- Email: [email protected]
00781 Mechanical analysis of fuel fretting problem
05
Nuclear fuels (scienfific, fechnical)
energy ENDF/B-V cross-section\. MCNP with continuous energy ENDFIBVI cross-sections and KENO-V.a with 27-group ENDFIB-IV cross-sections. For one case. the codes overestimated k,,, by more than 1%. The experiments were grouped for three parameter\: plutonium mass. horizontal spacing and vertical spacing. For each parameter, a bias value was found. In all cases MCNP with ENDFIB-VI cross-sections had the smallest difference in the average k,tt. This same result was found when all experiments were grouped together. However. if the one experiment that was significantly overestimated was excluded, the smallest difference in the average k,tt was found for KFNO-V.a with ENDFiB-IV cross-sections. A fast-running fuel management program for a 00/00775 CANDU reactor Choi, H. Annul.7 of Ntccl. Ever&?. 21100. 27. ( I ), l-10. A fast-running fuel management programme for a CANDU reactor has been developed. The basic principle of this programme is to select refuelling channels. under circumstances whereby the reference reactor conditions are maintained by applying several constraints and criteria. The constraints used in this programme are the channel and bundle power and the fuel burn-up. The final selection of the re-fuelling channel is determined based on the priority of candidate channels, which enhances the reactor power distrihution close to the time-average model. Flooding in the pressurizer surge line of AP600 00100776 plant and analyses of APEX data Takeuchi, K. er ul. Ntrcl. Engineering & Design. 1999. 192, (1). 45-58. For the passive AP600 plant, the three stages of ADS (automatic depressurization system) valves are attached to the top of pressurizer. The existence of these valves makes liquid flow into and out of the pressurizer, an important part of the dynamics during a small break loss-ofcoolant accident. In this paper, counter-current flow limit (CCFL) in the surge line was analysed. Specifically. CCFL in vertical piping, in slightly inclined horizontal piping and in horizontal and vertical elbows were compared. The CCFL in the vertical section of the surge line was found to be the most limiting section. That is, the vertical CCFL controls the pressurizer liquid drain rate. This conclusion was tested and verified by comparing the predicted vertical CCFL against the counter-current flow states in the surge line. observed in small break LOCA tests conducted at the APhOO scaled test facility (APEX). Gas-liquid
two-phase
flow in narrow horizontal
Ekberg, N. P. et al. Ntrcl. t’~gi,rern’ng & Derign. 1999, I Y2, (I ), 59-80. Experimental data associated with the two-phase flow regimes, void fraction and pressure drop in horizontal, narrow, concentric annuli are presented. Two transparent test sections, one with inner and outer diameters of 6.6 and 8.6 mm, and an overall length of 46.0 cm; the other with 33.2 and 35.2 mm diameters and 43.0 cm length, respectively, were used. Near-atmospheric air and water constituted the gas and liquid phases. The gas and liquid superficial velocities were varied in the 0.02-57 and 0.1-6.1 m s-t ranges, respectively. The major two-phase flow patterns observed included bubbly, slug/plug, churn, stratified and annular. Transitional regimes, where the characteristics of two distinct flow regimes could be observed in the test sections, included bubbly-plug, stratified-slug and annular-slug. The obtained flow regime maps were different than flow regime maps typical of large horizontal channels and microchannels with circular cross-sections. They were also different from the flow regimes in rectangular thin channels. The measured average void fractions for the two test sections were compared with predictions of several empirical correlations. Overall, a correlation proposed by Butterworth [Butterworth, D., 1975. A comparison of some void fraction relationships for co-current gas-liquid flow. Inr. J. Multiphase Flow, 1, 845-X50] provided the most accurate prediction of the measured void fractions. The measured pressure drops were compared with predictions of several empirical correlations. Intelligent system for transient data collection and 00100778 fatigue monitoring of pressurised water reactors nuclear steam supply system Morilhat, P. er al. Ntccl. Engineering & Design, 1999, 192, (I), 103-58. Electricite de France (EDF), the French national electricity company, is operating S4 standardized pressurized water reactors. This about 500 reactor-years experience in nuclear stations operation and maintenance area has allowed EDF to develop its own strategy for monitoring of agerelated degradations of NPP systems and components relevant for plant safety and reliability. After more than IS years of experience in regulatory transient data collection and seven years of successful fatigue monitoring prototypes experimentation, EDF decided to design a new system called SYSFAC (acronym for SYsteme de Surveillance en FAtigue de la Chaudiere) devoted to transient logging and thermal fatigue monitoring of the reactor coolant pressure boundary. The system is fully automatic and directly connected to the on-site data acquisition network without any complementary instrumentation. A functional transient detection module and a mechanical transient detection module are in charge of the general transient data collection. A fatigue monitoring module is aimed towards a precise surveillance of five specific zones particularly sensitive to thermal fatigue. After a first step of preliminary studies, the industrial phase of the SYSFAC project is currently going on, with hardware and software tests
88
Fuel and Energy Abstracts
March 2000
and implementation. The first SYSFAC system wtll he delivered to the pilot power plant by the beginning of 1996. The extension to all EDF’s nuclear YOUMW is planned after one more year of feedback experience. International benchmark experiment in FCA-CEA 00/00779 results Chaussonet, P. er al. Pro~rers in .Nuc/. Energy, IY99, 35, (2). 157-162. To reduce uncertainties on the effective delayed neutron fraction, ,jc,t, an experimental international benchmark was performed at the Fast Critical Assembly (FCA) in JAERI. CEA participation was limited to the two configurations XIX-I and XIX-3 with uranium and plutonium fuel, which can be considered as complementary to the BERENICE benchmark in MASIJRCA. The results ohtained by the noise technique. showed that measured i,,, values are coherent, within the associated uncertainties, with the values obtained hy the other teams and that the target limits (.S?/, at 20) were respected. Measurement of effective delayed neutron fraction 00100780 Qef+by covariance-to-mean method for benchmark experiments of pstt at FCA Sakurni. T. ef al. Progress in Nucl. Enqy, 1999, 35, (2). 203-208. Measurements of the effective delayed neutron fraction $,t were carried out at the FCA facility of the Japan Atomic Energy Research Institute to contribute to benchmark experiment\ of ~$.,t. These measurements were made in two cores: XIX-I core fueled with 93% enriched uranium and XIX-3 core fueled with plutonium. The experimental jc,, was determined with a covariance-to-mean ratio of counts of a pair of neutron detectors installed in the core. The values of .I,,, were 724 ? I3 pcm and 252 % 5 pcm in the cores of XIX-I and XIX-3 respectively. 00100781 Mechanical analysis of fuel fretting problem Kim, H.-K. Nucl. Engirwering& Design, 1999. 192, (I). X1-93. Fuel fretting is studied by contact mechanics approach. Shear force produced by flow-induced vibration is regarded as the major factor of the fuel fretting. Contact dimension is examined for the Korean PWR Fuel Assembly using finite element method. Axial direction is incorporated with transverse one for the shear force. As for the sequence of the shear. a closed rectangular as well as an oblique path are considered to simulate the actual behaviour due to the vibration. The shear stresses on the contact surface hetween fuel rod and spacer grid is evaluated numerically. It is supposed that a partial slip regime prevails on the contact at the early stage of fuel life. In case of gross slip. the present method can he applied without modification. The dissipation of friction energy on the contact is evaluated and discussed for a wear model and a spacer grid design. 00100782 Neutronics experiment on a mock-up of the ITER shielding blanket at the Frascati Neutron Generator Batistoni, P. et al. Fusion Engineering & Design, 1999, 47, (1). 25-60. An integral bulk shield experiment has been performed at the Frascati Neutron Generator (FNG) with the main objective to validate the ITER shielding system. To this end, a suitable mock-up of the ITER inboard shielding system including first wall, shielding blanket, vacuum vessel and toroidal field coil, had been assembled at ENEA Frascati. Neutron and ‘ray spectra as well as various nuclear responses were measured inside thts mock-up when irradiated hy 14.MeV neutrons. Measured and calculated data were compared to validate transport codes and nuclear data used in the design of the ITER system. In particular, the measured neutron and -,spectra, reaction rates and the nuclear heating were compared with the same quantities calculated with the MCNP Monte Carlo code using the Fusion Evaluated Nuclear Data Library (FENDL)-I.0 and the European Fusion File (EFF)-3.0 neutron cross-section data. The neutron-induced activation measured in stainless steel was compared with calculations performed with the FISPACT code and using the most recent activation cross-section data file FENDLIA-2.0. The paper includes a comparison of calculated over measured quantities (C/E) and a discussion of their relevance to the ITER nuclear design. It is found that, generally, the measured nuclear quantities are predicted by FENDL-I.0 and EFF-3.0 calculations within an uncertainty margin of about t 15% in the shielding blanket and vacuum vessel and ? 30% up to the region of the toroidal field coil. 00100783 Oblique injection system for ECH and ECCD in Heliotron E Manabe, Y. et al. Fusion Engineering & Design, 1999, 47, (I), 99-106. The oblique injection system of high power millimeter waves is designed and installed for plasma profile control and electron cyclotron current drive in the Heliotron E helical device. The previous 106.4 GHz electron cyclotron heating (ECH) perpendicular injection system has been modified by installing a movable mirror and a grating polarizer, which are required for toroidally oblique injection. Although the beam size at the resonant position is larger due to diffraction when scanning the beam direction, it is still smaller than the plasma radius so that the dependence of plasma profiles on the toroidal injection angle can be investigated experimentally. Low power and high power test results on the movable mirror and the polarizer have shown that the system works well as designed. The system has successfully been applied to the plasma experiment on Heliotron E. Plasma experimental results using this system are also shown.
Nuclear fuels (scienfific, fechnical)
energy ENDF/B-V cross-section\. MCNP with continuous energy ENDFIBVI cross-sections and KENO-V.a with 27-group ENDFIB-IV cross-sections. For one case. the codes overestimated k,,, by more than 1%. The experiments were grouped for three parameter\: plutonium mass. horizontal spacing and vertical spacing. For each parameter, a bias value was found. In all cases MCNP with ENDFIB-VI cross-sections had the smallest difference in the average k,tt. This same result was found when all experiments were grouped together. However. if the one experiment that was significantly overestimated was excluded, the smallest difference in the average k,tt was found for KFNO-V.a with ENDFiB-IV cross-sections. A fast-running fuel management program for a 00/00775 CANDU reactor Choi, H. Annul.7 of Ntccl. Ever&?. 21100. 27. ( I ), l-10. A fast-running fuel management programme for a CANDU reactor has been developed. The basic principle of this programme is to select refuelling channels. under circumstances whereby the reference reactor conditions are maintained by applying several constraints and criteria. The constraints used in this programme are the channel and bundle power and the fuel burn-up. The final selection of the re-fuelling channel is determined based on the priority of candidate channels, which enhances the reactor power distrihution close to the time-average model. Flooding in the pressurizer surge line of AP600 00100776 plant and analyses of APEX data Takeuchi, K. er ul. Ntrcl. Engineering & Design. 1999. 192, (1). 45-58. For the passive AP600 plant, the three stages of ADS (automatic depressurization system) valves are attached to the top of pressurizer. The existence of these valves makes liquid flow into and out of the pressurizer, an important part of the dynamics during a small break loss-ofcoolant accident. In this paper, counter-current flow limit (CCFL) in the surge line was analysed. Specifically. CCFL in vertical piping, in slightly inclined horizontal piping and in horizontal and vertical elbows were compared. The CCFL in the vertical section of the surge line was found to be the most limiting section. That is, the vertical CCFL controls the pressurizer liquid drain rate. This conclusion was tested and verified by comparing the predicted vertical CCFL against the counter-current flow states in the surge line. observed in small break LOCA tests conducted at the APhOO scaled test facility (APEX). Gas-liquid
two-phase
flow in narrow horizontal
Ekberg, N. P. et al. Ntrcl. t’~gi,rern’ng & Derign. 1999, I Y2, (I ), 59-80. Experimental data associated with the two-phase flow regimes, void fraction and pressure drop in horizontal, narrow, concentric annuli are presented. Two transparent test sections, one with inner and outer diameters of 6.6 and 8.6 mm, and an overall length of 46.0 cm; the other with 33.2 and 35.2 mm diameters and 43.0 cm length, respectively, were used. Near-atmospheric air and water constituted the gas and liquid phases. The gas and liquid superficial velocities were varied in the 0.02-57 and 0.1-6.1 m s-t ranges, respectively. The major two-phase flow patterns observed included bubbly, slug/plug, churn, stratified and annular. Transitional regimes, where the characteristics of two distinct flow regimes could be observed in the test sections, included bubbly-plug, stratified-slug and annular-slug. The obtained flow regime maps were different than flow regime maps typical of large horizontal channels and microchannels with circular cross-sections. They were also different from the flow regimes in rectangular thin channels. The measured average void fractions for the two test sections were compared with predictions of several empirical correlations. Overall, a correlation proposed by Butterworth [Butterworth, D., 1975. A comparison of some void fraction relationships for co-current gas-liquid flow. Inr. J. Multiphase Flow, 1, 845-X50] provided the most accurate prediction of the measured void fractions. The measured pressure drops were compared with predictions of several empirical correlations. Intelligent system for transient data collection and 00100778 fatigue monitoring of pressurised water reactors nuclear steam supply system Morilhat, P. er al. Ntccl. Engineering & Design, 1999, 192, (I), 103-58. Electricite de France (EDF), the French national electricity company, is operating S4 standardized pressurized water reactors. This about 500 reactor-years experience in nuclear stations operation and maintenance area has allowed EDF to develop its own strategy for monitoring of agerelated degradations of NPP systems and components relevant for plant safety and reliability. After more than IS years of experience in regulatory transient data collection and seven years of successful fatigue monitoring prototypes experimentation, EDF decided to design a new system called SYSFAC (acronym for SYsteme de Surveillance en FAtigue de la Chaudiere) devoted to transient logging and thermal fatigue monitoring of the reactor coolant pressure boundary. The system is fully automatic and directly connected to the on-site data acquisition network without any complementary instrumentation. A functional transient detection module and a mechanical transient detection module are in charge of the general transient data collection. A fatigue monitoring module is aimed towards a precise surveillance of five specific zones particularly sensitive to thermal fatigue. After a first step of preliminary studies, the industrial phase of the SYSFAC project is currently going on, with hardware and software tests
88
Fuel and Energy Abstracts
March 2000
and implementation. The first SYSFAC system wtll he delivered to the pilot power plant by the beginning of 1996. The extension to all EDF’s nuclear YOUMW is planned after one more year of feedback experience. International benchmark experiment in FCA-CEA 00/00779 results Chaussonet, P. er al. Pro~rers in .Nuc/. Energy, IY99, 35, (2). 157-162. To reduce uncertainties on the effective delayed neutron fraction, ,jc,t, an experimental international benchmark was performed at the Fast Critical Assembly (FCA) in JAERI. CEA participation was limited to the two configurations XIX-I and XIX-3 with uranium and plutonium fuel, which can be considered as complementary to the BERENICE benchmark in MASIJRCA. The results ohtained by the noise technique. showed that measured i,,, values are coherent, within the associated uncertainties, with the values obtained hy the other teams and that the target limits (.S?/, at 20) were respected. Measurement of effective delayed neutron fraction 00100780 Qef+by covariance-to-mean method for benchmark experiments of pstt at FCA Sakurni. T. ef al. Progress in Nucl. Enqy, 1999, 35, (2). 203-208. Measurements of the effective delayed neutron fraction $,t were carried out at the FCA facility of the Japan Atomic Energy Research Institute to contribute to benchmark experiment\ of ~$.,t. These measurements were made in two cores: XIX-I core fueled with 93% enriched uranium and XIX-3 core fueled with plutonium. The experimental jc,, was determined with a covariance-to-mean ratio of counts of a pair of neutron detectors installed in the core. The values of .I,,, were 724 ? I3 pcm and 252 % 5 pcm in the cores of XIX-I and XIX-3 respectively. 00100781 Mechanical analysis of fuel fretting problem Kim, H.-K. Nucl. Engirwering& Design, 1999. 192, (I). X1-93. Fuel fretting is studied by contact mechanics approach. Shear force produced by flow-induced vibration is regarded as the major factor of the fuel fretting. Contact dimension is examined for the Korean PWR Fuel Assembly using finite element method. Axial direction is incorporated with transverse one for the shear force. As for the sequence of the shear. a closed rectangular as well as an oblique path are considered to simulate the actual behaviour due to the vibration. The shear stresses on the contact surface hetween fuel rod and spacer grid is evaluated numerically. It is supposed that a partial slip regime prevails on the contact at the early stage of fuel life. In case of gross slip. the present method can he applied without modification. The dissipation of friction energy on the contact is evaluated and discussed for a wear model and a spacer grid design. 00100782 Neutronics experiment on a mock-up of the ITER shielding blanket at the Frascati Neutron Generator Batistoni, P. et al. Fusion Engineering & Design, 1999, 47, (1). 25-60. An integral bulk shield experiment has been performed at the Frascati Neutron Generator (FNG) with the main objective to validate the ITER shielding system. To this end, a suitable mock-up of the ITER inboard shielding system including first wall, shielding blanket, vacuum vessel and toroidal field coil, had been assembled at ENEA Frascati. Neutron and ‘ray spectra as well as various nuclear responses were measured inside thts mock-up when irradiated hy 14.MeV neutrons. Measured and calculated data were compared to validate transport codes and nuclear data used in the design of the ITER system. In particular, the measured neutron and -,spectra, reaction rates and the nuclear heating were compared with the same quantities calculated with the MCNP Monte Carlo code using the Fusion Evaluated Nuclear Data Library (FENDL)-I.0 and the European Fusion File (EFF)-3.0 neutron cross-section data. The neutron-induced activation measured in stainless steel was compared with calculations performed with the FISPACT code and using the most recent activation cross-section data file FENDLIA-2.0. The paper includes a comparison of calculated over measured quantities (C/E) and a discussion of their relevance to the ITER nuclear design. It is found that, generally, the measured nuclear quantities are predicted by FENDL-I.0 and EFF-3.0 calculations within an uncertainty margin of about t 15% in the shielding blanket and vacuum vessel and ? 30% up to the region of the toroidal field coil. 00100783 Oblique injection system for ECH and ECCD in Heliotron E Manabe, Y. et al. Fusion Engineering & Design, 1999, 47, (I), 99-106. The oblique injection system of high power millimeter waves is designed and installed for plasma profile control and electron cyclotron current drive in the Heliotron E helical device. The previous 106.4 GHz electron cyclotron heating (ECH) perpendicular injection system has been modified by installing a movable mirror and a grating polarizer, which are required for toroidally oblique injection. Although the beam size at the resonant position is larger due to diffraction when scanning the beam direction, it is still smaller than the plasma radius so that the dependence of plasma profiles on the toroidal injection angle can be investigated experimentally. Low power and high power test results on the movable mirror and the polarizer have shown that the system works well as designed. The system has successfully been applied to the plasma experiment on Heliotron E. Plasma experimental results using this system are also shown.