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NEA activities relating to innovative nuclear energy systems
05 Nuclear fuels (scientific, technical) 06•00624 New dynamic method to measure rod worths in zero power physics test at PWR startup Lee, E. K. et al. Annals of Nuclear Energy, 2005, 32, (13), 1457 1475. To measure and validate the worth of control (or shutdown) b a n k in zero power physics test at PWRs, a dynamic control rod reactivity m e a s u r e m e n t ( D C R M ) technique has been developed and applied to six start-ups of W e s t i n g h o u s e plants as well as Korea Standard Nuclear power Plants based on the C o m b u s t i o n E n g i n e e r i n g System 80 NSSS. W i t h this technique, just one test b a n k is inserted into the b o t t o m of the core at m a x i m u m stepping rate and withdrawn immediately to the all rod-out position. Specially designed inverse point kinetics equations are used to d e t e r m i n e the test b a n k worth from the m e a s u r e d ex-core detector signals, which are controlled by the neutron-to-response conversion factor and the dynamic-to-static conversion factor. These two p a r a m e t e r s are p r e d e t e r m i n e d by the t h r e e - d i m e n s i o n a l neutron adjoint flux distribution for both the top and b o t t o m ex-core detector and the t h r e e - d i m e n s i o n a l steady and transient core power distribution for test b a n k movement. To eliminate the g a m m a - r a y effect on ex-core detector signals, a simple method, using reactivity curve characteristics, was also developed. To verify the D C R M method, a total of 28 b a n k worths of six different P W R s was m e a s u r e d by the D C R M and c o m p a r e d with those of conventional method. Results show that the D C R M m e t h o d has a similar accuracy as the conventional technique. However, with the D C R M method, it only takes a r o u n d l 5 min per b a n k from the beginning of rod insertion to the d e t e r m i n a t i o n of m e a s u r e d static worth. F r o m its performance, one can conclude that the D C R M m e t h o d is an effective r e p l a c e m e n t for the conventional rod worth m e a s u r e m e n t method.
06•00625 Nuclear power for sustainable development and relevant IAEA activities for the future Omoto, A. Progress in Nuclear Energy, 2005, 47, (1 4), 16 26. The credible longer-term energy d e m a n d and supply analyses foresee a growing role for nuclear power for sustainable development. For instance the Special R e p o r t on Emissions Scenarios of the Interg o v e r n m e n t a l Panel on Climate C h a n g e (IPCC) shows an increase between 2000 and 2050 by a factor of 2.5 in global primary energy and the installed nuclear capacity will increase by about a factor of 4 5 as a m e d i a n value. The technologies for the nuclear energy are continuously improving towards the long-term goals of further i m p r o v e m e n t s in economics, very high levels of safety, increased proliferation resistance, and successful i m p l e m e n t a t i o n of solutions for radioactive waste disposal. By statute, the I A E A is authorized to encourage and assist the M e m b e r States efforts for the practical application of nuclear technology. The Agency's relevant activities are considered to contribute to assist the M e m b e r States to achieve their long-term goals. This paper overviews the current status of nuclear power in the world, discusses its future prospects and describes the I A E A ' s activities to s u p p o r t its M e m b e r States in their efforts for nuclear p r o g r a m m e for sustainable development.
06•00626 OECD/NEA activities relating to innovative nuclear energy systems Marcus, G. H. Progress in Nuclear Energy, 2005, 47, (1 4), 27 31. The mission of the O E C D Nuclear E n e r g y A g e n c y ( N E A ) is to assist its m e m b e r countries in maintaining and further developing, through i n t e r n a t i o n a l co-operation, the scientific, technological and legal bases required for the safe, environmentally friendly and economical use of nuclear energy for peaceful purposes. In fulfilling that mission, the N E A conducts technical, economic and policy studies in response to the needs and interests of its 28 m e m b e r countries. In recent years, a n u m b e r of these studies have a d d r e s s e d various aspects of the next g e n e r a t i o n of nuclear power plants. This paper will describe some of the major activities recently completed and currently underway that may be of particular interest to the C O E - I N E S program.
06•00627 On the possibility of the space-dependence of the stability indicator (decay ratio) of a BWR Demazi~re, C. et al. Pitzsit, I. Annals of Nuclear Energy, 2005, 32, (12), 1305 1322. A m o d e l is proposed for the explanation of the s p a c e - d e p e n d e n c e of the so-called decay ratio (DR) which is used to quantify the stability properties of boiling water reactors (BWRs). The study was p r o m p t e d by the observation of a strongly s p a c e - d e p e n d e n t decay ratio in an instability event at the Swedish Forsmark-1 BWR. Prior to that event, the s p a c e - d e p e n d e n c e of the D R was neither observed, nor assumed possible in the theoretical models of instability. The m o d e l proposed here is based on a previous suggestion by one of the authors on how to m o d e l the estimation of the D R in case of two different types of oscillations (instabilities) being present in the core simultaneously. The m o d e l was earlier only used in a s p a c e - i n d e p e n d e n t form, but here its applicability is extended such that s p a c e - d e p e n d e n c e of the oscillations is also accounted for, by using a noise simulator. The investigations show that the DR, as d e t e r m i n e d by the individual L P R M s (neutron detectors) at different positions, can be strongly s p a c e - d e p e n d e n t if at
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Fuel and Energy Abstracts
March 2006
least two different oscillations with differing D R and space-dependence exist in the core simultaneously. The observed s p a c e - d e p e n d e n c e of the D R in the F o r s m a r k case can be reconstructed by the model.
06•00628 Parametric studies on different gas turbine cycles for a high temperature gas-cooled reactor Wang, J. and Gu, Y. Nuclear Engineering ancl Design, 2005, 235, (16), 1761 1772. The high t e m p e r a t u r e gas-cooled reactor ( H T G R ) coupled with turbine cycle is considered as one of the leading candidates for future nuclear power plants. In this paper, the various types of H T G R gas turbine cycles are concluded as three typical cycles of direct cycle, closed indirect cycle and open indirect cycle. F u r t h e r m o r e they are theoretically converted to three Brayton cycles of helium, nitrogen and air. Those three types of Brayton cycles are thermodynamically analysed and optimized. The results show that the variety of gas affects the cycle pressure ratio more significantly than other cycle parameters, however, the optimized cycle efficiencies of the three Brayton cycles are almost the same. In addition, the turbo machines that are required for the three optimized Brayton cycles are aerodynamically analysed and c o m p a r e d and their f u n d a m e n t a l characteristics are obtained. H e l i u m turbo-compressor has lower stage pressure ratio and more stage n u m b e r than those for nitrogen and air machines, while h e l i u m and nitrogen turbo-compressors have shorter blade length than that for air machine.
06•00629 Prediction of fission mass-yield distributions based on cross section calculations Hambsch, F.-J. et al. Annals of Nuclear Energy, 2005, 32, (12), 1297 1304. For the first time, fission mass-yield distributions have been predicted based on an extended statistical m o d e l for fission cross-section calculations. In this model, the concept of the multi-modality of the fission process has b e e n incorporated. The three most d o m i n a n t fission modes, the two asymmetric standard I ($1) and standard II ($2) m o d e s and the symmetric s u p e r l o n g (SL) m o d e are t a k e n into account. Deconvoluted fission cross sections for $1, $2 and SL m o d e s for 23s'238U(n, f) and 237Np(n, f), based on e x p e r i m e n t a l branching ratios, were calculated for the first time in the incident n e u t r o n energy range from 0.01 to 5.5 M e V providing good a g r e e m e n t with the e x p e r i m e n t a l fission cross section data. The b r a n c h i n g ratios obtained from the modal fission cross section calculations have been used to deduce the corresponding fission yield distributions, including m e a n values also for incident n e u t r o n energies hitherto not accessible to experiment.
06•00630 Radiation damage studies on the first wall of a HYLIFE-II type fusion breeder Sahin, S. and Ubeyli, M. Energy Conversion and Management, 2005, 46, (20), 3185 3201. The radiation d a m a g e on the first wall [made of (1) a ferritic steel (9Cr 2WVTa), (2) a v a n a d i u m alloy ( V 4 C r 4Ti) and (3) S i C j S i C composite] of an inertial fusion energy (IFE) reactor of H Y L I F E - I I type is investigated. A protective liquid wall with variable thickness, containing Flibe + heavy metal salt (UF4 or ThF4) is used for first wall protection. The content of heavy metal salt is chosen as 4 and 12 mol%. Neutron transport calculations are performed with the aid of the SCALE4.3 System by solving the Boltzmann t r a n s p o r t equation with the X S D R N P M code in 238 energy groups and S~ P3 approximation. A flowing wall with a thickness of ~60 cm can extend the lifetime of the solid first wall structure to a plant lifetime of 30 years for 9Cr 2 W V T a and V 4 C r 4Ti, whereas the S i C j S i C composite as first wall needs a flowing wall with a thickness of ~85 cm to maintain the radiation d a m a g e limit. Substantial extra revenue can be gained through the insertion of a heavy m e t a l salt constituent into Flibe, which allows b r e e d i n g fissile fuel for external reactors and increasing energy multiplication (233U with a value of up to $1,000,O00,O00/year or 239Pu for few $100,O00,O00/year ).
06•00631 RELAP5/MOD3.2 investigation of reactor vessel YR line capabilities for primary side depressurization during the TLFW in VVER1000/V320 Gencheva, R. V. et al. Annals of Nuclear Energy, 2005, 32, (12), 1407 1434. D u r i n g the d e v e l o p m e n t of Symptom Based E m e r g e n c y O p e r a t i n g Procedures (SB-EOPs) for VVER-1000/V320 units at Kozloduy Nuclear Power Plant (NPP), a n u m b e r of analyses have been performed using the R E L A P 5 / M O D 3 . 2 computer code. One of them is 'Investigation of reactor vessel Y R line capabilities for primary side depressurization during the total loss of feed water (TLFW)'. The main purpose of these calculations is to evaluate the capabilities of Y R line located at the top of the reactor vessel for primary side depressurization to the set point of high pressure injection system (HPIS) actuation and the abilities for successful core cooling after Feed & Bleed procedure initiation. For the purpose of this, operator action with ' R e a c t o r vessel off-gas valve 0.032 m' opening has b e e n investigated.
06•00625 Nuclear power for sustainable development and relevant IAEA activities for the future Omoto, A. Progress in Nuclear Energy, 2005, 47, (1 4), 16 26. The credible longer-term energy d e m a n d and supply analyses foresee a growing role for nuclear power for sustainable development. For instance the Special R e p o r t on Emissions Scenarios of the Interg o v e r n m e n t a l Panel on Climate C h a n g e (IPCC) shows an increase between 2000 and 2050 by a factor of 2.5 in global primary energy and the installed nuclear capacity will increase by about a factor of 4 5 as a m e d i a n value. The technologies for the nuclear energy are continuously improving towards the long-term goals of further i m p r o v e m e n t s in economics, very high levels of safety, increased proliferation resistance, and successful i m p l e m e n t a t i o n of solutions for radioactive waste disposal. By statute, the I A E A is authorized to encourage and assist the M e m b e r States efforts for the practical application of nuclear technology. The Agency's relevant activities are considered to contribute to assist the M e m b e r States to achieve their long-term goals. This paper overviews the current status of nuclear power in the world, discusses its future prospects and describes the I A E A ' s activities to s u p p o r t its M e m b e r States in their efforts for nuclear p r o g r a m m e for sustainable development.
06•00626 OECD/NEA activities relating to innovative nuclear energy systems Marcus, G. H. Progress in Nuclear Energy, 2005, 47, (1 4), 27 31. The mission of the O E C D Nuclear E n e r g y A g e n c y ( N E A ) is to assist its m e m b e r countries in maintaining and further developing, through i n t e r n a t i o n a l co-operation, the scientific, technological and legal bases required for the safe, environmentally friendly and economical use of nuclear energy for peaceful purposes. In fulfilling that mission, the N E A conducts technical, economic and policy studies in response to the needs and interests of its 28 m e m b e r countries. In recent years, a n u m b e r of these studies have a d d r e s s e d various aspects of the next g e n e r a t i o n of nuclear power plants. This paper will describe some of the major activities recently completed and currently underway that may be of particular interest to the C O E - I N E S program.
06•00627 On the possibility of the space-dependence of the stability indicator (decay ratio) of a BWR Demazi~re, C. et al. Pitzsit, I. Annals of Nuclear Energy, 2005, 32, (12), 1305 1322. A m o d e l is proposed for the explanation of the s p a c e - d e p e n d e n c e of the so-called decay ratio (DR) which is used to quantify the stability properties of boiling water reactors (BWRs). The study was p r o m p t e d by the observation of a strongly s p a c e - d e p e n d e n t decay ratio in an instability event at the Swedish Forsmark-1 BWR. Prior to that event, the s p a c e - d e p e n d e n c e of the D R was neither observed, nor assumed possible in the theoretical models of instability. The m o d e l proposed here is based on a previous suggestion by one of the authors on how to m o d e l the estimation of the D R in case of two different types of oscillations (instabilities) being present in the core simultaneously. The m o d e l was earlier only used in a s p a c e - i n d e p e n d e n t form, but here its applicability is extended such that s p a c e - d e p e n d e n c e of the oscillations is also accounted for, by using a noise simulator. The investigations show that the DR, as d e t e r m i n e d by the individual L P R M s (neutron detectors) at different positions, can be strongly s p a c e - d e p e n d e n t if at
96
Fuel and Energy Abstracts
March 2006
least two different oscillations with differing D R and space-dependence exist in the core simultaneously. The observed s p a c e - d e p e n d e n c e of the D R in the F o r s m a r k case can be reconstructed by the model.
06•00628 Parametric studies on different gas turbine cycles for a high temperature gas-cooled reactor Wang, J. and Gu, Y. Nuclear Engineering ancl Design, 2005, 235, (16), 1761 1772. The high t e m p e r a t u r e gas-cooled reactor ( H T G R ) coupled with turbine cycle is considered as one of the leading candidates for future nuclear power plants. In this paper, the various types of H T G R gas turbine cycles are concluded as three typical cycles of direct cycle, closed indirect cycle and open indirect cycle. F u r t h e r m o r e they are theoretically converted to three Brayton cycles of helium, nitrogen and air. Those three types of Brayton cycles are thermodynamically analysed and optimized. The results show that the variety of gas affects the cycle pressure ratio more significantly than other cycle parameters, however, the optimized cycle efficiencies of the three Brayton cycles are almost the same. In addition, the turbo machines that are required for the three optimized Brayton cycles are aerodynamically analysed and c o m p a r e d and their f u n d a m e n t a l characteristics are obtained. H e l i u m turbo-compressor has lower stage pressure ratio and more stage n u m b e r than those for nitrogen and air machines, while h e l i u m and nitrogen turbo-compressors have shorter blade length than that for air machine.
06•00629 Prediction of fission mass-yield distributions based on cross section calculations Hambsch, F.-J. et al. Annals of Nuclear Energy, 2005, 32, (12), 1297 1304. For the first time, fission mass-yield distributions have been predicted based on an extended statistical m o d e l for fission cross-section calculations. In this model, the concept of the multi-modality of the fission process has b e e n incorporated. The three most d o m i n a n t fission modes, the two asymmetric standard I ($1) and standard II ($2) m o d e s and the symmetric s u p e r l o n g (SL) m o d e are t a k e n into account. Deconvoluted fission cross sections for $1, $2 and SL m o d e s for 23s'238U(n, f) and 237Np(n, f), based on e x p e r i m e n t a l branching ratios, were calculated for the first time in the incident n e u t r o n energy range from 0.01 to 5.5 M e V providing good a g r e e m e n t with the e x p e r i m e n t a l fission cross section data. The b r a n c h i n g ratios obtained from the modal fission cross section calculations have been used to deduce the corresponding fission yield distributions, including m e a n values also for incident n e u t r o n energies hitherto not accessible to experiment.
06•00630 Radiation damage studies on the first wall of a HYLIFE-II type fusion breeder Sahin, S. and Ubeyli, M. Energy Conversion and Management, 2005, 46, (20), 3185 3201. The radiation d a m a g e on the first wall [made of (1) a ferritic steel (9Cr 2WVTa), (2) a v a n a d i u m alloy ( V 4 C r 4Ti) and (3) S i C j S i C composite] of an inertial fusion energy (IFE) reactor of H Y L I F E - I I type is investigated. A protective liquid wall with variable thickness, containing Flibe + heavy metal salt (UF4 or ThF4) is used for first wall protection. The content of heavy metal salt is chosen as 4 and 12 mol%. Neutron transport calculations are performed with the aid of the SCALE4.3 System by solving the Boltzmann t r a n s p o r t equation with the X S D R N P M code in 238 energy groups and S~ P3 approximation. A flowing wall with a thickness of ~60 cm can extend the lifetime of the solid first wall structure to a plant lifetime of 30 years for 9Cr 2 W V T a and V 4 C r 4Ti, whereas the S i C j S i C composite as first wall needs a flowing wall with a thickness of ~85 cm to maintain the radiation d a m a g e limit. Substantial extra revenue can be gained through the insertion of a heavy m e t a l salt constituent into Flibe, which allows b r e e d i n g fissile fuel for external reactors and increasing energy multiplication (233U with a value of up to $1,000,O00,O00/year or 239Pu for few $100,O00,O00/year ).
06•00631 RELAP5/MOD3.2 investigation of reactor vessel YR line capabilities for primary side depressurization during the TLFW in VVER1000/V320 Gencheva, R. V. et al. Annals of Nuclear Energy, 2005, 32, (12), 1407 1434. D u r i n g the d e v e l o p m e n t of Symptom Based E m e r g e n c y O p e r a t i n g Procedures (SB-EOPs) for VVER-1000/V320 units at Kozloduy Nuclear Power Plant (NPP), a n u m b e r of analyses have been performed using the R E L A P 5 / M O D 3 . 2 computer code. One of them is 'Investigation of reactor vessel Y R line capabilities for primary side depressurization during the total loss of feed water (TLFW)'. The main purpose of these calculations is to evaluate the capabilities of Y R line located at the top of the reactor vessel for primary side depressurization to the set point of high pressure injection system (HPIS) actuation and the abilities for successful core cooling after Feed & Bleed procedure initiation. For the purpose of this, operator action with ' R e a c t o r vessel off-gas valve 0.032 m' opening has b e e n investigated.