- Email: [email protected]
00093 Calorimetric measurements on uraniumplutonium mixed oxides
05 Nuclear fuels (scientific, technical) and hence, the specific thermal power of the CPFC depend on the estimated maximum fuel temperature during re-entry following a launch accident, the storage time before launch and mission duration, and the helium release fraction from the fuel kernels. This p a p e r investigates, the advantage of replacing the four iridium-clad 23~PuO 2 fuel pellets, the two floating graphite membranes, and the two graphite impact shells in current State-Of-the-Art (SOA) GPHS with CPFC. The mass, thermal power, and specific thermal power of the CPFCGPHS are calculated as functions of the helium release fraction from the fuel kernels, up to a full release, and the maximum compact temperature during re-entry, from 1500 K to 2400 K. For the same mass and volume, the single-size particles CPFC-GPHS could generate 260 W at Beginning-Of-Mission (BOM), following 10 years of storage, versus 231 W for SOA GPHS. For an additional 10% higher mass, the CPFC-GPHS could generate 340 W at BOM at a specific thermal power of 214 W/kg, which are 47% and 34% higher than those of SOA GPHS, respectively.
05/00090 Analysis of a PWR core baffle considering irradiation induced creep Altstadt, E. et al. Annals of Nuclear Energy, 2004, 31, (7), 723 736. The core baffle of a PWR is loaded by the pressure difference between bypass and core and by temperature profiles developing from gamma and neutron heating and heat transfer into the coolant. Strain, deformation and gaps between the sheets resulting from this load are determined considering the effect of neutron irradiation induced creep of the core baffle bolts. The finite element code ANSYS C'~is applied for the thermal and mechanical analyses. The FE-model comprises a complete 45 ° sector of the core baffle structure including the core barrel, the formers, the core baffle sheets and about 230 bolt connections with non-linear contact between the single components and the effect of friction. The complete analysis requires three major steps. (1) Evaluation of the three-dimensional distribution of neutron flux an d gamma induced internal heating with the Monte Carlo code MCNP '~. These calculations are based on pin wise power distributions at the core edge for typical loading patterns. (2) Calculation of the temperature distribution in the core baffle for different operational conditions and core loading patterns, considering heat conduction in the components with internal heat sources and convectional boundary conditions (heat transfer coefficients and bulk temperature of the coolant). (3) Calculation of time dependent deformation, stresses and strains taking into account weight, pressure loads, temperature fields for different load situations, pre-stressing, irradiation induced creep of the bolts as correlated to neutron flux. The results show the equalizing effect of redistribution of bolt loads from high flux to lower flux exposure locations in a self controlled process, keeping the mechanical and geometrical stability of the core baffle structure and leaving the gaps between sheet edges unaffected.
05/00091 Application of artificial neural network for safety core parameters prediction in LWRRS Mazrou, H. and Hamadouche, M. Progress in Nuclear Energy, 2004, 44, (3), 263-275. This paper reports on the use of Artificial Neural Networks (ANNs) in predicting core safety parameters, such as multiplication factor Keff and fuel powers peaks P, .... in Light Water Research Reactors (LWRRs) using a personal computer. The purpose of this emulation, when it is associated with optimal incore fuel management, is to allow rapid and extensive exploration of potential good configurations of fuel in L W R R s cores. The key idea of the developed program lies in the use of an adaptive learning rate procedure in a typical back-propagation error model. The performance of the algorithm was improved to the proper setting of the learning rate. Which is adjusted according to the minimal error predicted. Using the 2-Dimensional neutronic diffusion code MUDICO-2D for the referenced target values, the neural network development process is reported for a slight modified version of the IAEA 10Mw benchmark LEU core.
05/00092 Atmospheric dispersion of argon-41 from a nuclear research reactor: measurement and modelling of plume geometry and gamma radiation field, Lauritzen, B. et al. International Journal of' Environment and Pollution, 2003, 20, (1-6), 47 54. An atmospheric dispersion experiment was conducted using a visible tracer along with the routine releases of 41Ar from the BR1 research reactor in Mol, Belgium. Simultaneous measurements of plume geometry and radiation field form 41Ar decay were performed as well as measurements of the 41Ar source term and the meteorological parameters. Good overall agreement is found between measurement data and model results using the mesoscale atmospheric dispersion and dose rate model RIMPUFF.
14
Fuel and Energy Abstracts
January 2005
05/00093 Calorimetric measurements on uraniumplutonium mixed oxides Kandan, R. et al. Journal of Nuclear Materials, 2004, 324, (2-3), 215 219. Enthalpy increments of Ul_yPuyO2 solid solutions with y = 0.21, 0.28 and 0.40 were measured using a high-temperature differential calorimeter by employing the method of inverse drop calorimetry in the temperature range 1000-1780 K. From the fit equations for the enthalpy increments, other thermodynamic functions such as heat capacity, entropy and Gibbs energy function have been computed in the temperature range 298-1800 K. The results indicate that the enthalpies of (U,Pu)O2 solid solutions obey the N e u m a n n - K o p p molar additivity rule.
05/00094 Development of a 30 m long NbaAI superconducting conductor jacketed with a round-in-square type stainless steel conduit for the toroidal field coils in JT-60SC Tsuchiya, K. et al. Fusion Engineering and Design, 2004, 70, (2), 131 140. The modification of JT-60 was programmed to be a superconducting tokamak (JT-60SC) that has superconducting toroidal field (TF) and poloidal field coils. The Nb3AI superconductor was developed in order to apply to the TF coil system in JT-60SC. The TF coil conductor was designed with a cable-in-conduit (CIC) type. This conductor was Nb3AI cable jacketed into stainless steel (SS) conduit. As the first step of SS conduit production, the 10 m long round-in-square type tubes were successfully fabricated within eccentricity tolerance of the tube hole (<10%). A 30 m SS conduit was made of three units of 10 m long tubes, which were butt-welded in series. After the welding process, penetrating beads were observed by CCD camera, and it was visually clear that the penetrating beads were successfully controlled by adjustment of back-shielding gas pressure. Therefore, regarding the production of SS conduit, the procedures of fabrication and inspection were established. Then, the inserting of cable into this conduit and compaction were also carried out. Consequently, the 30 m Nb3A1 CIC conductor was successfully fabricated, so that a feasibility of massproducing CIC conductor applied to the real machine was obtained.
05/00095 Development of environment for remote participation in fusion research on JT-60 Oshima, T. et al. Fusion Engineering and Design, 2004, 71, (1-4), 239 244. In the JT-60 tokamak at JAERI, an environment for remote participation is to be developed by concentrating ideas of experts of nuclear fusion research from research organizations and universities distributed throughout the country. The study involves constructing a hierarchical collaborative remote research system, which consists of remote experiment, remote analysis and remote diagnostics. In remote collaboration, it is important to share information, atmosphere and presence between the participants as well as to maintain the security of the system. To implement a collaborative environment, a video conferencing system, a video streaming system and remote analysis system called VizAnalysis have been developed for the communication. To assist the remote participation for the fusion research, the document management system based on the web browser called VizSquare has been developed. In JT-60, security and authentication methods utilizing a computer network and a new communication tool have been developed, and they are applicable to the remote participation of ITER.
05/00096 Developmental prototype for replacement of JT-60 timing system Akasaka, H. et al. Fusion Engineering and Design, 2004, 71, (1-4), 2934. The present CAMAC-based timing system has been used for synchronizing sequential events of the discharge and the data collection of the interesting JT-60U experiment plasma phenomena. However, a more flexible and sophisticated state-of-the-art timing system now is required to realize advanced plasma control with minimal maintenance costs. In this context, the versa module Europe system with a high-speed data communication network using reflective memory (RM) modules and user-friendly application software based on MATLAB TM tools has been selected to develop the new prototype timing system. In the ZENKEI, the supervisory control system of the JT-60, the supervisory timing system provides the 50-~ts master clock pulses, the various timing signal preparation logic, which is built into the digital signal processing module in conjunction with the discharge sequence event signals, and the 6.2 MB/s high-speed communication data link provided by the RM module. Except the clock pulse generator module, no other special timing module is necessary for this new timing system. The timing signal is prepared by software logic in conjunction with sequential events and the preset timer, is transferred to the subsystems through the RM module, where it is synchronized to the 50~ts clock pulses. The timing system of the subsystems also consists of
05/00090 Analysis of a PWR core baffle considering irradiation induced creep Altstadt, E. et al. Annals of Nuclear Energy, 2004, 31, (7), 723 736. The core baffle of a PWR is loaded by the pressure difference between bypass and core and by temperature profiles developing from gamma and neutron heating and heat transfer into the coolant. Strain, deformation and gaps between the sheets resulting from this load are determined considering the effect of neutron irradiation induced creep of the core baffle bolts. The finite element code ANSYS C'~is applied for the thermal and mechanical analyses. The FE-model comprises a complete 45 ° sector of the core baffle structure including the core barrel, the formers, the core baffle sheets and about 230 bolt connections with non-linear contact between the single components and the effect of friction. The complete analysis requires three major steps. (1) Evaluation of the three-dimensional distribution of neutron flux an d gamma induced internal heating with the Monte Carlo code MCNP '~. These calculations are based on pin wise power distributions at the core edge for typical loading patterns. (2) Calculation of the temperature distribution in the core baffle for different operational conditions and core loading patterns, considering heat conduction in the components with internal heat sources and convectional boundary conditions (heat transfer coefficients and bulk temperature of the coolant). (3) Calculation of time dependent deformation, stresses and strains taking into account weight, pressure loads, temperature fields for different load situations, pre-stressing, irradiation induced creep of the bolts as correlated to neutron flux. The results show the equalizing effect of redistribution of bolt loads from high flux to lower flux exposure locations in a self controlled process, keeping the mechanical and geometrical stability of the core baffle structure and leaving the gaps between sheet edges unaffected.
05/00091 Application of artificial neural network for safety core parameters prediction in LWRRS Mazrou, H. and Hamadouche, M. Progress in Nuclear Energy, 2004, 44, (3), 263-275. This paper reports on the use of Artificial Neural Networks (ANNs) in predicting core safety parameters, such as multiplication factor Keff and fuel powers peaks P, .... in Light Water Research Reactors (LWRRs) using a personal computer. The purpose of this emulation, when it is associated with optimal incore fuel management, is to allow rapid and extensive exploration of potential good configurations of fuel in L W R R s cores. The key idea of the developed program lies in the use of an adaptive learning rate procedure in a typical back-propagation error model. The performance of the algorithm was improved to the proper setting of the learning rate. Which is adjusted according to the minimal error predicted. Using the 2-Dimensional neutronic diffusion code MUDICO-2D for the referenced target values, the neural network development process is reported for a slight modified version of the IAEA 10Mw benchmark LEU core.
05/00092 Atmospheric dispersion of argon-41 from a nuclear research reactor: measurement and modelling of plume geometry and gamma radiation field, Lauritzen, B. et al. International Journal of' Environment and Pollution, 2003, 20, (1-6), 47 54. An atmospheric dispersion experiment was conducted using a visible tracer along with the routine releases of 41Ar from the BR1 research reactor in Mol, Belgium. Simultaneous measurements of plume geometry and radiation field form 41Ar decay were performed as well as measurements of the 41Ar source term and the meteorological parameters. Good overall agreement is found between measurement data and model results using the mesoscale atmospheric dispersion and dose rate model RIMPUFF.
14
Fuel and Energy Abstracts
January 2005
05/00093 Calorimetric measurements on uraniumplutonium mixed oxides Kandan, R. et al. Journal of Nuclear Materials, 2004, 324, (2-3), 215 219. Enthalpy increments of Ul_yPuyO2 solid solutions with y = 0.21, 0.28 and 0.40 were measured using a high-temperature differential calorimeter by employing the method of inverse drop calorimetry in the temperature range 1000-1780 K. From the fit equations for the enthalpy increments, other thermodynamic functions such as heat capacity, entropy and Gibbs energy function have been computed in the temperature range 298-1800 K. The results indicate that the enthalpies of (U,Pu)O2 solid solutions obey the N e u m a n n - K o p p molar additivity rule.
05/00094 Development of a 30 m long NbaAI superconducting conductor jacketed with a round-in-square type stainless steel conduit for the toroidal field coils in JT-60SC Tsuchiya, K. et al. Fusion Engineering and Design, 2004, 70, (2), 131 140. The modification of JT-60 was programmed to be a superconducting tokamak (JT-60SC) that has superconducting toroidal field (TF) and poloidal field coils. The Nb3AI superconductor was developed in order to apply to the TF coil system in JT-60SC. The TF coil conductor was designed with a cable-in-conduit (CIC) type. This conductor was Nb3AI cable jacketed into stainless steel (SS) conduit. As the first step of SS conduit production, the 10 m long round-in-square type tubes were successfully fabricated within eccentricity tolerance of the tube hole (<10%). A 30 m SS conduit was made of three units of 10 m long tubes, which were butt-welded in series. After the welding process, penetrating beads were observed by CCD camera, and it was visually clear that the penetrating beads were successfully controlled by adjustment of back-shielding gas pressure. Therefore, regarding the production of SS conduit, the procedures of fabrication and inspection were established. Then, the inserting of cable into this conduit and compaction were also carried out. Consequently, the 30 m Nb3A1 CIC conductor was successfully fabricated, so that a feasibility of massproducing CIC conductor applied to the real machine was obtained.
05/00095 Development of environment for remote participation in fusion research on JT-60 Oshima, T. et al. Fusion Engineering and Design, 2004, 71, (1-4), 239 244. In the JT-60 tokamak at JAERI, an environment for remote participation is to be developed by concentrating ideas of experts of nuclear fusion research from research organizations and universities distributed throughout the country. The study involves constructing a hierarchical collaborative remote research system, which consists of remote experiment, remote analysis and remote diagnostics. In remote collaboration, it is important to share information, atmosphere and presence between the participants as well as to maintain the security of the system. To implement a collaborative environment, a video conferencing system, a video streaming system and remote analysis system called VizAnalysis have been developed for the communication. To assist the remote participation for the fusion research, the document management system based on the web browser called VizSquare has been developed. In JT-60, security and authentication methods utilizing a computer network and a new communication tool have been developed, and they are applicable to the remote participation of ITER.
05/00096 Developmental prototype for replacement of JT-60 timing system Akasaka, H. et al. Fusion Engineering and Design, 2004, 71, (1-4), 2934. The present CAMAC-based timing system has been used for synchronizing sequential events of the discharge and the data collection of the interesting JT-60U experiment plasma phenomena. However, a more flexible and sophisticated state-of-the-art timing system now is required to realize advanced plasma control with minimal maintenance costs. In this context, the versa module Europe system with a high-speed data communication network using reflective memory (RM) modules and user-friendly application software based on MATLAB TM tools has been selected to develop the new prototype timing system. In the ZENKEI, the supervisory control system of the JT-60, the supervisory timing system provides the 50-~ts master clock pulses, the various timing signal preparation logic, which is built into the digital signal processing module in conjunction with the discharge sequence event signals, and the 6.2 MB/s high-speed communication data link provided by the RM module. Except the clock pulse generator module, no other special timing module is necessary for this new timing system. The timing signal is prepared by software logic in conjunction with sequential events and the preset timer, is transferred to the subsystems through the RM module, where it is synchronized to the 50~ts clock pulses. The timing system of the subsystems also consists of