- Email: [email protected]
00356 FAMSAP: A code to analyze the dynamic response of PWR fuel assemblies in LOCA and seismic conditions and its application to QNPP
05
Dose control at nuclear power stations 96lQO346 NCRP Report No.124 National Council on Radiation Protection & Measurements, Maryland, USA, $20.00, 138 pp. The recommendations in this report are based on practical techniques and
programmes that have been generally successful in controlling and reducmg exposure to workers. Effect of detector size and posltlon on measured 96lQO347 vlbratlon spectra of strings and rods Lipcsei, S. et al., Progress in Nuclear Energy, 1994, 28, (2), 91-102. The effect of detector size and position on the measured spectra has been
investigated and the main characteristics of the transfer function were calculated by using a simple model of standing waves. The theoretical estimation was checked in laboratory rod vibration experiments, and the main features were also found in pressure fluctuation spectra measured at Paks Nuclear Power Station. Fundamental results have been achieved using the theory of the superposition of running waves and their reflection on the clamped ends of the rod. Effects of cryogenic lrradlatlon on electrlcal 96lQQ346 strength of candldate ITER lnsulatlon materlals Schutz, J. B. et al., Cryogenics, 1995, 35, (ll), 759-762. The electrical strength of insulation systems is critical to the success of
complex fusion reactor systems such as the International Thermonuclear Experimental Reactor (ITER). Candidate toroidal field coil insulation materials were subjected to irradiation at 4 K by a fast neutron fluence of 3.1 X lO**nrn-*. Electrical strengths of irradiated materials were evaluated at 76 K and are compared to electrical strengths of unirradiated control specimens. Materials evaluated include several classes of organic matrix insulation, barrier materials including mica paper and polyimide film, and selected hybrid insulation systems. Effects of neutron/gamma lrradlatlon 96100349 shear and compressive propertles of lnsulatlon Reed, R. et al., Cryogenics, 1995, 35, (ll), 739-741.
at 4 K on
The effects of irradiation in a fission reactor at 4 K on the shear and compressive properties of 4 K of candidate insulation systems for the International Thermonuclear Experimental Reactor have been examined. The shear strength was measured by the short-beam shear test technique. Fast neutron fluences from 0.9 to 3 X 102*nrn-* were obtained at 4 K. Results indicate that the shear strength and flexural modulus were both sensitive to the neutron fluence; the compressive strength and modulus were less sensitive. Systems of flexibilized diglycidyl ether of bisphenol-A epoxy were more sensitive to irradiation than systems based on tetraglycidyl diaminodiphenyl methane epoxy. 96100350
EPR: The regulators’ vlew
Kovan, D. Nuclear Engng. Int., Nov. 1995, 40, (496), 44-45. When the French and German partners in the EPR (European Pressurizedwater Reactor) project agreed to invest another $150 million to complete the basic design phase, it was after receiving an ‘OK’ from both national safety authorities on the project’s approach to safety. The authorities want the reactors to have substantially reduced levels of risk and require the main safety features, particularly those concerning severe accidents, to be incorporated early in the design phase. Estlmate of total control rod worth based on lower 96100351 dlmenslonal modelllng Sathyabama, N. et al., Ann. Nucl. Energy, Jan. 1995, 22, (l), 61-66.
Estimation of control rod worths by three dimensional methods is very time consuming. Even with the powerful computers, available today, it is difficult to perform multigroup (25 groups and above) calculations in three dimensions for large fast power reactors. In this paper, a simplified theoretical formula for estimating the total control rod worth from its lower dimensional worth estimattons has been framed. The control rod worth estimated in 2-D cylindrical geometry modelling has the cylindricity error and that estimated in 2-D hex/X-Y geometry modelling has the buckling error and the worth estimated in 1-D cylindrical geometry modelling has both the cylindricity and the buckling errors. 96lOO352 Evaluatlon of cold neutron scattering cross sectlons for light and heavy water Morishima, N. and Aoki, Y. Ann. Nucl. Energy, Mar.-Apr. 1995, 22, (3),
147-158. Cross section models for cold and thermal neutron scattering from light and heavy water are developed. They are based on the proper treatment of molecular motions in the water in terms of the diffusive and oscillatory translational motions of a water molecule and the intermolecular spacetime correlations for highly bonded molecules. Intramolecular motions such as hindered rotations and internal harmonic vibrations are also included. The total and differential scattering cross sections are calculated for a wide range of incident neutron energies from very cold to epithermal. Good agreement with available experimental results permits us to generate basic cross section libraries for the evaluation of cold and ultracold neutron production.
Nuclear fuels (scientific, technical)
96100353 transport
Evaluatlon of hl her K-elgnevalues of the neutron equatlon by S,-met % od Modak, R. S. et al., Ann. Nucl. Energy, Jun. 1995, 22, (6), 359-366.
Many semi-analytical methods are being developed to evaluate higher eigenvalue of monoenergetic neutron transport equation. In this apar, a numerical approach is presented where the well-known S,,-metho8, IS used to generate higher K-eigenvalues of neutron transport equation via the generation of a fission matrix. Although less accurate than the semi-analytical methods, the inclusion of spatial inhomogeneities, scattering anisotropies and even more energy groups would be straightforward in this method owing to the versatility of the &-codes. It also shows that the Keigenvalues are always real in the mono-energetic case even if scattering is anisotropic. Exergy analysis of an operatlng bolllng-water-reac96100354 tor nuclear power statlon L$blagIi, W. R. et al., Energy Convers. Mgmt., Mar. 1995, 36, (3),
A Second Law analysis is performed on the LaSalle County Nuclear Station of the Commonwealth Edison Company to evaluate plant and subsystem irreversibility. The results disclose that over 80% of the exergy destroyed during plant operation is a result of the highly irreversible fission and heat transport processes within the reactor vessel. Plant efficiency and effectiveness are found to be 34.4%, well below the 40-45% efficiencies of typical fossil-fuel-fired power generating stations. Based on these wellknown numbers, and the results of the exergy analysis, one recommendation is to reevaluate the integration of fossil-fuel-fired superheat/reheat units located downstream of the reactor vessel. 96100355 An extended discharge burnup optlmlzstlon technlque using Penn State’s fuel management package and CASMO-3/SIMULATE-3 Feltus, M. A. Ann. Nucl. Energy, May 1995, 22, (5), 267-274.
The paper addresses the advantages and disadvantages of using very high fuel burnup, reinsertion, and low leakage designs in advanced fuel cycle light water reactor cores as a technique to reduce vessel fluence, and total volume of spent fuel discharged into the waste management stream. 96100356 FAMSAP: A code to analyze the dynamic response of PWR fuel assemblies In LOCA and selsmlc condltlons and It8 appllcatlon to QNPP Chongzhu, Z. and Zhongyue, Z. Progress in Nuclear Energy, 1994, 28, (2), 170-176.
FAMSAP is used to calculate the non-linear lateral transient response of PWR fuel assemblies to seismic or LOCA excitation. It is mainly composed of three subcodes, FAMREG, FAFRES and PLOT, two of which are for physics calculation and another one is for plotting purposes. FAMREC (Fuel Assembly Mechanical Response Code) is mamly used to calculate fuel assembly displacements, impacts and spacer grid crushing forces response. Its input data include fuel assembly mode shapes matrix, circular frequency matrix and some static properties obtained from fuel assemblies prototype test. FAFRES is called to calculate dynamic behaviour matrics and static behaviour matrix if prototype testing cannot be performed. PLOT code can plot input acceleration of the core plates, assembly displacement response, impact forces, spacer grid crushing forces and stresses on the worst deformed assembly. 96lOO357 Feaslblllty of zero temperature coefflclent core with hlghly diluted fuel and graphlte moderator Ann. Nucl. Energy, Jun. 1995, 22, (6), C),4ayj7T and Sekimoto, H.
The feasibility of the special core of which temperature coefficient is almost zero over wide temperature range and becomes negative at very high temperature is shown in the present study. To avoid the effect of the power density distribution which causes temperature distribution in the core, homogenous core is assumed in this study. The graphite is adopted as the moderator from its small neutron absorption. The fuel is highly diluted to make the thermal neutron flux peak high and to make epithermal and fast neutron flux low. If the plutonium and Er-167 are mixed in this core properly, it is possible to make the core temperature coefficient almost constant between 300 and 900 K and to make it suddenly negative above 900 K. Fuel assembly ldentlflcatlon - ANSVANS-57.64995 96lOO356 American Nuclear Society, 555 North Kensington Ave., La Grange Park, IL.60526, USA, $20.00 (Order No.240205), 1995.
Revision of ANSUANS-57.8-1978. This standard describes requirements for the unique identification of fuel assemblies utilized in nuclear power plants. It defines the characters and proposed sequence to be used in assigning identification to fuel assembles. 96/00359 Future human actlons at dlsposal sites OECD Nuclear Energy Agency, Paris, France, 70 pp.
The report by a Nuclear Engineering Agency working group discusses the treatment of risk in postclosure safety assessments of some deep geological waste repositories which may arise from future human actions.
Fuel and Energy Abstracts
January
1996
21
Dose control at nuclear power stations 96lQO346 NCRP Report No.124 National Council on Radiation Protection & Measurements, Maryland, USA, $20.00, 138 pp. The recommendations in this report are based on practical techniques and
programmes that have been generally successful in controlling and reducmg exposure to workers. Effect of detector size and posltlon on measured 96lQO347 vlbratlon spectra of strings and rods Lipcsei, S. et al., Progress in Nuclear Energy, 1994, 28, (2), 91-102. The effect of detector size and position on the measured spectra has been
investigated and the main characteristics of the transfer function were calculated by using a simple model of standing waves. The theoretical estimation was checked in laboratory rod vibration experiments, and the main features were also found in pressure fluctuation spectra measured at Paks Nuclear Power Station. Fundamental results have been achieved using the theory of the superposition of running waves and their reflection on the clamped ends of the rod. Effects of cryogenic lrradlatlon on electrlcal 96lQQ346 strength of candldate ITER lnsulatlon materlals Schutz, J. B. et al., Cryogenics, 1995, 35, (ll), 759-762. The electrical strength of insulation systems is critical to the success of
complex fusion reactor systems such as the International Thermonuclear Experimental Reactor (ITER). Candidate toroidal field coil insulation materials were subjected to irradiation at 4 K by a fast neutron fluence of 3.1 X lO**nrn-*. Electrical strengths of irradiated materials were evaluated at 76 K and are compared to electrical strengths of unirradiated control specimens. Materials evaluated include several classes of organic matrix insulation, barrier materials including mica paper and polyimide film, and selected hybrid insulation systems. Effects of neutron/gamma lrradlatlon 96100349 shear and compressive propertles of lnsulatlon Reed, R. et al., Cryogenics, 1995, 35, (ll), 739-741.
at 4 K on
The effects of irradiation in a fission reactor at 4 K on the shear and compressive properties of 4 K of candidate insulation systems for the International Thermonuclear Experimental Reactor have been examined. The shear strength was measured by the short-beam shear test technique. Fast neutron fluences from 0.9 to 3 X 102*nrn-* were obtained at 4 K. Results indicate that the shear strength and flexural modulus were both sensitive to the neutron fluence; the compressive strength and modulus were less sensitive. Systems of flexibilized diglycidyl ether of bisphenol-A epoxy were more sensitive to irradiation than systems based on tetraglycidyl diaminodiphenyl methane epoxy. 96100350
EPR: The regulators’ vlew
Kovan, D. Nuclear Engng. Int., Nov. 1995, 40, (496), 44-45. When the French and German partners in the EPR (European Pressurizedwater Reactor) project agreed to invest another $150 million to complete the basic design phase, it was after receiving an ‘OK’ from both national safety authorities on the project’s approach to safety. The authorities want the reactors to have substantially reduced levels of risk and require the main safety features, particularly those concerning severe accidents, to be incorporated early in the design phase. Estlmate of total control rod worth based on lower 96100351 dlmenslonal modelllng Sathyabama, N. et al., Ann. Nucl. Energy, Jan. 1995, 22, (l), 61-66.
Estimation of control rod worths by three dimensional methods is very time consuming. Even with the powerful computers, available today, it is difficult to perform multigroup (25 groups and above) calculations in three dimensions for large fast power reactors. In this paper, a simplified theoretical formula for estimating the total control rod worth from its lower dimensional worth estimattons has been framed. The control rod worth estimated in 2-D cylindrical geometry modelling has the cylindricity error and that estimated in 2-D hex/X-Y geometry modelling has the buckling error and the worth estimated in 1-D cylindrical geometry modelling has both the cylindricity and the buckling errors. 96lOO352 Evaluatlon of cold neutron scattering cross sectlons for light and heavy water Morishima, N. and Aoki, Y. Ann. Nucl. Energy, Mar.-Apr. 1995, 22, (3),
147-158. Cross section models for cold and thermal neutron scattering from light and heavy water are developed. They are based on the proper treatment of molecular motions in the water in terms of the diffusive and oscillatory translational motions of a water molecule and the intermolecular spacetime correlations for highly bonded molecules. Intramolecular motions such as hindered rotations and internal harmonic vibrations are also included. The total and differential scattering cross sections are calculated for a wide range of incident neutron energies from very cold to epithermal. Good agreement with available experimental results permits us to generate basic cross section libraries for the evaluation of cold and ultracold neutron production.
Nuclear fuels (scientific, technical)
96100353 transport
Evaluatlon of hl her K-elgnevalues of the neutron equatlon by S,-met % od Modak, R. S. et al., Ann. Nucl. Energy, Jun. 1995, 22, (6), 359-366.
Many semi-analytical methods are being developed to evaluate higher eigenvalue of monoenergetic neutron transport equation. In this apar, a numerical approach is presented where the well-known S,,-metho8, IS used to generate higher K-eigenvalues of neutron transport equation via the generation of a fission matrix. Although less accurate than the semi-analytical methods, the inclusion of spatial inhomogeneities, scattering anisotropies and even more energy groups would be straightforward in this method owing to the versatility of the &-codes. It also shows that the Keigenvalues are always real in the mono-energetic case even if scattering is anisotropic. Exergy analysis of an operatlng bolllng-water-reac96100354 tor nuclear power statlon L$blagIi, W. R. et al., Energy Convers. Mgmt., Mar. 1995, 36, (3),
A Second Law analysis is performed on the LaSalle County Nuclear Station of the Commonwealth Edison Company to evaluate plant and subsystem irreversibility. The results disclose that over 80% of the exergy destroyed during plant operation is a result of the highly irreversible fission and heat transport processes within the reactor vessel. Plant efficiency and effectiveness are found to be 34.4%, well below the 40-45% efficiencies of typical fossil-fuel-fired power generating stations. Based on these wellknown numbers, and the results of the exergy analysis, one recommendation is to reevaluate the integration of fossil-fuel-fired superheat/reheat units located downstream of the reactor vessel. 96100355 An extended discharge burnup optlmlzstlon technlque using Penn State’s fuel management package and CASMO-3/SIMULATE-3 Feltus, M. A. Ann. Nucl. Energy, May 1995, 22, (5), 267-274.
The paper addresses the advantages and disadvantages of using very high fuel burnup, reinsertion, and low leakage designs in advanced fuel cycle light water reactor cores as a technique to reduce vessel fluence, and total volume of spent fuel discharged into the waste management stream. 96100356 FAMSAP: A code to analyze the dynamic response of PWR fuel assemblies In LOCA and selsmlc condltlons and It8 appllcatlon to QNPP Chongzhu, Z. and Zhongyue, Z. Progress in Nuclear Energy, 1994, 28, (2), 170-176.
FAMSAP is used to calculate the non-linear lateral transient response of PWR fuel assemblies to seismic or LOCA excitation. It is mainly composed of three subcodes, FAMREG, FAFRES and PLOT, two of which are for physics calculation and another one is for plotting purposes. FAMREC (Fuel Assembly Mechanical Response Code) is mamly used to calculate fuel assembly displacements, impacts and spacer grid crushing forces response. Its input data include fuel assembly mode shapes matrix, circular frequency matrix and some static properties obtained from fuel assemblies prototype test. FAFRES is called to calculate dynamic behaviour matrics and static behaviour matrix if prototype testing cannot be performed. PLOT code can plot input acceleration of the core plates, assembly displacement response, impact forces, spacer grid crushing forces and stresses on the worst deformed assembly. 96lOO357 Feaslblllty of zero temperature coefflclent core with hlghly diluted fuel and graphlte moderator Ann. Nucl. Energy, Jun. 1995, 22, (6), C),4ayj7T and Sekimoto, H.
The feasibility of the special core of which temperature coefficient is almost zero over wide temperature range and becomes negative at very high temperature is shown in the present study. To avoid the effect of the power density distribution which causes temperature distribution in the core, homogenous core is assumed in this study. The graphite is adopted as the moderator from its small neutron absorption. The fuel is highly diluted to make the thermal neutron flux peak high and to make epithermal and fast neutron flux low. If the plutonium and Er-167 are mixed in this core properly, it is possible to make the core temperature coefficient almost constant between 300 and 900 K and to make it suddenly negative above 900 K. Fuel assembly ldentlflcatlon - ANSVANS-57.64995 96lOO356 American Nuclear Society, 555 North Kensington Ave., La Grange Park, IL.60526, USA, $20.00 (Order No.240205), 1995.
Revision of ANSUANS-57.8-1978. This standard describes requirements for the unique identification of fuel assemblies utilized in nuclear power plants. It defines the characters and proposed sequence to be used in assigning identification to fuel assembles. 96/00359 Future human actlons at dlsposal sites OECD Nuclear Energy Agency, Paris, France, 70 pp.
The report by a Nuclear Engineering Agency working group discusses the treatment of risk in postclosure safety assessments of some deep geological waste repositories which may arise from future human actions.
Fuel and Energy Abstracts
January
1996
21