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0091 Investigation of efficient 131I production fromnatural uranium at Tehran research reactor
05 Nuclear fuels (scientific, technical) in the assembly was checked with activation foils. The mea s ure me nt results were analysed with the three-dimensional Monte Carlo code MCNP-4C and three different beryllium cross-section evaluations taken from FENDL/MC-2.0, FENDL/E-2.0, and EFF-3.03. The library FENDL/MC-2.0 was also used for the other materials in the assembly. All three beryllium libraries model the fast flux in the assembly well. Differences were seen for indium foils. An overestimation was observed for the tritium production.
06•00086 Hybrid soliton nuclear reactors: a model and simulation (encapsulated long living accelerator driven system) Gaveau, B. et al. Nuclear Engineering and Design, 2005, 235, (15), 1665 1674. The purpose of this study was to explore the potential of Hybrid Soliton Reactors (Rdacteur Hybride h Soliton, RHYS) for producing energy. In this case an encapsulated long-living fission reactor is driven by a proton accelerator, who produces neutrons on a target. In a first part of the paper the mathematical approach of such a sub-critical reactor is given, as an extension of the 'Soliton Reactor' recently proposed by other studies. In the second part results of simulations are given and the possibilities to control such a system explored.
06•00087 Identification and localization of absorbers of variable strength in nuclear reactors Demazi~re, C. and Andhill, G. Annals o/Nuclear Energy, 2005, 32, (8), 812 842. This paper investigates the possibility of localizing a noise source of the type 'absorber of variable strength' (or reactor oscillator) from as few as five neutron detectors evenly distributed throughout the core of a commercial nuclear reactor. The novelty of this investigation lies with the fact that the calculations are performed for a realistic 2-D heterogeneous reactor in the 2-group diffusion approximation, via the prior determination of the corresponding reactor transfer function. It is first demonstrated that the response of such a reactor to a localized perturbation deviates significantly from point-kinetics. The space-dependence of the induced neutron noise thus carries enough information about the location of the noise source, which makes it possible to determine its position from a few detector readings. The identification of the type of noise source is easily performed from the in-phase behaviour of the induced neutron noise. Different unfolding techniques are finally tested. All these techniques rely on the use of the reactor transfer function. One of these techniques is based on the comparison between the actual measured neutron noise and the neutron noise calculated for every possible location of the noise source. This technique is very reliable and almost insensitive to the contamination of the detector signals by background noise, but also extremely CPU consuming. Another technique, based on the piecewise inversion of the reactor transfer function and requiring little CPU effort, was developed. Although this technique is much less reliable when background noise is present, this technique is useful to indicate a region of the reactor where a noise source is likely to be located.
analyses based on first-order perturbation theory calculations have been performed using the deterministic code E R A N O S (Version 2.0) in conjunction with its adjusted nuclear data library ERALIB-1. It is found that the M4SC2 configuration, independent of the external source, is quite representative of the different XADSs for actinide capture reactions at the centre of the fuel zone, relative to 2 3 9 p u fission at the same location. For the case of a threshold fission reaction, such as that in 239U, the sensitivity to the external source is significantly higher. With respect to the corresponding spectral index, M4SC2 with the D(d,n)He 3 source remains quite representative of the He- and Nacooled XADSs. For the system with Pb/Bi coolant, on the other hand, effects of uncertainties associated with the data for these two nuclides and their low content in the MUSE configuration result in significantly lower associated representativity factors. A better overall representativity of the Pb/Bi-cooled XADS is expected to be achieved by the new MUSE_Na/Pb configuration.
06•00090 Improved core design of the high temperature supercritical-pressure light water reactor Yamaji, A. et al. Annals el'Nuclear Energy, 2005, 32, (7), 651 670. A new coolant flow scheme has been devised to raise the average coolant core outlet t e mpe ra t ure of the High Temperature Supercritical Pressure Light Water Reactor (S C LWR H). A new equilibrium core is designed with this flow scheme to show the feasibility of an SCLWR-H core with an average coolant core outlet temperature of 530°C. In previous studies, the average coolant core outlet temperature was limited by the relatively low t e mpe ra t ure outlet coolant from the core periphery. In order to achieve an average coolant core outlet t e mpe ra t ure of 500°C, each fuel assembly had to be horizontally divided into four sub-assemblies by coolant flow separation plates, and coolant flow rate had to be adjusted for each sub-assembly by an inlet orifice. However, the difficulty of raising the outlet coolant temperature from the core periphery remained. In this study, a new coolant flow scheme is devised, in which the fuel assemblies loaded on the core periphery are cooled by a descending flow. The new flow scheme has eliminated the need for raising the outlet coolant t e mpe ratu r e from the core periphery and removed the coolant flow separation plates from the fuel assemblies.
06•00091 Investigation of efficient 1311 production from natural uranium at Tehran research reactor Khalafi, H. et al. Annals o/Nuclear Energy, 2005, 32, (7), 729 740. Iodine-131, which has a half-life of 8.05 days, is the one of the most widely used radionuclides in medical diagnosis and treats some diseases of thyroid gland. Optimization of 1- 1I production in Tehran research reactor (TR R ) was studied by two different methods. Primarily, standard nuclear codes such as O R I G E N , WIMS and C I T A T I O N were applied and then analytical solutions technique was followed. Calculated results and experimental works in the bench scale
06•00088 Immobilization of radioactive evaporator concentrate in mortar matrix
irradiation in the unique production line.
Plecas, I. B. and Dimovic, S. Progress in Nuclear Energy, 2005, 46, (2), 151 157. Traditional methods of processing evaporator concentrates from NPP are evaporation and cementation. These methods allow the transformation of a liquid radioactive waste into an inert form, suitable for a final disposal. To assess the safety for disposal of radioactive mortar-waste composition, the leaching of - C s from immobilized radioactive evaporator concentrate into a surrounding fluid has been studied. Leaching tests were carried out in accordance with a method recommended by IAEA. Determination of retardation factors, KF, and coefficients of distribution, kd, using a simplified mathematical model for analysing the migration of radionuclides, has been developed. The experiment achieved the lowest leaching values after 60 days in samples. Results presented in this paper are examples of results obtained in a 20-year mortar and concrete testing project, which will influence the design of the engineered trenches system for a future central Serbian radioactive waste disposal centre.
06•00092 Measurement of (n, n') reaction cross-sections of 7gBr, 9°Zr, 197Au and 2°7pb with pulsed d-D neutrons
06/00089 Importance of the MUSE experiments for emerging ADS concepts from the nuclear data viewpoint Plaschy, M. et al. Anna/s o/Nuclear Energy, 2005, 32, (8), 843 856. The current investigation, conducted in the general framework of the MUSE program ('MUltiplication avec une Source Externe'), considers the representativity of a specific configuration of its fourth phase (M4SC2), which is driven by an external D(d,n)He 3 or T(d,n)He 4 neutron source, with respect to current concepts of eXperimental Accelerator Driven Systems (XADSs) with gas (He), Na and Pb/Bi coolants. The study has been carried out from the nuclear data viewpoint, with the external source being accounted for in an appropriate manner. In this context, data sensitivity/uncertainty
14
Fuel and Energy Abstracts
January 2006
Shimizu, T. et al. Annals of Nuclear Energy, 2005, 32, (9), 949 963. Activation cross-sections for the (n, n') reaction were m eas u r ed by means of the activation method at neutron energies of 3.1 and 2.54 MeV using a p u l s e d neutron beam. The target nuclei were 79 Br, 90 Zr, 197 Au, and 20TrPb whose half-lives were between 0.8 and 8 s. The value of the 9°Zr(n, n f) 9°mZr reaction was obtained for the first time. In order to confirm the pulsed neutron beam measuring method, the cross-section data of 79 Br and 197 Au were compared with previous data obtained using a pneumatic sample transport system. The results of this comparison were in agreement within the range of experimental error. The d-D neutrons were generated by bombarding a deuterated titanium target with a 350-keV d+-beam at the 80 ° beam line of the Fusion Neutronics Source (FNS) at the Japan Atomic Energy Research Institute. In order to obtain reliable activation cross-sections, careful attention was paid to correct the efficiency for a volume source, and the self-absorption of gamma rays in irradiated samples. The systematics of the (n, n') reaction at a neutron energy of 3.0 MeV, which can predict cross-section of (n, n') reaction with an accuracy of 50%, was proposed for the first time on the basis of the data.
06/00093 Modeling of flashing-induced instabilities in the start-up phase of natural-circulation BWRs using the twophase flow code FLOCAL Manera, A. et al. Nuclear Engineering and Design, 2005, 235, (14), 1517 1535. This paper reports on the modelling and simulation of flashing-induced instabilities in natural-circulation systems, with special emphasis on natural-circulation boiling water reactors (BWRs). For the modelling
06•00086 Hybrid soliton nuclear reactors: a model and simulation (encapsulated long living accelerator driven system) Gaveau, B. et al. Nuclear Engineering and Design, 2005, 235, (15), 1665 1674. The purpose of this study was to explore the potential of Hybrid Soliton Reactors (Rdacteur Hybride h Soliton, RHYS) for producing energy. In this case an encapsulated long-living fission reactor is driven by a proton accelerator, who produces neutrons on a target. In a first part of the paper the mathematical approach of such a sub-critical reactor is given, as an extension of the 'Soliton Reactor' recently proposed by other studies. In the second part results of simulations are given and the possibilities to control such a system explored.
06•00087 Identification and localization of absorbers of variable strength in nuclear reactors Demazi~re, C. and Andhill, G. Annals o/Nuclear Energy, 2005, 32, (8), 812 842. This paper investigates the possibility of localizing a noise source of the type 'absorber of variable strength' (or reactor oscillator) from as few as five neutron detectors evenly distributed throughout the core of a commercial nuclear reactor. The novelty of this investigation lies with the fact that the calculations are performed for a realistic 2-D heterogeneous reactor in the 2-group diffusion approximation, via the prior determination of the corresponding reactor transfer function. It is first demonstrated that the response of such a reactor to a localized perturbation deviates significantly from point-kinetics. The space-dependence of the induced neutron noise thus carries enough information about the location of the noise source, which makes it possible to determine its position from a few detector readings. The identification of the type of noise source is easily performed from the in-phase behaviour of the induced neutron noise. Different unfolding techniques are finally tested. All these techniques rely on the use of the reactor transfer function. One of these techniques is based on the comparison between the actual measured neutron noise and the neutron noise calculated for every possible location of the noise source. This technique is very reliable and almost insensitive to the contamination of the detector signals by background noise, but also extremely CPU consuming. Another technique, based on the piecewise inversion of the reactor transfer function and requiring little CPU effort, was developed. Although this technique is much less reliable when background noise is present, this technique is useful to indicate a region of the reactor where a noise source is likely to be located.
analyses based on first-order perturbation theory calculations have been performed using the deterministic code E R A N O S (Version 2.0) in conjunction with its adjusted nuclear data library ERALIB-1. It is found that the M4SC2 configuration, independent of the external source, is quite representative of the different XADSs for actinide capture reactions at the centre of the fuel zone, relative to 2 3 9 p u fission at the same location. For the case of a threshold fission reaction, such as that in 239U, the sensitivity to the external source is significantly higher. With respect to the corresponding spectral index, M4SC2 with the D(d,n)He 3 source remains quite representative of the He- and Nacooled XADSs. For the system with Pb/Bi coolant, on the other hand, effects of uncertainties associated with the data for these two nuclides and their low content in the MUSE configuration result in significantly lower associated representativity factors. A better overall representativity of the Pb/Bi-cooled XADS is expected to be achieved by the new MUSE_Na/Pb configuration.
06•00090 Improved core design of the high temperature supercritical-pressure light water reactor Yamaji, A. et al. Annals el'Nuclear Energy, 2005, 32, (7), 651 670. A new coolant flow scheme has been devised to raise the average coolant core outlet t e mpe ra t ure of the High Temperature Supercritical Pressure Light Water Reactor (S C LWR H). A new equilibrium core is designed with this flow scheme to show the feasibility of an SCLWR-H core with an average coolant core outlet temperature of 530°C. In previous studies, the average coolant core outlet temperature was limited by the relatively low t e mpe ra t ure outlet coolant from the core periphery. In order to achieve an average coolant core outlet t e mpe ra t ure of 500°C, each fuel assembly had to be horizontally divided into four sub-assemblies by coolant flow separation plates, and coolant flow rate had to be adjusted for each sub-assembly by an inlet orifice. However, the difficulty of raising the outlet coolant temperature from the core periphery remained. In this study, a new coolant flow scheme is devised, in which the fuel assemblies loaded on the core periphery are cooled by a descending flow. The new flow scheme has eliminated the need for raising the outlet coolant t e mpe ratu r e from the core periphery and removed the coolant flow separation plates from the fuel assemblies.
06•00091 Investigation of efficient 1311 production from natural uranium at Tehran research reactor Khalafi, H. et al. Annals o/Nuclear Energy, 2005, 32, (7), 729 740. Iodine-131, which has a half-life of 8.05 days, is the one of the most widely used radionuclides in medical diagnosis and treats some diseases of thyroid gland. Optimization of 1- 1I production in Tehran research reactor (TR R ) was studied by two different methods. Primarily, standard nuclear codes such as O R I G E N , WIMS and C I T A T I O N were applied and then analytical solutions technique was followed. Calculated results and experimental works in the bench scale
06•00088 Immobilization of radioactive evaporator concentrate in mortar matrix
irradiation in the unique production line.
Plecas, I. B. and Dimovic, S. Progress in Nuclear Energy, 2005, 46, (2), 151 157. Traditional methods of processing evaporator concentrates from NPP are evaporation and cementation. These methods allow the transformation of a liquid radioactive waste into an inert form, suitable for a final disposal. To assess the safety for disposal of radioactive mortar-waste composition, the leaching of - C s from immobilized radioactive evaporator concentrate into a surrounding fluid has been studied. Leaching tests were carried out in accordance with a method recommended by IAEA. Determination of retardation factors, KF, and coefficients of distribution, kd, using a simplified mathematical model for analysing the migration of radionuclides, has been developed. The experiment achieved the lowest leaching values after 60 days in samples. Results presented in this paper are examples of results obtained in a 20-year mortar and concrete testing project, which will influence the design of the engineered trenches system for a future central Serbian radioactive waste disposal centre.
06•00092 Measurement of (n, n') reaction cross-sections of 7gBr, 9°Zr, 197Au and 2°7pb with pulsed d-D neutrons
06/00089 Importance of the MUSE experiments for emerging ADS concepts from the nuclear data viewpoint Plaschy, M. et al. Anna/s o/Nuclear Energy, 2005, 32, (8), 843 856. The current investigation, conducted in the general framework of the MUSE program ('MUltiplication avec une Source Externe'), considers the representativity of a specific configuration of its fourth phase (M4SC2), which is driven by an external D(d,n)He 3 or T(d,n)He 4 neutron source, with respect to current concepts of eXperimental Accelerator Driven Systems (XADSs) with gas (He), Na and Pb/Bi coolants. The study has been carried out from the nuclear data viewpoint, with the external source being accounted for in an appropriate manner. In this context, data sensitivity/uncertainty
14
Fuel and Energy Abstracts
January 2006
Shimizu, T. et al. Annals of Nuclear Energy, 2005, 32, (9), 949 963. Activation cross-sections for the (n, n') reaction were m eas u r ed by means of the activation method at neutron energies of 3.1 and 2.54 MeV using a p u l s e d neutron beam. The target nuclei were 79 Br, 90 Zr, 197 Au, and 20TrPb whose half-lives were between 0.8 and 8 s. The value of the 9°Zr(n, n f) 9°mZr reaction was obtained for the first time. In order to confirm the pulsed neutron beam measuring method, the cross-section data of 79 Br and 197 Au were compared with previous data obtained using a pneumatic sample transport system. The results of this comparison were in agreement within the range of experimental error. The d-D neutrons were generated by bombarding a deuterated titanium target with a 350-keV d+-beam at the 80 ° beam line of the Fusion Neutronics Source (FNS) at the Japan Atomic Energy Research Institute. In order to obtain reliable activation cross-sections, careful attention was paid to correct the efficiency for a volume source, and the self-absorption of gamma rays in irradiated samples. The systematics of the (n, n') reaction at a neutron energy of 3.0 MeV, which can predict cross-section of (n, n') reaction with an accuracy of 50%, was proposed for the first time on the basis of the data.
06/00093 Modeling of flashing-induced instabilities in the start-up phase of natural-circulation BWRs using the twophase flow code FLOCAL Manera, A. et al. Nuclear Engineering and Design, 2005, 235, (14), 1517 1535. This paper reports on the modelling and simulation of flashing-induced instabilities in natural-circulation systems, with special emphasis on natural-circulation boiling water reactors (BWRs). For the modelling