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03302 Moderator temperature effect on reactivity in light water moderated experimental reactors
05 Nuclear
fuels (scientific, technical)
(usually the elongation at break is used as the critical parameter). A method based on the measurement of the thermo-oxidative stability by differential scanning calorimetry has been developed in the case of the cables installed in the NPP hermetic zone. Design options for the safety injection system of 00/03299 Korean next generation reactor Bae, K.H. er ol. Annuls o/’ NW/. Encvgy. 2000, 27, (1 I). 101 lLlO2X. Loss of coolant accident (LOCA) analyses for various configurations of safety injection system (SIS) are performed to optimize the emergency core cooling system (ECCS) performance for the Korean next generation reactor (KNGR). The KNGR is an advanced light water reactor (ALWR) adopting the advanced design feature of a direct vessel injection (DVI) configuration and passive fluidic device in the discharge line of the safety injection tank (SIT). To determine the feasible SIS configuration and the optimum capacities of the SIT and high pressure safety injection pump (HPSIP), licensing &sign basis and best estimate LOCA analyses are performed for the limiting large break and small break spectrum, respectively. The analyses results show that the four-train DVI injection with the current system design is a more feasible configuration than the other ones considered and the adoption of a fluidic device SIT enhances the ECCS performance for large break LOCA. For small break LOCA, in the case of cold leg break, the DV14 configuration is better than other configurations and also meets the EPRI ALWR requirement of no core uncovery for up to a 15.24 cm (6 in) all break. However, in the case of DVI line break, slight core uncovery is predicted and also the system behaviour is significantly affected by reactor vessel (RV) downcomer modelling. Therefore, the DV14 configuration is more feasible for KNGR ECCS performance, but further investigations are required to resolve the ECCS bypass issues for large break LOCA and to develop a proper RV downcomer model for analysis of DVI line break in small break LOCA. Improvement of the axial power distribution control 00103300 capabilities in VVER-1000 reactors Yousefpour, F. and Ghofrani, M. B. Annrrh of NW/. Energy, 2000, 27, (IO), 949-957. This article describes an automatic reactor power control system for a VVER-1000 reactor and reports simulation analysis results for a typical daily load follow operation. The associated reactor control algorithm is called ‘mode G’ that uses a heavy-worth bank (H-bank) dedicated to axial power shape control and the light-grey banks (G-banks) for reactor power change and reactivity compensation. The simulation results for daily load follow operation in three burn-up states of first cycle illustrate that the load follow capability of VVER-1000 reactors using this algorithm will be improved. Instability analysis on xenon spatial oscillation in a 00/03301 CANDU-6 reactor with DUPIC fuel Jeong, C. J. and Choi, H. Anmls qfNuc/. Energy, 2000, 27, (lo), 887-899. The instability induced by xenon spatial oscillation of a CANDU-6 reactor with DUPIC fuel has been assessed for three important harmonic perturbations: top-to-bottom, side-to-side and front-to-back oscillations. For each oscillation, the instability index of the DUPIC fuel core has been calculated and compared with that of the natural uranium core. Parametric calculations have also been performed to analyse the effect of the power level and axial power shape on the xenon oscillation. This study has shown that the instability due to xenon oscillation increases for the DUPIC fuel core compared with the natural uranium core. However, this study has also shown that the current reactivity device system suppresses the xenon oscillation completely for both the natural uranium and the DUPIC fuel cores. Moderator temperature effect on reactivity in light 00103302 water moderated experimental reactors Mohapatra, D.K. and Mohanakrishnan, P. Annals Oj‘Nucl. Energy, 2000, 27, (I I), 969-983. The moderator temperature coefficient of reactivity has been measured in a U-233 fuelled plate type light water moderated reactor (KAMINI). Using neutron cross sections based on WIMS library, lattice homogenization code SMAXY and 3D core calculation code COMESH, the moderator temperature coefficient of reactivity is predicted very well (within 0.6 pcm per “C). For U-235 and Pu-239 fuelled KRITZ experiments also, moderator temperature coefficients of reactivity have been predicted and compared with Studsvik results using lattice code MURLI and 2D core calculations. The predictions of the moderator temperature coefficient of reactivity are reasonable for these cores. For U-233-fuelled KAMINI, the close agreement on moderator temperature coefficient of reactivity was not found using more recent U-233 cross-sections derived from ENDWB-IV (WIMKAL). By comparing the o (capture to fission ratio) of U-233 at low energies between WIMS and ENDFIB-IV and VI, the better comparison with WIMS is attributed to the flatter CI behaviour with energy. There appears to be a need for measurement of CLof U-233 below 0.1 eV. Neutron cross-sections of 238U in 00103303 0.01-5.5 MeV Vladuca, G. el ul. Anno/:, of Ntrcl. Energy, 2000. 27, Calculated neutron cross-sections in agreement with have been obtained for “‘U in the energy range compound nucleus calculations have been performed
372
Fuel and Energy Abstracts
the energy
range
(1 I), 995-1010. the experimental data 0.01-5.5 MeV. The with a version of the
November 2000
HRTW statistical model extended to subbarrier excitation energies, which includes the calculation of the isomeric fission. The results confirm the power of prediction of this model for the actinide nuclei. 00/03304 Neutron spectrum measurements from 1-16 MeV in beryllium assemblies with a central D-T neutron source Koohi-Fayegh, R. el crl. Annals of Nucl. Energy, 2000, 27, (I I), 959-967. Beryllium has been proposed as both a neutron multiplier and as a plasma facing material for fusion reactors, and its neutron cross-section up to 15 MeV or so is consequently of considerable importance, particularly for the (n, 2n) reaction. This paper reports on the results of a series of 14 MeV neutron transmission measurements using an NE213 scintillation spectrometer with three different thicknesses of beryllium shells. The experimental results are then compared with MCNP Monte Carlo calculations using the ENDFIB-VI data set. For all three shells the experimental results lie above those calculated for neutron energies between 8 and 11 MeV, while between 1 and 4 MeV they lie below. It is concluded that there are continuing uncertainties in the data. 00103305 Optimized conventional island design for the European pressurized water reactor (EPR) Schuberth, U. er al. ATW. Inr. Z. Kernenerg., 1999, 44, (6), 375-381. The main technical objectives of the EPR, an evolutionary nuclear reactor design derived from the German Konvoi series and the French N4 series, are as follows. The EPR had to be able to compete economically with other nuclear plant designs as well as hard-coal fired power plants. It had to provide satisfactory performance at a generating capacity of 1750-1850 MW,, a plant service lifetime of 60 years, at improved safety. The basic design of the nuclear island was developed to improve the efficiency, the reliability, and economy of the EPR. 00103306 Reactor physics calculations and their experimental validation for conversion and upgrading of a typical swimming pool type research reactor Khan, L. A. er ol. Anna/s 01 NW/. Energy, 2000, 27, (IO), 873-885. Detailed neutronic analysrs of a typical swimming pool type research reactor, Pakistan Research Reactor-l (PARR-l), was carried out for conversion of its core from 93% highly enriched uranium to 20% low enriched uranium fuel with power upgrading from 5 to 10 MW. Standard computer codes WIMS-D/4 and CITATION were employed to calculate core excess reactivity, power defect, reactivity effect of xenon and samarium, reactivity worth of fuel element, worth of control rods, shutdown margin, reactivity feedback coefficients, neutron flux and power peaking factors. A series of low and high power tests were performed on the newly converted core to determine its performance. A comparison between the calculated and measured results is preSented. The agreement is generally good. 00/03307 Some peculiar features for kinetics of fission gas release from nuclear fuel on fast heating up Ivanov. A. S. and Ivanov. D. A. Annuls of Nuclear Energy, 2000, 27, (8), 697-71 I. This paper presents studies of some anomalies observed in kinetics of fission products based on the model of porous fuel. A non-m onotone twopeak structure of a fission product flux is shown to be deter mined by the effect of the system of open- pores. It has been established that on heating, an explosive growth of a flux in the region of relatively high temperatures is associated with a high non-linearity of the time dependence of phenomenological parameters describing a fission gas release from a solid phase. The analysis has been performed, of the processes accompanying a decay of a fission product solid solution. The possibility of a fuel material dispersion has been noted under the conditions of a fast reactivity accident due to a sharp rise of the gas phase pressure in the system of closed pores. 00103308 The importance of flow patterns Nedderman, J. Ntrcl. Eng. In?., 2000, 45, (551), 32-33. A coolant leak at Tsuruga revealed limitations in the design of heat exchangers. The leak in question was traced to cracks in an elbow joint in the regenerative heat exchanger of the charging and volume control system (CVCS). This prompted further investigation of thermal stratification. 00103309 Thermodynamic study of the behavior of uranium and plutonium during thermal treatment under reducing and oxidizing modes Ho, T. C. l+‘u:rrsre ~Manugemen!, 2000, 20, (5/6), 355-361. This study investigated the equilibrium compositions of uranium and plutonium under various thermal treatment conditions using an incineration equilibrium calculation programme. The treatment conditions examined included temperature, oxygen level (either reducing or oxidizing), and the existence of chlorine. In a simulation, a selected waste containing either uranium or plutonium was input to the programme along with the desired treatment conditions. The programme then performed the free energy calculations and searched for the optimum composition which minimizes the total system free energy. The simulation results have indicated that, under a reducing mode, uranium tends to stay in a solid phase as U02(s) up to 1500°C; however, under an oxidizing mode, it will exist as U30s(s) up to 1100°C. As the temperature increases, the solid-phase compounds either vaporize or decompose into various vapour-phase compounds. Under a reducing mode all the preferred compounds will be in vapour phase when
fuels (scientific, technical)
(usually the elongation at break is used as the critical parameter). A method based on the measurement of the thermo-oxidative stability by differential scanning calorimetry has been developed in the case of the cables installed in the NPP hermetic zone. Design options for the safety injection system of 00/03299 Korean next generation reactor Bae, K.H. er ol. Annuls o/’ NW/. Encvgy. 2000, 27, (1 I). 101 lLlO2X. Loss of coolant accident (LOCA) analyses for various configurations of safety injection system (SIS) are performed to optimize the emergency core cooling system (ECCS) performance for the Korean next generation reactor (KNGR). The KNGR is an advanced light water reactor (ALWR) adopting the advanced design feature of a direct vessel injection (DVI) configuration and passive fluidic device in the discharge line of the safety injection tank (SIT). To determine the feasible SIS configuration and the optimum capacities of the SIT and high pressure safety injection pump (HPSIP), licensing &sign basis and best estimate LOCA analyses are performed for the limiting large break and small break spectrum, respectively. The analyses results show that the four-train DVI injection with the current system design is a more feasible configuration than the other ones considered and the adoption of a fluidic device SIT enhances the ECCS performance for large break LOCA. For small break LOCA, in the case of cold leg break, the DV14 configuration is better than other configurations and also meets the EPRI ALWR requirement of no core uncovery for up to a 15.24 cm (6 in) all break. However, in the case of DVI line break, slight core uncovery is predicted and also the system behaviour is significantly affected by reactor vessel (RV) downcomer modelling. Therefore, the DV14 configuration is more feasible for KNGR ECCS performance, but further investigations are required to resolve the ECCS bypass issues for large break LOCA and to develop a proper RV downcomer model for analysis of DVI line break in small break LOCA. Improvement of the axial power distribution control 00103300 capabilities in VVER-1000 reactors Yousefpour, F. and Ghofrani, M. B. Annrrh of NW/. Energy, 2000, 27, (IO), 949-957. This article describes an automatic reactor power control system for a VVER-1000 reactor and reports simulation analysis results for a typical daily load follow operation. The associated reactor control algorithm is called ‘mode G’ that uses a heavy-worth bank (H-bank) dedicated to axial power shape control and the light-grey banks (G-banks) for reactor power change and reactivity compensation. The simulation results for daily load follow operation in three burn-up states of first cycle illustrate that the load follow capability of VVER-1000 reactors using this algorithm will be improved. Instability analysis on xenon spatial oscillation in a 00/03301 CANDU-6 reactor with DUPIC fuel Jeong, C. J. and Choi, H. Anmls qfNuc/. Energy, 2000, 27, (lo), 887-899. The instability induced by xenon spatial oscillation of a CANDU-6 reactor with DUPIC fuel has been assessed for three important harmonic perturbations: top-to-bottom, side-to-side and front-to-back oscillations. For each oscillation, the instability index of the DUPIC fuel core has been calculated and compared with that of the natural uranium core. Parametric calculations have also been performed to analyse the effect of the power level and axial power shape on the xenon oscillation. This study has shown that the instability due to xenon oscillation increases for the DUPIC fuel core compared with the natural uranium core. However, this study has also shown that the current reactivity device system suppresses the xenon oscillation completely for both the natural uranium and the DUPIC fuel cores. Moderator temperature effect on reactivity in light 00103302 water moderated experimental reactors Mohapatra, D.K. and Mohanakrishnan, P. Annals Oj‘Nucl. Energy, 2000, 27, (I I), 969-983. The moderator temperature coefficient of reactivity has been measured in a U-233 fuelled plate type light water moderated reactor (KAMINI). Using neutron cross sections based on WIMS library, lattice homogenization code SMAXY and 3D core calculation code COMESH, the moderator temperature coefficient of reactivity is predicted very well (within 0.6 pcm per “C). For U-235 and Pu-239 fuelled KRITZ experiments also, moderator temperature coefficients of reactivity have been predicted and compared with Studsvik results using lattice code MURLI and 2D core calculations. The predictions of the moderator temperature coefficient of reactivity are reasonable for these cores. For U-233-fuelled KAMINI, the close agreement on moderator temperature coefficient of reactivity was not found using more recent U-233 cross-sections derived from ENDWB-IV (WIMKAL). By comparing the o (capture to fission ratio) of U-233 at low energies between WIMS and ENDFIB-IV and VI, the better comparison with WIMS is attributed to the flatter CI behaviour with energy. There appears to be a need for measurement of CLof U-233 below 0.1 eV. Neutron cross-sections of 238U in 00103303 0.01-5.5 MeV Vladuca, G. el ul. Anno/:, of Ntrcl. Energy, 2000. 27, Calculated neutron cross-sections in agreement with have been obtained for “‘U in the energy range compound nucleus calculations have been performed
372
Fuel and Energy Abstracts
the energy
range
(1 I), 995-1010. the experimental data 0.01-5.5 MeV. The with a version of the
November 2000
HRTW statistical model extended to subbarrier excitation energies, which includes the calculation of the isomeric fission. The results confirm the power of prediction of this model for the actinide nuclei. 00/03304 Neutron spectrum measurements from 1-16 MeV in beryllium assemblies with a central D-T neutron source Koohi-Fayegh, R. el crl. Annals of Nucl. Energy, 2000, 27, (I I), 959-967. Beryllium has been proposed as both a neutron multiplier and as a plasma facing material for fusion reactors, and its neutron cross-section up to 15 MeV or so is consequently of considerable importance, particularly for the (n, 2n) reaction. This paper reports on the results of a series of 14 MeV neutron transmission measurements using an NE213 scintillation spectrometer with three different thicknesses of beryllium shells. The experimental results are then compared with MCNP Monte Carlo calculations using the ENDFIB-VI data set. For all three shells the experimental results lie above those calculated for neutron energies between 8 and 11 MeV, while between 1 and 4 MeV they lie below. It is concluded that there are continuing uncertainties in the data. 00103305 Optimized conventional island design for the European pressurized water reactor (EPR) Schuberth, U. er al. ATW. Inr. Z. Kernenerg., 1999, 44, (6), 375-381. The main technical objectives of the EPR, an evolutionary nuclear reactor design derived from the German Konvoi series and the French N4 series, are as follows. The EPR had to be able to compete economically with other nuclear plant designs as well as hard-coal fired power plants. It had to provide satisfactory performance at a generating capacity of 1750-1850 MW,, a plant service lifetime of 60 years, at improved safety. The basic design of the nuclear island was developed to improve the efficiency, the reliability, and economy of the EPR. 00103306 Reactor physics calculations and their experimental validation for conversion and upgrading of a typical swimming pool type research reactor Khan, L. A. er ol. Anna/s 01 NW/. Energy, 2000, 27, (IO), 873-885. Detailed neutronic analysrs of a typical swimming pool type research reactor, Pakistan Research Reactor-l (PARR-l), was carried out for conversion of its core from 93% highly enriched uranium to 20% low enriched uranium fuel with power upgrading from 5 to 10 MW. Standard computer codes WIMS-D/4 and CITATION were employed to calculate core excess reactivity, power defect, reactivity effect of xenon and samarium, reactivity worth of fuel element, worth of control rods, shutdown margin, reactivity feedback coefficients, neutron flux and power peaking factors. A series of low and high power tests were performed on the newly converted core to determine its performance. A comparison between the calculated and measured results is preSented. The agreement is generally good. 00/03307 Some peculiar features for kinetics of fission gas release from nuclear fuel on fast heating up Ivanov. A. S. and Ivanov. D. A. Annuls of Nuclear Energy, 2000, 27, (8), 697-71 I. This paper presents studies of some anomalies observed in kinetics of fission products based on the model of porous fuel. A non-m onotone twopeak structure of a fission product flux is shown to be deter mined by the effect of the system of open- pores. It has been established that on heating, an explosive growth of a flux in the region of relatively high temperatures is associated with a high non-linearity of the time dependence of phenomenological parameters describing a fission gas release from a solid phase. The analysis has been performed, of the processes accompanying a decay of a fission product solid solution. The possibility of a fuel material dispersion has been noted under the conditions of a fast reactivity accident due to a sharp rise of the gas phase pressure in the system of closed pores. 00103308 The importance of flow patterns Nedderman, J. Ntrcl. Eng. In?., 2000, 45, (551), 32-33. A coolant leak at Tsuruga revealed limitations in the design of heat exchangers. The leak in question was traced to cracks in an elbow joint in the regenerative heat exchanger of the charging and volume control system (CVCS). This prompted further investigation of thermal stratification. 00103309 Thermodynamic study of the behavior of uranium and plutonium during thermal treatment under reducing and oxidizing modes Ho, T. C. l+‘u:rrsre ~Manugemen!, 2000, 20, (5/6), 355-361. This study investigated the equilibrium compositions of uranium and plutonium under various thermal treatment conditions using an incineration equilibrium calculation programme. The treatment conditions examined included temperature, oxygen level (either reducing or oxidizing), and the existence of chlorine. In a simulation, a selected waste containing either uranium or plutonium was input to the programme along with the desired treatment conditions. The programme then performed the free energy calculations and searched for the optimum composition which minimizes the total system free energy. The simulation results have indicated that, under a reducing mode, uranium tends to stay in a solid phase as U02(s) up to 1500°C; however, under an oxidizing mode, it will exist as U30s(s) up to 1100°C. As the temperature increases, the solid-phase compounds either vaporize or decompose into various vapour-phase compounds. Under a reducing mode all the preferred compounds will be in vapour phase when