for 29 cycles of PWRs for which ultrasonic inspections were performed at the end of fuel cycles and compared with the ultrasonic inspection results. CAAP was found to give a better agreement with the ultrasonic test data than existing models. In addition, CAAP calculations performed to estimate the region and the burn-up of the failed fuel rods for two cycles, whose locations of failed fuel rods were known, show that CAAP can accurately predict both the region and the burn-up of the failed fuel rods. Diagnostics of detector tube impacting with wave98102952 let techniques R&z, A. and PBzsit, 1. Ann. Nucl. Energy, 25, (6), 387-400. A method based on neutron noise is proposed for detecting impacting of detector tubes in BWRs. The basic idea relies on the assumption that nonstationary transients may be induced at impacting. It is difficult to detect such short-lived transients by spectral analysis methods, but their presence in the detector signal can be detected by wavelet analysis. The Haar transform is a simple wavelet technique, which is suggested impacting detection. Successful testing has been performed on the method both on simulated data with controlled impacting as well as with real measurement data. The simulation model as well as the results of the wavelet analysis are reported in this paper. The source codes written in MATLAB. are available at a public ftp site. Empirical process modelling in fast breeder 98102953 reactors Ikonomopoulos, A. and Endou, A. Ann. Nucl. Energy, 1998, 25, (9), 609621. For monitoring vital system parameters in a nuclear reactor environment, a non-linear multi-input/single output (MISO) empirical model is introduced. The proposed methodology employs a scheme of non-parametric smoothing
272
Fuel and Energy Abstracts July 1998
that models the local dynamics of each fitting point individually, as opposed to global modelling techniques that attempt to capture the dynamics of the entire design space. This approach alleviates the effect of negative interference between old and new observations enhancing the model prediction capabilities. Modelling the behaviour of any given system comes down to a trade-off between variance and bias. The proposed approach tailors each component to each data set through two separate, adaptive procedures in order to optimize the bias-variance reconciliation. Heteroassociative schemes of the technique presented exhibit insensitivity to sensor noise and provide the operator with accurate predictions of the actual process signals. A comparison between the local model and MLP prediction capabilities is performed and the results favour the first method. The data used to demonstrate the potential of local regression have been obtained during two start-up periods of the Monju fast breeder reactor (FBR). The forgotten effect of the finite measurement time gal02954 on various noise analysis techniques Wallerbos, E. J. M. and Hoogenboom, J. E. Ann. Nucl. Energy, 1998, 25. (lo), 733-746. When the reactor is critical, the conventional noise analysis expressions for functions, like the auto- and cross-correlation function, the variance to mean ratio, and the Rossi-a formula, diverge. This problem arises because one pole of the zero-power reactor transfer function is zero. However, in a finite measurement time, a zero frequency cannot be measured and the divergence will not be found experimentally. New expressions for the expectation values of the experimental quantities of various pulse counting techniques are derived which also take into account the dead time of the detector. These expressions do not suffer from divergence at critical. A Feynman-(1 experiment was simulated in two, neutronically different systems. A bias occurs in the inferred reactivity value with the use of the conventional equations for the analysis of the experiments. Fuel alternatives for oil sands development-the 98102955 nuclear option Bock, D. and Donnelly, J. K. Annu. Conf. Proc. Can. Nucl. Sot., 1995, 16, (2), Paper 4.114, I6 pp. In order to determine the magnitude of the bitumen resource which can he economically exploited using the Steam Assisted Gravity Drainage (SAGD) technology, a reservoir screening study was conducted. Fuels other than natural gas must clearly be used if the full potential of oil sands is to he realized. Alternate fuels which are in sufficient supply to have a significant impact on the energy requirements for oil sands development are nuclear energy and those derived from bitumen and coal. As a possible alternative for providing steam for the deep commercial in tim oil sand projects which were initiated over 10 years ago, Canadian nuclear energy is explored. As the in situ technology of that time required steam at pressures > IO MPa. the nuclear option required the development of new reactor technology or the use of steam compressors which was not economical. The current SAGD technology requires steam at pressures of <5 MPa which is in the reach of existing Canadian nuclear technology. The cost of supplying steam for a SAGD in siru project using a CANDU 3 nuclear reactor were developed. For gas prices in excess of $2.50 per GJ, replacing natural gas fuel with a nuclear reactor is economically feasible for in situ projects > 123 thousand barrels per day. 98102958 Improvement of direct contact condensation model of RELAPS/MODB.I for passive high-pressure injection system Lee, S. I. and No, H. C. Ann. Nucl. Energy, 1998, 25, (9), 677-688. The authors developed a simple set of the transition criterion of the condensation regimes and the heat transfer coefficients on the direct contact condensation in the core makeup tank, implemented in RELAPSI MOD3.1. The condensation regimes were divided into two: supply limit and condensation limit. In modelling the transition criterion hetween two regimes, a large-eddy model developed by Theofanous was used. The modified code better predicted the experiments on the core makeup tank using a smaller scale test facility than the original code. 98102957 Interpretation of transmutation rates of minor actinides in thermal and fast reactors Takeda, T. et al. Ann. Nucl. Energy, 1998, 25, (9), 653-665. The four components of the commonly used transmutation rate of minor actinides in nuclear reactors are: overall fission rate, plutonium production rate, MA production rate, and element production rate. The paper describes the physical meanings of these factors. The transmutation rates of minor actinides in two types of highly-moderated PWRs, a MOX fuelled sodium cooled fast reactor, and a metal fuelled lead cooled fast reactor are interpreted using the four components. 98lO2958 Investigation of fuel rod behaviour under extended 1 burnup conditions with ROFEM fuel performance code Horhoianu, G. ef al. Ann. Nucl. Energy, 1998, 25, (lo), 695-708. The economics of water-reactor operation can be improved by extending burn-up via enhanced fuel utilization and reduced spent fuel volume. A dedicated fuel behaviour modelling computer code (entitled ROFEM-IB) has been developed to analyse high burn-up fuel performance. The code was benchmarked on an experimental data base which include a significant number of irradiation experiments performed in TRIGA-INR Pitesti research reactor. Five fuel rod behaviour during irradiation up to 50
06 MWd kg-‘UOz-’ burn-up have been analysed by the code in the framework of the first phase of the international FUMEX code exercise co-ordinated by IAFA, Vienna. The input experimental data package has been prepared by IFE-OECD Halden, Norway laboratory. In the second phase of the FUMEX exercise the participants have analysed eight simplified theoretical cases. The results obtained with ROFEM-are presented and discussed and the comparison between code predictions and experimental data are reported. 98102959 Measurement of eigenvalue separation by using position sensitive proportional counter Ishitani. K. ef al. Ann. Nucl. Energy, 1998, 2.5, (IO), 721-732. From the viewpoint of neutronics, an important index of power reactor stability is eigenvalue separation. In recent years, a long, slender position sensitive proportional counter has been developed, which facilitates the rapid measurement of spatial neutron flux distribution of a nuclear reactor. Using these characteristics, the authors intend to develop the experimental method evaluating eigenvalue separation as an index for reactor stability. A method based on spatial integrals of neutron flux measured at a subcritical system with an external source is proposed. Experiments were carried out at Kyoto University Critical Assembly to verify this method. It was confirmed that eigenvalue separation can be evaluated within a relative error of 10%. A model for the fuel pin behavior in a pressurized98102960 water reactor operating in transitional and load-following states Proshkin, A. A. et al. AI. Eflerg?, 1997, 82, (h), 469-473. The paper performs data analysis on the effect of the pin behaviour upon power stepping as well as the effect of thermomechanical interaction between fuel and cladding upon power stepping from Russian and foreign sources. Simulation of corrosion product activity in pressur98102961 ized water reactors under flow rate transients Mirza, A. M. et al. Ann. Nucl. Etlergy, 1998, 25, (6). 331-345. Coolant activation due to corrosion products and impurities in a typical pressurized water reactor has been simulated under flow rate transients. In order to calculate the coolant specific activity, an approach was developed employing time dependent production and losses of corrosion products in the primary coolant path. The results obtained reveal that the specific activity decreases and the rate of decrease depends on pump half time and the reactor scram conditions.
Economics, Policy, Supplies, Forecasts Bayesian analysis of public views on the safety of 98102962 nuclear developments Yamagata, H. and Kanda, K. Ann. Nucl. Energy, 1998, 25, (lo), 709-720. Public distrust of nuclear development policy of Japan has arisen as a result of the sodium leak accident at Monju in combination with the PNC’s management of the accident. To learn from these events and assess how a relationship of mutual trust can be established with the public, this paper simulates the processes of establishing public confidence in reliability of a nuclear plant in cases of accidents, no accident, a cover-up, etc. Public confidence is defined here as the public evaluation of cumulative probability under a certain level of accident rate, conditional on the information available to the public. The conditional probability is estimated by use of Bayes’ Theorem. The simulation shows that public confidence is lost by only one accident in an early stage of operation and can then be recovered only by many subsequent years of accident-free operation, but never by a cover-up. Also it claims that the more information that is provided to the public, the better the relationship of mutual trust that will be established, especially at an early stage of plant operations. Coal subsidization and nuclear phase-out in a 98102963 general equilibrium model for Germany Welsch, H. Energy Economics, 1998, 20, (2), 203-222. In German energy policy, the subsidization of domestic hard coal and the future of nuclear power generation are highly controversial issues. This paper offers an analysis of a nuclear phase-out scenario and a scenario of accelerated dismantling of coal subsidies within a dynamic general equilibrium model for Germany. The German economy is modelled in the context of the European Union, because European energy markets are becoming increasingly integrated. It is found that the nuclear phase-out has an effect both on the sectoral and the macroeconomic level, including a modest decrease in GDP, and a substantial increase in CO2 emissions. The accelerated removal of coal subsidies has a noticeable effect on the sectoral and macroeconomic structure as well, but the effect on GDP is found to be rather negligible. CO2 emissions are only reduced slightly, because domestic coal is mainly replaced by imported coal. On the basis of these findings it is conjectured that policy making on these issues is likely to be based on politics rather than strictly macroeconomic considerations.
Electrical power supply and utilization (scientific, technical)
98lO2964 The European PWR (EPR) Kasper, K. J. VGB Krujiwerkstech., 1998, 78, (2), 29-34. (In German) The European PWR is presented as a co-operation of various German energy support companies and the Framatome. Preliminary the political aspects of energy production and the novel German regulation on nuclear power production is discussed. The basic design and its optimization of the European PWR, as well as safety improvement and economical aspects are reported. Compared to a coal fired electric power plant, the advantages of the European PWR are demonstrated. 98lO2965 Nuclear fuel cycle cost analysis using a probabilistic simulation technique Ko, W. I. ef al. Ann. Nucl. Energy, 1998, 25, (IO), 771-789. The paper describes a simple approach to incorporate the Monte Carlo simulation technique into a fuel cycle cost estimate. As a case study, the once-through and recycle fuel cycle options were tested with some alternatives and the simulation results were compared with the values calculated by a deterministic method. A three-estimate approach was used for converting cost inputs into the statistical parameters of assumed probabilistic distributions. It was indicated that the Monte Carlo simulation by a Latin Hypercube Sampling technique and subsequent sensitivity analyses were useful for examining uncertainty propagation of fuel cycle costs and could more efficiently provide information to decisions makers than a deterministic method. The results obtained are presented in detail and could be useful in applications to another options, such as the DUPIC (Direct Use of PWR spent fuel In CANDU reactors) cycle with high cost uncertainty. 98102966 The risks of the nuclear policies Romerio, F. Energy Policy, 1998, 26, (3), 239-246. An evaluation of the risks of nuclear policies is undertaken from the viewpoint of security of energy supplies and of the environment. A model is defined which helps to explain and evaluate the choice of the nuclear sector, electric utilities and governments. This model takes into consideration the main elements which may restrain nuclear energy development, in particular, electric consumption evolution, production costs, fuel resources, major accidents, disposal of highly radioactive wastes and development of new technologies allowing to reduce emissions from the coal fired power station. Also examined are the problem of the gap existing between ‘the reality’ and ‘the objectives of the actors’, because it allows to understand some energy policy decisions. The risks provoked by energy strategies are investigated which try to realize their objectives through an unilateral technological choice. The economic and political advantages of diversification and flexibility are highlighted. The importance of objectivity and transparency is also stressed. This paper intends to he a contribution to discussions on sustainable development and on the future of nuclear energy. 98lO2967 United States energy supplies for the Zlst century Penner, S. S. Energy, 23, (2), 71-78. The likely sources energy supplies necessary for the US for the 21st century are outlined.
06
ELECTRICAL POWER SUPPLY AND UTILIZATION Scientific, Technical
98lO2968 The ‘advanced DIR-MCFC development’ project, an overview Kortbeek, P. J. and Ottervanger, R. .I. Power Sources, 1998, 71, (l/2), 223225. The approach and mid-term status of the joint European ‘Advanced DIRMCFC Development’ project, in which BCN, BG plc, GDF, ECN, Stork, Schelde and Sydkraft co-operate, are presented. Hospitals are identified as an attractive initial market for cogeneration direct internal reforming molten carbonate fuel cell (DIR-MCFC) systems in the size of 400 kW,. Innovative system and stack design concepts are being developed for this application. The ‘SMARTER’ system, based on DIR stacks, combines high electric efficiency and a wide operational window with optimal system simplicity and low cost.
Fuel and Energy Abstracts
July 1998
273
Nuclear
fuels (scientific, technical)
Zeolite synthesis from fly ash and cement kiln dust 98io2949 Grutzeck, M. W. Ceram. Trans., 1996, 72, 353-364. When added to portland cement paste, zeolites normally undergo a pozzolanic reaction. However, if the composition of the cement is modified by blending it with fly ash, the calcium silicate hydrate (C-S-H) that forms has a low CaO/SiOz ratio which allows it to coexist with a zeolite. If alkali is added to the system, it then becomes possible to nucleate and grow a zeolitic phase with C-S-H. When the fly ash and NaOH are mixed with cement kiln dust, cancrinite-like phases and tobermorite form instead of the usual Nap-1 and analcime. This implies that a zeolite-containing monolith could be produced that would exhibit both the cation-exchange and adsorptive properties of zeolites while retaining the characteristic strength and ease of use attributable to cement based materials. The potential of these composites as a new class of inexpensive cation exchange and/or chemical adsorbents for use in large-scale applications is promising.
05
NUCLEAR Scientific,
FUELS
Technical
Analysis of the Rowlands uranium oxide pin-cell 98102950 benchmark with an updated WIMS-D library Trkov, A. Ann. Nucl. Energy, 1998, 25, (lo), 747-752. In order to obtain a self consistent set which can be used as reference, the Rowlands uranium oxide light water reactor pin-cell numerical benchmark results from the literature were analysed. The materials relevant to the benchmark from the JEF-2.2 evaluated nuclear data file were processed with the NJOY code and the WIMS-D multi-group library was updated. An input for WIMSD-5A was prepared. Integral parameters, which include reaction rates and multiplication factors for the pin cell at different temperatures, moderator density and leakage were calculated. A comparison with the previously defined reference values was completed. 98102951 ;u$rd
Development of a computer code to estimate the failure using primary coolant activities of operatmg
Chun, M.-H. et ui. Ann. Nucl. Energy, 1998, 25, (IO), 753-763. Entitled as ‘CAAP (Coolant Activity Analysis Program)‘, a Windows computer code has been developed to evaluate the number, the degree of failures and the location of failed fuel rods using primary coolant radioactivity data obtained from operating PWRS. New models are developed and incorporated into the CAAP program to improve some of the drawbacks of existing computer codes. The iodine and noble gas activities obtained by grab sampling and that obtained from the primary coolant radioactivity monitoring system can be used to estimate fuel rod failures with CAAP. CAAP has been combined with the primary coolant radioactivity monitoring system at Kori 3 and 4 operating nuclear power plants for an on-line evaluation of the number of failed fuel rods and the degree of failures. The validity of the computational models of CAAP has been examined using nuclear power plant data collected. The number of failed fuel rods has been estimated with CAAP for 29 cycles of PWRs for which ultrasonic inspections were performed at the end of fuel cycles and compared with the ultrasonic inspection results. CAAP was found to give a better agreement with the ultrasonic test data than existing models. In addition, CAAP calculations performed to estimate the region and the burn-up of the failed fuel rods for two cycles, whose locations of failed fuel rods were known, show that CAAP can accurately predict both the region and the burn-up of the failed fuel rods. Diagnostics of detector tube impacting with wave98102952 let techniques R&z, A. and PBzsit, 1. Ann. Nucl. Energy, 25, (6), 387-400. A method based on neutron noise is proposed for detecting impacting of detector tubes in BWRs. The basic idea relies on the assumption that nonstationary transients may be induced at impacting. It is difficult to detect such short-lived transients by spectral analysis methods, but their presence in the detector signal can be detected by wavelet analysis. The Haar transform is a simple wavelet technique, which is suggested impacting detection. Successful testing has been performed on the method both on simulated data with controlled impacting as well as with real measurement data. The simulation model as well as the results of the wavelet analysis are reported in this paper. The source codes written in MATLAB. are available at a public ftp site. Empirical process modelling in fast breeder 98102953 reactors Ikonomopoulos, A. and Endou, A. Ann. Nucl. Energy, 1998, 25, (9), 609621. For monitoring vital system parameters in a nuclear reactor environment, a non-linear multi-input/single output (MISO) empirical model is introduced. The proposed methodology employs a scheme of non-parametric smoothing
272
Fuel and Energy Abstracts July 1998
that models the local dynamics of each fitting point individually, as opposed to global modelling techniques that attempt to capture the dynamics of the entire design space. This approach alleviates the effect of negative interference between old and new observations enhancing the model prediction capabilities. Modelling the behaviour of any given system comes down to a trade-off between variance and bias. The proposed approach tailors each component to each data set through two separate, adaptive procedures in order to optimize the bias-variance reconciliation. Heteroassociative schemes of the technique presented exhibit insensitivity to sensor noise and provide the operator with accurate predictions of the actual process signals. A comparison between the local model and MLP prediction capabilities is performed and the results favour the first method. The data used to demonstrate the potential of local regression have been obtained during two start-up periods of the Monju fast breeder reactor (FBR). The forgotten effect of the finite measurement time gal02954 on various noise analysis techniques Wallerbos, E. J. M. and Hoogenboom, J. E. Ann. Nucl. Energy, 1998, 25. (lo), 733-746. When the reactor is critical, the conventional noise analysis expressions for functions, like the auto- and cross-correlation function, the variance to mean ratio, and the Rossi-a formula, diverge. This problem arises because one pole of the zero-power reactor transfer function is zero. However, in a finite measurement time, a zero frequency cannot be measured and the divergence will not be found experimentally. New expressions for the expectation values of the experimental quantities of various pulse counting techniques are derived which also take into account the dead time of the detector. These expressions do not suffer from divergence at critical. A Feynman-(1 experiment was simulated in two, neutronically different systems. A bias occurs in the inferred reactivity value with the use of the conventional equations for the analysis of the experiments. Fuel alternatives for oil sands development-the 98102955 nuclear option Bock, D. and Donnelly, J. K. Annu. Conf. Proc. Can. Nucl. Sot., 1995, 16, (2), Paper 4.114, I6 pp. In order to determine the magnitude of the bitumen resource which can he economically exploited using the Steam Assisted Gravity Drainage (SAGD) technology, a reservoir screening study was conducted. Fuels other than natural gas must clearly be used if the full potential of oil sands is to he realized. Alternate fuels which are in sufficient supply to have a significant impact on the energy requirements for oil sands development are nuclear energy and those derived from bitumen and coal. As a possible alternative for providing steam for the deep commercial in tim oil sand projects which were initiated over 10 years ago, Canadian nuclear energy is explored. As the in situ technology of that time required steam at pressures > IO MPa. the nuclear option required the development of new reactor technology or the use of steam compressors which was not economical. The current SAGD technology requires steam at pressures of <5 MPa which is in the reach of existing Canadian nuclear technology. The cost of supplying steam for a SAGD in siru project using a CANDU 3 nuclear reactor were developed. For gas prices in excess of $2.50 per GJ, replacing natural gas fuel with a nuclear reactor is economically feasible for in situ projects > 123 thousand barrels per day. 98102958 Improvement of direct contact condensation model of RELAPS/MODB.I for passive high-pressure injection system Lee, S. I. and No, H. C. Ann. Nucl. Energy, 1998, 25, (9), 677-688. The authors developed a simple set of the transition criterion of the condensation regimes and the heat transfer coefficients on the direct contact condensation in the core makeup tank, implemented in RELAPSI MOD3.1. The condensation regimes were divided into two: supply limit and condensation limit. In modelling the transition criterion hetween two regimes, a large-eddy model developed by Theofanous was used. The modified code better predicted the experiments on the core makeup tank using a smaller scale test facility than the original code. 98102957 Interpretation of transmutation rates of minor actinides in thermal and fast reactors Takeda, T. et al. Ann. Nucl. Energy, 1998, 25, (9), 653-665. The four components of the commonly used transmutation rate of minor actinides in nuclear reactors are: overall fission rate, plutonium production rate, MA production rate, and element production rate. The paper describes the physical meanings of these factors. The transmutation rates of minor actinides in two types of highly-moderated PWRs, a MOX fuelled sodium cooled fast reactor, and a metal fuelled lead cooled fast reactor are interpreted using the four components. 98lO2958 Investigation of fuel rod behaviour under extended 1 burnup conditions with ROFEM fuel performance code Horhoianu, G. ef al. Ann. Nucl. Energy, 1998, 25, (lo), 695-708. The economics of water-reactor operation can be improved by extending burn-up via enhanced fuel utilization and reduced spent fuel volume. A dedicated fuel behaviour modelling computer code (entitled ROFEM-IB) has been developed to analyse high burn-up fuel performance. The code was benchmarked on an experimental data base which include a significant number of irradiation experiments performed in TRIGA-INR Pitesti research reactor. Five fuel rod behaviour during irradiation up to 50
06 MWd kg-‘UOz-’ burn-up have been analysed by the code in the framework of the first phase of the international FUMEX code exercise co-ordinated by IAFA, Vienna. The input experimental data package has been prepared by IFE-OECD Halden, Norway laboratory. In the second phase of the FUMEX exercise the participants have analysed eight simplified theoretical cases. The results obtained with ROFEM-are presented and discussed and the comparison between code predictions and experimental data are reported. 98102959 Measurement of eigenvalue separation by using position sensitive proportional counter Ishitani. K. ef al. Ann. Nucl. Energy, 1998, 2.5, (IO), 721-732. From the viewpoint of neutronics, an important index of power reactor stability is eigenvalue separation. In recent years, a long, slender position sensitive proportional counter has been developed, which facilitates the rapid measurement of spatial neutron flux distribution of a nuclear reactor. Using these characteristics, the authors intend to develop the experimental method evaluating eigenvalue separation as an index for reactor stability. A method based on spatial integrals of neutron flux measured at a subcritical system with an external source is proposed. Experiments were carried out at Kyoto University Critical Assembly to verify this method. It was confirmed that eigenvalue separation can be evaluated within a relative error of 10%. A model for the fuel pin behavior in a pressurized98102960 water reactor operating in transitional and load-following states Proshkin, A. A. et al. AI. Eflerg?, 1997, 82, (h), 469-473. The paper performs data analysis on the effect of the pin behaviour upon power stepping as well as the effect of thermomechanical interaction between fuel and cladding upon power stepping from Russian and foreign sources. Simulation of corrosion product activity in pressur98102961 ized water reactors under flow rate transients Mirza, A. M. et al. Ann. Nucl. Etlergy, 1998, 25, (6). 331-345. Coolant activation due to corrosion products and impurities in a typical pressurized water reactor has been simulated under flow rate transients. In order to calculate the coolant specific activity, an approach was developed employing time dependent production and losses of corrosion products in the primary coolant path. The results obtained reveal that the specific activity decreases and the rate of decrease depends on pump half time and the reactor scram conditions.
Economics, Policy, Supplies, Forecasts Bayesian analysis of public views on the safety of 98102962 nuclear developments Yamagata, H. and Kanda, K. Ann. Nucl. Energy, 1998, 25, (lo), 709-720. Public distrust of nuclear development policy of Japan has arisen as a result of the sodium leak accident at Monju in combination with the PNC’s management of the accident. To learn from these events and assess how a relationship of mutual trust can be established with the public, this paper simulates the processes of establishing public confidence in reliability of a nuclear plant in cases of accidents, no accident, a cover-up, etc. Public confidence is defined here as the public evaluation of cumulative probability under a certain level of accident rate, conditional on the information available to the public. The conditional probability is estimated by use of Bayes’ Theorem. The simulation shows that public confidence is lost by only one accident in an early stage of operation and can then be recovered only by many subsequent years of accident-free operation, but never by a cover-up. Also it claims that the more information that is provided to the public, the better the relationship of mutual trust that will be established, especially at an early stage of plant operations. Coal subsidization and nuclear phase-out in a 98102963 general equilibrium model for Germany Welsch, H. Energy Economics, 1998, 20, (2), 203-222. In German energy policy, the subsidization of domestic hard coal and the future of nuclear power generation are highly controversial issues. This paper offers an analysis of a nuclear phase-out scenario and a scenario of accelerated dismantling of coal subsidies within a dynamic general equilibrium model for Germany. The German economy is modelled in the context of the European Union, because European energy markets are becoming increasingly integrated. It is found that the nuclear phase-out has an effect both on the sectoral and the macroeconomic level, including a modest decrease in GDP, and a substantial increase in CO2 emissions. The accelerated removal of coal subsidies has a noticeable effect on the sectoral and macroeconomic structure as well, but the effect on GDP is found to be rather negligible. CO2 emissions are only reduced slightly, because domestic coal is mainly replaced by imported coal. On the basis of these findings it is conjectured that policy making on these issues is likely to be based on politics rather than strictly macroeconomic considerations.
Electrical power supply and utilization (scientific, technical)
98lO2964 The European PWR (EPR) Kasper, K. J. VGB Krujiwerkstech., 1998, 78, (2), 29-34. (In German) The European PWR is presented as a co-operation of various German energy support companies and the Framatome. Preliminary the political aspects of energy production and the novel German regulation on nuclear power production is discussed. The basic design and its optimization of the European PWR, as well as safety improvement and economical aspects are reported. Compared to a coal fired electric power plant, the advantages of the European PWR are demonstrated. 98lO2965 Nuclear fuel cycle cost analysis using a probabilistic simulation technique Ko, W. I. ef al. Ann. Nucl. Energy, 1998, 25, (IO), 771-789. The paper describes a simple approach to incorporate the Monte Carlo simulation technique into a fuel cycle cost estimate. As a case study, the once-through and recycle fuel cycle options were tested with some alternatives and the simulation results were compared with the values calculated by a deterministic method. A three-estimate approach was used for converting cost inputs into the statistical parameters of assumed probabilistic distributions. It was indicated that the Monte Carlo simulation by a Latin Hypercube Sampling technique and subsequent sensitivity analyses were useful for examining uncertainty propagation of fuel cycle costs and could more efficiently provide information to decisions makers than a deterministic method. The results obtained are presented in detail and could be useful in applications to another options, such as the DUPIC (Direct Use of PWR spent fuel In CANDU reactors) cycle with high cost uncertainty. 98102966 The risks of the nuclear policies Romerio, F. Energy Policy, 1998, 26, (3), 239-246. An evaluation of the risks of nuclear policies is undertaken from the viewpoint of security of energy supplies and of the environment. A model is defined which helps to explain and evaluate the choice of the nuclear sector, electric utilities and governments. This model takes into consideration the main elements which may restrain nuclear energy development, in particular, electric consumption evolution, production costs, fuel resources, major accidents, disposal of highly radioactive wastes and development of new technologies allowing to reduce emissions from the coal fired power station. Also examined are the problem of the gap existing between ‘the reality’ and ‘the objectives of the actors’, because it allows to understand some energy policy decisions. The risks provoked by energy strategies are investigated which try to realize their objectives through an unilateral technological choice. The economic and political advantages of diversification and flexibility are highlighted. The importance of objectivity and transparency is also stressed. This paper intends to he a contribution to discussions on sustainable development and on the future of nuclear energy. 98lO2967 United States energy supplies for the Zlst century Penner, S. S. Energy, 23, (2), 71-78. The likely sources energy supplies necessary for the US for the 21st century are outlined.
06
ELECTRICAL POWER SUPPLY AND UTILIZATION Scientific, Technical
98lO2968 The ‘advanced DIR-MCFC development’ project, an overview Kortbeek, P. J. and Ottervanger, R. .I. Power Sources, 1998, 71, (l/2), 223225. The approach and mid-term status of the joint European ‘Advanced DIRMCFC Development’ project, in which BCN, BG plc, GDF, ECN, Stork, Schelde and Sydkraft co-operate, are presented. Hospitals are identified as an attractive initial market for cogeneration direct internal reforming molten carbonate fuel cell (DIR-MCFC) systems in the size of 400 kW,. Innovative system and stack design concepts are being developed for this application. The ‘SMARTER’ system, based on DIR stacks, combines high electric efficiency and a wide operational window with optimal system simplicity and low cost.
Fuel and Energy Abstracts
July 1998
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