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03695 Core design of a high-temperature fast reactor cooled by supercritical light water
05
Nuclear fuels (scientific, technical)
The use of polymers and coal combustion byproducts for amelioration of crusting in disturbed soils
99103694
Stoddard, A. A., III 1998, 147 pp. Avail. UMI, Order From Diss. Abstr. Inl., B, 1998, 59, (6). 2514.
05
NUCLEAR Scientific,
No. DA9836987.
FUELS
Technical
Dynamic response analyses of the upper guide 99103699 structure assembly for Korean next generation reactor
Ji, Y. K. and Lee, Y. S. Annals Nucl. Energy, 1999, 26, (16). 1457-1464. Developed in this paper is a finite element model of the upper guide structure assembly for the Korean next generation reactor. The model considers the annulus effects due to water gaps existing between the upper guide structure and the inner barrel assembly/the core support barrel. The responses of the upper guide structure assembly to dynamic flow loads, which are random pressure fluctuations due to turbulent flow and harmonic pump pulsation pressures, are obtained. An efficient procedure of random response analysis is developed to facilitate the combination of the significant mode responses under correlated and uncorrelated cases. The transient response evaluation is proposed to compensate uncertainty for random and pump pulsation loads.
H, filtering for dynamic corn ensation of selfpoey;zd neutron detectors-a linear matrPx inequality based
99/03700
Core desi n of a high-temperature fast reactor cooled by supercrltica Blight water
Mukohara, T. er al. Annals Nucl. Energy, 1999, 26, (16), 1423-1436. A high-temperature large fast reactor cooled by supercritical water (SCFRH) is designed for assessing its technical feasibility and potential economical improvement. The coolant system is a once-through, direct cycle where whole core coolant flows to the turbine. The goal is to achieve the high coolant outlet temperature over 500°C. The reactors with blankets cooled by ascending and descending flow are studied. SCFR-H adopts a radial heterogeneous core with zirconium-hydride layers between the driver core and the blankets for making coolant void reactivity negative. The coolant outlet temperature of the core with blankets cooled by ascending flow is low. 467°C.
Park, M.-G. et al. Annals Nucl. Energy, 1999, 26, (18), 1669-1682. A method is described to develop an H, filtering method for the dynamic compensation of self-powered neutron detectors normally used for fixed incore instruments, An H, norm of the filter transfer matrix is used as the optimization criteria in the worst-case estimation error sense. Filter modelling is performed for both continuousand discrete-time models. The filter gains are optimized in the sense of noise attenuation level of H, setting. By introducing Bounded Real Lemma, the conventional algebraic Riccati inequalities are converted into Linear Matrix Inequalities (LMIs). Finally, the filter design problem is solved via the convex optimization framework using LMIs. The simulation results show that remarkable improvements are achieved in view of the filter response time and the filter design efficiency.
99103696
99lQ3701
Ricciardi, M. V. el al. Annals Nucl. Energy, 1999, 26, (18), 1643-1656. With regard to the ISOL technique, the dependence of intermediate mass (80-160 A) production on target and proton beam features has been investigated by means of hadronic transport Monte Carlo codes. The analysis of the target material (mechanism of production), dimensions (length and width) and composition as well as the proton beam energy has shown that in principle, regardless of technological difficulties, the production of neutron-rich isotopes of intermediate mass (SO-160 A) can be successfully achieved with proton beams of low-medium energy into uranium thick targets. Additionally, the power density in uranium targets bombarded by low-medium energy protons has been analysed.
Van Dam, H. Annals Nucl. Energy, 1999, 26, (16), 1489-1495. The consequences of loss-of-cooling-without-scram events are analysed for critical and subcritical (accelerator driven) reactors. Use is made of point dynamic models with different levels of approximation: the most realistic model is based on six groups of delayed neutrons and 33 groups of fission products. It is shown that the behaviour of subcritical systems is unfavourable compared to critical systems. Treatment of neutron kinetics of moderately and strongly subcritical systems can be based on the very simple model of prompt multiplication, but to obtain realistic results, decay heat of fission products should be taken into account.
99103695
Dependence of exotic nuciides production on target and beam features
Loss-of-cooling consequences in critical and in accelerator driven reactor systems
Parametric stochastic stabilltv and decav ratio for a stochastic nonlinear BWR model below ttie Hopf bif&rcatlon 99lO3702
Development of a multi-dimensional thermalhydraulic system code, MARS 1.3.1
99103697
Jeong, J.-J. et al. Annals Nucl. Energy, 1999, 26, (18), 1611-1642. A multi-dimensional thermal-hydraulic system code, MARS, has been developed by consolidating and restructuring the RELAPS/MOD3.2.1.2 and COBRA-TF codes. The two codes were adopted to take advantage of the very general, versatile features of RELAPS and the realistic threedimensional hydrodynamic module of COBRA-TF. In the course of code development, major features of each code were consolidated into a single code first. The resulting source programs were rewritten in standard FORTRAN 90, and then were restructured using modular data structures based on ‘derived type variables’ and a new ‘dynamic memory allocation’ scheme. In additiont’the Windows graphics features were implkmented for user friendliness. This paper presents the developmental activities up to MARS version 1.3.1 including the code consolidation, the code restructuring and modernization and the results of the developmental assessment.
Development of conceptual design tool for liquid metal cooled-reactors
99103696
Lee, K. G. and Chang, S. H. Annals Nucl. Energy, 1999, 26, (18). 16571668. A conceotual design tool for three tvnes of liauid metal reactors (LMRs), ultra-loig life, accklerator-driven sudiritical transmutation and large-scaleh LMRs has been proposed. This tool is developed for system design with artificial intelligence, scaling of design parameters and validation analysis. The system design consists of design synthesis and design analysis. System design decides the optimal structure and layout of a LMR using design synthesis with rule bases and databases and design analysis with design constraints. The designed system is scaled by scale laws to be optimal with desired power level and then a specific design basis accident is analysed in validation part. Design synthesis contains the data about each component and general LMRs and the rules about selection of each component and general LMRs. Design analysis contains several design constraints for formulation and solution of a design problem. In these two parts, a designer’s intention can be externalized through emphasis on design requirements. For the purpose of demonstration, the conceptual design tool was applied to an ultra long life LMR with 35 MWe power level. This ultra-long life LMR was designed from knowledge structure based on all the characteristics of ultra long life, accelerator-driven subcritical transmutation and large-scaled LMRs and then optimally scaled to the 35 MWe power level.
390
Fuel and Energy Abstracts
November 1999
Konno, H. et al. Annals Nucl. Energy, 1999, 26, (16), 1465-1487. The effects of the reactivity noise p,(t) ( a coloured parametric noise) upon the neutron fluctuation has been studied with the use of a non-linear BWR It is clarified that there appears model below the Hopf bifurcation. temporal variation of a local stability in .the neutron field under the influence of p,(t). It is also clarified that p,(t) works to change a global stability of the neutron field. The feature of global variations in the decay ratio (DR) for an exponentially correlated noise cp,(l)p,(t’)> = exp(-ylt-t’l) in the 2D (7, ) parameter space is shown in conjunction with physical interpretation of the correlation function of the neutron fluctuation.
99103703 Thermal behaviour of CANDU t pe fuel rods during steady state and transient operating condlt ron8 Horhoianu, G. et al. Annals Nucl. Energy, 1999, 26, (16), 1427-1445. The paper oresents in-core measurements on fuel rods’ thermal behaviour duririg steaby-state, as well as transient, power conditions. Based on these data, the influence of filling gas, linear power and burn-up on CANDU-type fuel centre temperature evolution are discussed. Thermal behaviour characteristics up to -69 MWh/kgU of helium and xenon prefilled fuel rods instrumented with centreline thermocouples are analysed
TRU incineration characteristics of thermal and fast subcritical reactors
99lQ3704
Park, W. S. et al. Annals Nucl. Energy, 1999, 26, (16), 1497-1508. In terms of the incineration capability for each TRU nuclide, TRU inventory, beam current fluctuation (net multiplication swing) and power peaking, etc., accelerator driven thermal and fast neutron systems were investigated. The thermal system was modelled on a CANDU reactor and the fast system was constructed from a typical LMR with lead-bismuth coolant. TRU mixed with thorium were employed as a fuel and the TRU inventory in each system was adjusted to make the system eigenvalue 0.97. The beam energy was assumed to be 1 GeV and the beam current was controlled to maintain the system power 1000 MWth. The thermal system was found to need 754.3 kg of TRU loading while the fast system needed 2331 kg at BOL condition. The required beam currents were 28.346 mA and 10.19 mA for thermal and fast systems, respectively, at BOL. The fast system showed the superiority in terms of uniform homing capability, beam current fluctuation, power peaking while the thermal system did in terms of TRU inventory. Especially related to power peaking problems, a liquid-type
Nuclear fuels (scientific, technical)
The use of polymers and coal combustion byproducts for amelioration of crusting in disturbed soils
99103694
Stoddard, A. A., III 1998, 147 pp. Avail. UMI, Order From Diss. Abstr. Inl., B, 1998, 59, (6). 2514.
05
NUCLEAR Scientific,
No. DA9836987.
FUELS
Technical
Dynamic response analyses of the upper guide 99103699 structure assembly for Korean next generation reactor
Ji, Y. K. and Lee, Y. S. Annals Nucl. Energy, 1999, 26, (16). 1457-1464. Developed in this paper is a finite element model of the upper guide structure assembly for the Korean next generation reactor. The model considers the annulus effects due to water gaps existing between the upper guide structure and the inner barrel assembly/the core support barrel. The responses of the upper guide structure assembly to dynamic flow loads, which are random pressure fluctuations due to turbulent flow and harmonic pump pulsation pressures, are obtained. An efficient procedure of random response analysis is developed to facilitate the combination of the significant mode responses under correlated and uncorrelated cases. The transient response evaluation is proposed to compensate uncertainty for random and pump pulsation loads.
H, filtering for dynamic corn ensation of selfpoey;zd neutron detectors-a linear matrPx inequality based
99/03700
Core desi n of a high-temperature fast reactor cooled by supercrltica Blight water
Mukohara, T. er al. Annals Nucl. Energy, 1999, 26, (16), 1423-1436. A high-temperature large fast reactor cooled by supercritical water (SCFRH) is designed for assessing its technical feasibility and potential economical improvement. The coolant system is a once-through, direct cycle where whole core coolant flows to the turbine. The goal is to achieve the high coolant outlet temperature over 500°C. The reactors with blankets cooled by ascending and descending flow are studied. SCFR-H adopts a radial heterogeneous core with zirconium-hydride layers between the driver core and the blankets for making coolant void reactivity negative. The coolant outlet temperature of the core with blankets cooled by ascending flow is low. 467°C.
Park, M.-G. et al. Annals Nucl. Energy, 1999, 26, (18), 1669-1682. A method is described to develop an H, filtering method for the dynamic compensation of self-powered neutron detectors normally used for fixed incore instruments, An H, norm of the filter transfer matrix is used as the optimization criteria in the worst-case estimation error sense. Filter modelling is performed for both continuousand discrete-time models. The filter gains are optimized in the sense of noise attenuation level of H, setting. By introducing Bounded Real Lemma, the conventional algebraic Riccati inequalities are converted into Linear Matrix Inequalities (LMIs). Finally, the filter design problem is solved via the convex optimization framework using LMIs. The simulation results show that remarkable improvements are achieved in view of the filter response time and the filter design efficiency.
99103696
99lQ3701
Ricciardi, M. V. el al. Annals Nucl. Energy, 1999, 26, (18), 1643-1656. With regard to the ISOL technique, the dependence of intermediate mass (80-160 A) production on target and proton beam features has been investigated by means of hadronic transport Monte Carlo codes. The analysis of the target material (mechanism of production), dimensions (length and width) and composition as well as the proton beam energy has shown that in principle, regardless of technological difficulties, the production of neutron-rich isotopes of intermediate mass (SO-160 A) can be successfully achieved with proton beams of low-medium energy into uranium thick targets. Additionally, the power density in uranium targets bombarded by low-medium energy protons has been analysed.
Van Dam, H. Annals Nucl. Energy, 1999, 26, (16), 1489-1495. The consequences of loss-of-cooling-without-scram events are analysed for critical and subcritical (accelerator driven) reactors. Use is made of point dynamic models with different levels of approximation: the most realistic model is based on six groups of delayed neutrons and 33 groups of fission products. It is shown that the behaviour of subcritical systems is unfavourable compared to critical systems. Treatment of neutron kinetics of moderately and strongly subcritical systems can be based on the very simple model of prompt multiplication, but to obtain realistic results, decay heat of fission products should be taken into account.
99103695
Dependence of exotic nuciides production on target and beam features
Loss-of-cooling consequences in critical and in accelerator driven reactor systems
Parametric stochastic stabilltv and decav ratio for a stochastic nonlinear BWR model below ttie Hopf bif&rcatlon 99lO3702
Development of a multi-dimensional thermalhydraulic system code, MARS 1.3.1
99103697
Jeong, J.-J. et al. Annals Nucl. Energy, 1999, 26, (18), 1611-1642. A multi-dimensional thermal-hydraulic system code, MARS, has been developed by consolidating and restructuring the RELAPS/MOD3.2.1.2 and COBRA-TF codes. The two codes were adopted to take advantage of the very general, versatile features of RELAPS and the realistic threedimensional hydrodynamic module of COBRA-TF. In the course of code development, major features of each code were consolidated into a single code first. The resulting source programs were rewritten in standard FORTRAN 90, and then were restructured using modular data structures based on ‘derived type variables’ and a new ‘dynamic memory allocation’ scheme. In additiont’the Windows graphics features were implkmented for user friendliness. This paper presents the developmental activities up to MARS version 1.3.1 including the code consolidation, the code restructuring and modernization and the results of the developmental assessment.
Development of conceptual design tool for liquid metal cooled-reactors
99103696
Lee, K. G. and Chang, S. H. Annals Nucl. Energy, 1999, 26, (18). 16571668. A conceotual design tool for three tvnes of liauid metal reactors (LMRs), ultra-loig life, accklerator-driven sudiritical transmutation and large-scaleh LMRs has been proposed. This tool is developed for system design with artificial intelligence, scaling of design parameters and validation analysis. The system design consists of design synthesis and design analysis. System design decides the optimal structure and layout of a LMR using design synthesis with rule bases and databases and design analysis with design constraints. The designed system is scaled by scale laws to be optimal with desired power level and then a specific design basis accident is analysed in validation part. Design synthesis contains the data about each component and general LMRs and the rules about selection of each component and general LMRs. Design analysis contains several design constraints for formulation and solution of a design problem. In these two parts, a designer’s intention can be externalized through emphasis on design requirements. For the purpose of demonstration, the conceptual design tool was applied to an ultra long life LMR with 35 MWe power level. This ultra-long life LMR was designed from knowledge structure based on all the characteristics of ultra long life, accelerator-driven subcritical transmutation and large-scaled LMRs and then optimally scaled to the 35 MWe power level.
390
Fuel and Energy Abstracts
November 1999
Konno, H. et al. Annals Nucl. Energy, 1999, 26, (16), 1465-1487. The effects of the reactivity noise p,(t) ( a coloured parametric noise) upon the neutron fluctuation has been studied with the use of a non-linear BWR It is clarified that there appears model below the Hopf bifurcation. temporal variation of a local stability in .the neutron field under the influence of p,(t). It is also clarified that p,(t) works to change a global stability of the neutron field. The feature of global variations in the decay ratio (DR) for an exponentially correlated noise cp,(l)p,(t’)> =
99103703 Thermal behaviour of CANDU t pe fuel rods during steady state and transient operating condlt ron8 Horhoianu, G. et al. Annals Nucl. Energy, 1999, 26, (16), 1427-1445. The paper oresents in-core measurements on fuel rods’ thermal behaviour duririg steaby-state, as well as transient, power conditions. Based on these data, the influence of filling gas, linear power and burn-up on CANDU-type fuel centre temperature evolution are discussed. Thermal behaviour characteristics up to -69 MWh/kgU of helium and xenon prefilled fuel rods instrumented with centreline thermocouples are analysed
TRU incineration characteristics of thermal and fast subcritical reactors
99lQ3704
Park, W. S. et al. Annals Nucl. Energy, 1999, 26, (16), 1497-1508. In terms of the incineration capability for each TRU nuclide, TRU inventory, beam current fluctuation (net multiplication swing) and power peaking, etc., accelerator driven thermal and fast neutron systems were investigated. The thermal system was modelled on a CANDU reactor and the fast system was constructed from a typical LMR with lead-bismuth coolant. TRU mixed with thorium were employed as a fuel and the TRU inventory in each system was adjusted to make the system eigenvalue 0.97. The beam energy was assumed to be 1 GeV and the beam current was controlled to maintain the system power 1000 MWth. The thermal system was found to need 754.3 kg of TRU loading while the fast system needed 2331 kg at BOL condition. The required beam currents were 28.346 mA and 10.19 mA for thermal and fast systems, respectively, at BOL. The fast system showed the superiority in terms of uniform homing capability, beam current fluctuation, power peaking while the thermal system did in terms of TRU inventory. Especially related to power peaking problems, a liquid-type