- Email: [email protected]
02036 Inelastic analysis of cylindrical steel containment vessels under internal accident conditions
05 Nuclear fuels (scientific, technical) found that in the first case, core volume reduces with increasing fuel loadings per plate but r e q u i r e m e n t of fuel also increases. In the second and third case, core v o l u m e as well as fuel r e q u i r e m e n t decreases with increasing fuel loadings per plate. However in the second case, core v o l u m e reduces more rapidly than in case 3 with increasing fuel loadings per plate. E m p l o y i n g standard computer code P A R E T , steady state t h e r m a l hydraulic analysis of all these cores was performed. The t h e r m a l hydraulic analysis reveals that cores with higher densities and fixed w a t e r channel width are better from t h e r m a l hydraulic point of view and have fuel and clad t e m p e r a t u r e s within the acceptable limits. But the core with higher densities and o p t i m u m w a t e r channel width is a b e t t e r choice in terms of core compaction, less e3SU loading and higher neutron fluxes. Finally, the core was compacted in three steps to exploit the benefits of both types of cores. The strategy resulted in 36% reduction in the core volume, 50% increase in t h e r m a l n e u t r o n flux for irradiation and isotope production and a slight reduction in e3SU loading. All this was achieved with acceptable p e a k clad and p e a k fuel centreline temperatures.
fission chambers and m o d e r a t o r are given. Thus this b e n c h m a r k is just a m a t h e m a t i c a l test that allows testing the accuracy of the n e u t r o n t r a n s p o r t equation solution with different methods and codes. In this paper the G e n e r a l First Collision Probabilities M e t h o d ( G F C P M ) is used to analyse the two-dimensional configuration of this b e n c h m a r k . A linear flux approximation is used in the reflector. Different calculation schemes in the reflector region have been used. The o u t p u t results, I~ff and the pin powers have been analysed. The convergence of the results has been analysed both as a function of the subdivision scheme of the reflector region and of the n u m b e r of points in the calculation scheme for general first collision probabilities. C o m p a r i s o n has been carried out for Keff and pin powers both with the reference results (external convergence) and with the results of different approximations of G F C P M (internal convergence).
05•02034 Hydro-thermal-mechanical analysis of thermal fatigue in a mixing tee
A l b a r h o u m , M. Annals of Nuclear Energy, 2004, 31, (18), 2203 2209. The possibility of extending the operable time of the Syrian M N S R is investigated through a t h r e e - d i m e n s i o n a l detailed model of the reactor constructed for this purpose. G o o d a g r e e m e n t between calculated and m e a s u r e d reactor p a r a m e t e r values were obtained for the reactor before modification. The operable time is increased by increasing the initial available excess reactivity. The latter is increased by adding Top Beryllium Shims in the Shim Tray. The increased initial excess reactivity is c o m p e n s a t e d for by increasing control rod worth by substituting the actual c a d m i u m absorber by 5B10 a b s o r b e r . The shut down margin is also e n h a n c e d and safer reactor is obtained.
Chapuliot, S. et al. Nuclear Engineering andDesign, 2005, 235, (5), 575 596. This paper covers work carried out by the C E A to study the m e c h a n i s m s leading to cracking of piping as a result of t h e r m a l loading in flow mixing zones. The main goal of the work is to analyse, by calculation, the t h e r m a l loading caused by t u r b u l e n t mixing in tees and to u n d e r s t a n d the m e c h a n i s m of initiation and propagation of cracks in such components. This work is s u p p o r t e d by IRSN. This t h e r m a l fatigue p h e n o m e n o n is still not fully understood. One of the main obstacles to its u n d e r s t a n d i n g resides in the m u l t i - d o m a i n nature of the loading and associated damage, involving three c o m p l e m e n t a r y scientific disciplines: thermal-hydraulic field, t h e r m o - m e c h a n i c a l field and materials science. This paper describes the approach adopted by the C E A to establish natural m e c h a n i s m s (turbulence, pulsing and instability) which might be the cause of any substantial thermomechanical loading in the piping. A l t h o u g h turbulence may be the cause of the t h e r m a l stripping (presence of high-frequency t h e r m a l fluctuations on the inner surface of the component), it c a n n o t alone explain the p r o p a g a t i o n of deep cracks. The main reason is the 'highpass filter' effect of convection. The wall c a n n o t be subjected to convection-related t h e r m a l fluctuations and frequencies less than the inverse of the turbulence transit time. A straightforward frequencybased analysis of the loading, carried out as a first stage, made it possible to establish the limits of the loading created by these highfrequency events. However, turbulence can give rise to flow instability (such as pulsing) of lower frequency. But this cannot explain everything. The geometry u p s t r e a m of the tee, particularly the sequence of straight sections and bends can, in certain cases, d a m p the pulses or greatly amplify them. The use of suitable thermalhydraulic m o d e l l i n g is discussed in the second part of this article. The final result of the t h e r m o - h y d r o - m e c h a n i c a l link-up on application to the complex 3D geometry of the Civaux unit 1 case (which includes a mixing tee, bends and straight sections) enabled the observations m a d e in this plant case to be highlighted and correlated. One of the originalities of this study is to carry out the overall analysis (thermalhydraulic and thermo-mechanical) with a single computer code, the C A S T 3 M code developed by the CEA.
05•02032 Helios, current coupling collision probability method, applied for solving the NEA C5G7 MOX benchmark
05•02035 Improved approach for obtaining rotational components of seismic motion
05•02030 Evaluation of the burst characteristics for axial notches on SG tubings Hwang, S. S. et al. Nuclear Engineering andDesign, 2004, 232, (2), 139 143. Some events of steam generator tubes have been reported in some nuclear power plants a r o u n d the world. Main causes of the leakage are from various types of corrosion in the s t e a m g e n e r a t o r (SG) tubing. Primary water stress corrosion cracking (PWSCC) of steam generator tubing have occurred in many tubes in Korean plants, and they were repaired using sleeves or plugs. In order to develop p r o p e r repair criteria, it is necessary to ascertain the leak behaviour of the tubings. A high-pressure leak and burst testing system was manufactured. Various types of electro-discharged-machined ( E D M ) notches having different lengths were m a c h i n e d on the outside d i a m e t e r of test tubes to study SG tube behaviour. L e a k rate and ligament rupture pressure as well as the burst pressure were m e a s u r e d for the tubes at r o o m temperature. R u p t u r e pressure of the part through-wall defect tubes depends on the defect depth and length. W a t e r flow rates after the rupture were i n d e p e n d e n t of the flaw types; tubes having 20 60 m m long E D M notches showed similar flow rates regardless of the initial defect depth. A fast pressurization rate g e n e r a t e d a lower burst pressure than the case of a slow pressurization.
05•02031
Extending the operable time of the Syrian MNSR
Ivanov, B. D. et al. Progress in Nuclear Energy, 2004, 45, (2 4), 119 124. As part of an effort to test the ability of current t r a n s p o r t codes to treat reactor core problems w i t h o u t spatial homogenization, the lattice code H E L I O S was employed to perform criticality calculations. The test consists in seven-group calculations of the C5 M O X fuel assembly problem specified by earlier studies and comprises two cases - two and t h r e e - d i m e n s i o n a l geometry. There are four fuel assemblies two with M O X fuel, the other two with UO2 fuel. Each fuel assembly is m a d e up of a 17x17 lattice of square fuel-pin cells. Fuel pin compositions are specified in the B e n c h m a r k Specification, which also provides sevengroup transport-corrected isotropic scattering cross-sections for UO2, the three M O X enrichments, the guide tubes, the fission c h a m b e r and the moderator. This p a p e r preset is the m e t h o d o l o g y employed in solving the C5G7 M O X Fuel Assembly Problem using the transport code H E L I O S .
Li, H.-N. et al. Nuclear Engineering andDesign, 2004, 232, (2), 131 137. The rotational c o m p o n e n t of seismic strong-motion is attracting attention since it is b e c o m i n g evident that it may contribute considerably to the overall response of structures to e a r t h q u a k e motions. This paper presents an improved m e t h o d for calculating the time histories of torsional and rocking c o m p o n e n t s of ground motion corresponding to a set of three recorded o r t h o g o n a l translational components. The m a t h e m a t i c a l m o d e l is based on a detailed r e p r e s e n t a t i o n of soil i m p e d a n c e and contributions of body waves. The d e p e n d e n c e of the angle of wave incidence on the frequency of wave is properly given in the calculation of rotational c o m p o n e n t s with consideration of critical incident angles. N u m e r i c a l results of the torsion and rocking obtained from a set of three recorded translational c o m p o n e n t s are also presented.
05•02033 Homogenization-free reactor core analysiswith general first collision probabilities method
L a n d e s m a n n , A. and de M i r a n d a Batista, E. Nuclear Engineering and Design, 2005, 235, (5), 541 555. The present paper is concerned with the structural safety assessment of a proposed nuclear steel c o n t a i n m e n t shell during a postulated loss-ofcoolant accident scenario. The structural evaluation is p e r f o r m e d using a c o m p u t a t i o n a l second-order refined plastic-hinge method, which is capable of accurately predicting all possible modes of failure in an efficient and computationally less expensive way than the general F E M formulation. A tangent modulus m o d e l and a gradual reduction of the inelastic resistance surface are used to take into account directly the structural strength and stability performances in the e l e m e n t formu-
Poveschenko, T. S. Progress in Nuclear Energy, 2004, 45, (2 4), 143 152. The C5G7 M O X B e n c h - m a r k for current codes has been proposed as a basis to test the ability of current transport codes to teat reactor core problems w i t h o u t spatial homogenization. This is a seven-group form of the C5G7 M O X fuel assembly problem specified by earlier studies. There are four fuel assemblies, two contain UO2 fuel e l e m e n t s and two contain M O X fuel elements. Seven group cross-sections for different kinds of fuel (three e n r i c h m e n t of M O X and UO2), the guide tubes, the
302
Fuel and Energy Abstracts
September 2005
05•02036 Inelastic analysis of cylindrical steel containment vessels under internal accident conditions
05 Nuclear fuels (scientific, technical) lation. The implemented numerical method provides more reliable safety margins and maintainability, exhibiting a more uniform structural safety level than the linear elastic analysis. A simplified non-linear heat transfer model, developed for symmetrical crosssections, is used to determine the steel temperature gradient and to establish a link between the thermo and the mechanical analysis. The load resulting from pressure and temperature thermodynamic calculations, obtained for the accident scenario, are considered in the structural quasi static analysis, so that the structural response can be tracked for the entire duration of the simulated accident.
05•02037
Instrumentation and control system design
Saito, K. et al. Nuclear Engineering and Design, 2004, 233, (1 3), 125 133. The instrumentation and control system of the high temperature engineering test reactor consists of the instrumentation, control equipments and safety protection systems. There are not many differences in the instrumentation and control equipments design between the high temperature engineering test reactor (HTTR) and light water reactors except for some features. Various kinds of R&D of reactor instrumentation were performed taking into account the HTTR operational conditions, and a plant dynamic analysis was carried out considering the operational conditions of the HTTR in order to design the control system. These systems are required to have a high reliability in respect to safety. In the rise-to-power test it was confirmed that the instrumentation has a high reliability and the control system has a high stability and reasonable damped characteristics for various disturbances.
05•02038 Investigations of the SAD design parameters for optimum experimental performance Domangska, G. et al. Annals of Nuclear Energy, 2004, 31, (18), 2127 2138. A project of fast experimental accelerator-driven system called SAD Subcritical Assembly in Dubna containing MOX fuel and driven by 660 MeV proton beam is described and analysed. It is shown by design calculations that the necessity exists for certain modifications, allowing for better reliability of measurements of system time characteristics. Different solutions such as: cadmium separation of the biological concrete shield, admixtures of B203 to the concrete and certain slowing down of neutrons were analysed. Experiments on a bare spallation targets were conducted and the production of radionuclides in the lead target were measured and compared with calculations.
05•02039
Nuclear design
Fujimoto, N. et al. Nuclear Engineering and Design, 2004, 233, (1 3), 23 36. The high-temperature engineering test reactor (HTTR) has been designed for an outlet temperature of 950~'C. That is the highest temperature in the world for a block-type high-temperature gas-cooled reactor (HTGR). The functions of the reactivity control system are determined considering the operational conditions, and the reactivity balance is planned so that the design requirements are fully satisfied. Moreover, the reactivity coefficients are evaluated to confirm the safety characteristics of the reactor. The power distribution in the core was optimized by changing the uranium enrichment to maintain the fuel temperature at less than the limit (1600~'C). Deviation from the optimized distribution due to the burnup of fissile materials was avoided by flattening time-dependent changes in local reactivities. Flattening was achieved by optimizing the specifications of the burnable poisons. The original nuclear design model had to be modified based on the first critical experiments. The Monte Carlo code MVP was also used to predict criticality of the initial core. The predicted excess reactivities are now in good agreement with the experimental results.
05•02040
Overview of HTTR design features
Shiozawa, S. et al. Nuclear Engineering and Design, 2004, 233, (1 3), 11 21. The Japan Atomic Energy Research Institute (JAERI) designed and constructed the high temperature engineering test reactor (HTTR) in order to establish and upgrade the technology basis for the high temperature gas-cooled reactor (HTGR) and develop the technology for high temperature heat applications. The HTTR is a helium-cooled and graphite-moderated HTGR with a thermal power of 30 MW and a maximum reactor outlet coolant temperature of 950c'C. The first criticality was attained on November 10, 1998 and the rated power operation with the reactor outlet coolant temperature of 850~'C was achieved on December 7, 2001. Since 2002, safety demonstration tests simulating anticipated operational occurrences such as decrease of primary coolant flow-rate and reactivity insertion have been carried out to ensure safety during off-normal conditions. From 2005, various irradiation tests for fuels and materials will start. In addition, a hydrogen production test facility will be coupled with the HTTR by
2015 to produce hydrogen by nuclear. The history and future plan, major design features and R&D programs of the HTTR are summarized in this paper.
05•02041 Parameter estimation during a transient application to BWR stability Tambouratzis, T. and Antonopoulos-Domis, M. Annals o f Nuclear Energy, 2004, 31, (18), 2077 2092. The estimation of system parameters is of obvious practical interest. During transient operation, these parameters are expected to change, whereby the system is rendered time-varying and classical signal processing techniques are not applicable. A novel methodology is proposed here, which combines wavelet multi-resolution analysis and selective wavelet coefficient removal with classical signal processing techniques in order to provide short-term estimates of the system parameters of interest. The use of highly overlapping time-windows further monitors the gradual changes in system parameter values. The potential of the proposed methodology is demonstrated with numerical experiments for the problem of stability evaluation of boiling water reactors during a transient.
05•02042
Performance test of HTTR
Nakagawa, S. et al. Nuclear Engineering and Design, 2004, 233, (1 3), 291 300. The high temperature gas-cooled reactor (HTGR) is particularly attractive due to its capability of producing high temperature helium gas and due to its inherent safety characteristics. The high temperature engineering test reactor (HTTR), which is the first HTGR in Japan, was successfully constructed at the Oarai Research Establishment of the Japan Atomic Energy Research Institute. The HTTR achieved full power of 30 MW at a reactor outlet coolant temperature of about 850c'C on 7 December 2001 during the 'rise-to-power tests'. Two kinds of tests were carried out during the 'rise-to-power tests'. One is commissioning test to get operation permit by the government and another is test to confirm a performance of the reactor, heat exchanger, and control system. From the test results of the 'rise-to-power tests' up to 30 MW, the functionality of the reactor and the cooling system were confirmed, and it was also confirmed that an operation of reactor facility could be performed safely.
05•02043 Potential of thorium molten salt reactorsdetailed calculations and concept evolution with a view to large scale energy production Nuttin, A. eta[. Progress in Nuclear Energy, 2005, 46, (1), 77 99. The paper discusses the concept of a thorium molten salt reactor dedicated to future nuclear energy production. The fuel of such reactors being liquid, it can be easily reprocessed to overcome neutronic limits. In the late sixties, the MSBR project showed that breeding is possible with thorium in a thermal spectrum, provided that an efficient pyrochemical reprocessing is added. With tools developed around the Monte Carlo MCNP code, the performance of a MSBR-like reference system with 232Th / 232U fuel was re-evaluated. An important reduction of inventories and induced radiotoxicities was found at equilibrium compared to other fuel cycles, with a doubling time of about 30 years. The study then considered how to start this interesting reference system with the plutonium from PWR spent fuel. Such a transition appears slow and difficult, since it is very sensitive to the fissile quality of the plutonium used. Deployment scenarios of Z3ZTh/ 232U MSBR-like systems from the existing French PWRs demonstrate the advantage of an upstream 232U production in other reactors, allowing a direct start of the MSBR-like systems with 232U. This finally leads to the exploration of alternatives to some MSBR features, for energy production with 232Th/ 232U fuel from the start. Different options were then tested, especially in terms of core neutronics optimization and reprocessing unit adaptation.
05•02044 sheet
Purex co-processing of spent LWR fuels: flow
Zabuno~lu, O. H. and Ozdemir, L. Annals of Nuclear Energy, 2005, 32, (2), 151 162. Purex co-processing of spent LWR fuel is investigated. In purex coprocessing, uranium and plutonium in spent fuel are processed and recovered together as a single stream, while in standard purex reprocessing uranium and plutonium are obtained as separate streams. A two-step (co-decontamination and co-stripping) flow sheet for purex co-processing is devised; concentrations, recoveries and decontamination factors are calculated; and methods to co-convert uranium plutonium nitrate to mixed oxide are reviewed. A closed nuclear fuel cycle in which at no point uranium and plutonium are separated from each other is reached.
05•02045
R&D on core seismic design
Iyoku, T. et al. Nuclear Engineering and Design, 2004, 233, (1 3), 225 234.
Fuel and Energy Abstracts
September 2005
303
fission chambers and m o d e r a t o r are given. Thus this b e n c h m a r k is just a m a t h e m a t i c a l test that allows testing the accuracy of the n e u t r o n t r a n s p o r t equation solution with different methods and codes. In this paper the G e n e r a l First Collision Probabilities M e t h o d ( G F C P M ) is used to analyse the two-dimensional configuration of this b e n c h m a r k . A linear flux approximation is used in the reflector. Different calculation schemes in the reflector region have been used. The o u t p u t results, I~ff and the pin powers have been analysed. The convergence of the results has been analysed both as a function of the subdivision scheme of the reflector region and of the n u m b e r of points in the calculation scheme for general first collision probabilities. C o m p a r i s o n has been carried out for Keff and pin powers both with the reference results (external convergence) and with the results of different approximations of G F C P M (internal convergence).
05•02034 Hydro-thermal-mechanical analysis of thermal fatigue in a mixing tee
A l b a r h o u m , M. Annals of Nuclear Energy, 2004, 31, (18), 2203 2209. The possibility of extending the operable time of the Syrian M N S R is investigated through a t h r e e - d i m e n s i o n a l detailed model of the reactor constructed for this purpose. G o o d a g r e e m e n t between calculated and m e a s u r e d reactor p a r a m e t e r values were obtained for the reactor before modification. The operable time is increased by increasing the initial available excess reactivity. The latter is increased by adding Top Beryllium Shims in the Shim Tray. The increased initial excess reactivity is c o m p e n s a t e d for by increasing control rod worth by substituting the actual c a d m i u m absorber by 5B10 a b s o r b e r . The shut down margin is also e n h a n c e d and safer reactor is obtained.
Chapuliot, S. et al. Nuclear Engineering andDesign, 2005, 235, (5), 575 596. This paper covers work carried out by the C E A to study the m e c h a n i s m s leading to cracking of piping as a result of t h e r m a l loading in flow mixing zones. The main goal of the work is to analyse, by calculation, the t h e r m a l loading caused by t u r b u l e n t mixing in tees and to u n d e r s t a n d the m e c h a n i s m of initiation and propagation of cracks in such components. This work is s u p p o r t e d by IRSN. This t h e r m a l fatigue p h e n o m e n o n is still not fully understood. One of the main obstacles to its u n d e r s t a n d i n g resides in the m u l t i - d o m a i n nature of the loading and associated damage, involving three c o m p l e m e n t a r y scientific disciplines: thermal-hydraulic field, t h e r m o - m e c h a n i c a l field and materials science. This paper describes the approach adopted by the C E A to establish natural m e c h a n i s m s (turbulence, pulsing and instability) which might be the cause of any substantial thermomechanical loading in the piping. A l t h o u g h turbulence may be the cause of the t h e r m a l stripping (presence of high-frequency t h e r m a l fluctuations on the inner surface of the component), it c a n n o t alone explain the p r o p a g a t i o n of deep cracks. The main reason is the 'highpass filter' effect of convection. The wall c a n n o t be subjected to convection-related t h e r m a l fluctuations and frequencies less than the inverse of the turbulence transit time. A straightforward frequencybased analysis of the loading, carried out as a first stage, made it possible to establish the limits of the loading created by these highfrequency events. However, turbulence can give rise to flow instability (such as pulsing) of lower frequency. But this cannot explain everything. The geometry u p s t r e a m of the tee, particularly the sequence of straight sections and bends can, in certain cases, d a m p the pulses or greatly amplify them. The use of suitable thermalhydraulic m o d e l l i n g is discussed in the second part of this article. The final result of the t h e r m o - h y d r o - m e c h a n i c a l link-up on application to the complex 3D geometry of the Civaux unit 1 case (which includes a mixing tee, bends and straight sections) enabled the observations m a d e in this plant case to be highlighted and correlated. One of the originalities of this study is to carry out the overall analysis (thermalhydraulic and thermo-mechanical) with a single computer code, the C A S T 3 M code developed by the CEA.
05•02032 Helios, current coupling collision probability method, applied for solving the NEA C5G7 MOX benchmark
05•02035 Improved approach for obtaining rotational components of seismic motion
05•02030 Evaluation of the burst characteristics for axial notches on SG tubings Hwang, S. S. et al. Nuclear Engineering andDesign, 2004, 232, (2), 139 143. Some events of steam generator tubes have been reported in some nuclear power plants a r o u n d the world. Main causes of the leakage are from various types of corrosion in the s t e a m g e n e r a t o r (SG) tubing. Primary water stress corrosion cracking (PWSCC) of steam generator tubing have occurred in many tubes in Korean plants, and they were repaired using sleeves or plugs. In order to develop p r o p e r repair criteria, it is necessary to ascertain the leak behaviour of the tubings. A high-pressure leak and burst testing system was manufactured. Various types of electro-discharged-machined ( E D M ) notches having different lengths were m a c h i n e d on the outside d i a m e t e r of test tubes to study SG tube behaviour. L e a k rate and ligament rupture pressure as well as the burst pressure were m e a s u r e d for the tubes at r o o m temperature. R u p t u r e pressure of the part through-wall defect tubes depends on the defect depth and length. W a t e r flow rates after the rupture were i n d e p e n d e n t of the flaw types; tubes having 20 60 m m long E D M notches showed similar flow rates regardless of the initial defect depth. A fast pressurization rate g e n e r a t e d a lower burst pressure than the case of a slow pressurization.
05•02031
Extending the operable time of the Syrian MNSR
Ivanov, B. D. et al. Progress in Nuclear Energy, 2004, 45, (2 4), 119 124. As part of an effort to test the ability of current t r a n s p o r t codes to treat reactor core problems w i t h o u t spatial homogenization, the lattice code H E L I O S was employed to perform criticality calculations. The test consists in seven-group calculations of the C5 M O X fuel assembly problem specified by earlier studies and comprises two cases - two and t h r e e - d i m e n s i o n a l geometry. There are four fuel assemblies two with M O X fuel, the other two with UO2 fuel. Each fuel assembly is m a d e up of a 17x17 lattice of square fuel-pin cells. Fuel pin compositions are specified in the B e n c h m a r k Specification, which also provides sevengroup transport-corrected isotropic scattering cross-sections for UO2, the three M O X enrichments, the guide tubes, the fission c h a m b e r and the moderator. This p a p e r preset is the m e t h o d o l o g y employed in solving the C5G7 M O X Fuel Assembly Problem using the transport code H E L I O S .
Li, H.-N. et al. Nuclear Engineering andDesign, 2004, 232, (2), 131 137. The rotational c o m p o n e n t of seismic strong-motion is attracting attention since it is b e c o m i n g evident that it may contribute considerably to the overall response of structures to e a r t h q u a k e motions. This paper presents an improved m e t h o d for calculating the time histories of torsional and rocking c o m p o n e n t s of ground motion corresponding to a set of three recorded o r t h o g o n a l translational components. The m a t h e m a t i c a l m o d e l is based on a detailed r e p r e s e n t a t i o n of soil i m p e d a n c e and contributions of body waves. The d e p e n d e n c e of the angle of wave incidence on the frequency of wave is properly given in the calculation of rotational c o m p o n e n t s with consideration of critical incident angles. N u m e r i c a l results of the torsion and rocking obtained from a set of three recorded translational c o m p o n e n t s are also presented.
05•02033 Homogenization-free reactor core analysiswith general first collision probabilities method
L a n d e s m a n n , A. and de M i r a n d a Batista, E. Nuclear Engineering and Design, 2005, 235, (5), 541 555. The present paper is concerned with the structural safety assessment of a proposed nuclear steel c o n t a i n m e n t shell during a postulated loss-ofcoolant accident scenario. The structural evaluation is p e r f o r m e d using a c o m p u t a t i o n a l second-order refined plastic-hinge method, which is capable of accurately predicting all possible modes of failure in an efficient and computationally less expensive way than the general F E M formulation. A tangent modulus m o d e l and a gradual reduction of the inelastic resistance surface are used to take into account directly the structural strength and stability performances in the e l e m e n t formu-
Poveschenko, T. S. Progress in Nuclear Energy, 2004, 45, (2 4), 143 152. The C5G7 M O X B e n c h - m a r k for current codes has been proposed as a basis to test the ability of current transport codes to teat reactor core problems w i t h o u t spatial homogenization. This is a seven-group form of the C5G7 M O X fuel assembly problem specified by earlier studies. There are four fuel assemblies, two contain UO2 fuel e l e m e n t s and two contain M O X fuel elements. Seven group cross-sections for different kinds of fuel (three e n r i c h m e n t of M O X and UO2), the guide tubes, the
302
Fuel and Energy Abstracts
September 2005
05•02036 Inelastic analysis of cylindrical steel containment vessels under internal accident conditions
05 Nuclear fuels (scientific, technical) lation. The implemented numerical method provides more reliable safety margins and maintainability, exhibiting a more uniform structural safety level than the linear elastic analysis. A simplified non-linear heat transfer model, developed for symmetrical crosssections, is used to determine the steel temperature gradient and to establish a link between the thermo and the mechanical analysis. The load resulting from pressure and temperature thermodynamic calculations, obtained for the accident scenario, are considered in the structural quasi static analysis, so that the structural response can be tracked for the entire duration of the simulated accident.
05•02037
Instrumentation and control system design
Saito, K. et al. Nuclear Engineering and Design, 2004, 233, (1 3), 125 133. The instrumentation and control system of the high temperature engineering test reactor consists of the instrumentation, control equipments and safety protection systems. There are not many differences in the instrumentation and control equipments design between the high temperature engineering test reactor (HTTR) and light water reactors except for some features. Various kinds of R&D of reactor instrumentation were performed taking into account the HTTR operational conditions, and a plant dynamic analysis was carried out considering the operational conditions of the HTTR in order to design the control system. These systems are required to have a high reliability in respect to safety. In the rise-to-power test it was confirmed that the instrumentation has a high reliability and the control system has a high stability and reasonable damped characteristics for various disturbances.
05•02038 Investigations of the SAD design parameters for optimum experimental performance Domangska, G. et al. Annals of Nuclear Energy, 2004, 31, (18), 2127 2138. A project of fast experimental accelerator-driven system called SAD Subcritical Assembly in Dubna containing MOX fuel and driven by 660 MeV proton beam is described and analysed. It is shown by design calculations that the necessity exists for certain modifications, allowing for better reliability of measurements of system time characteristics. Different solutions such as: cadmium separation of the biological concrete shield, admixtures of B203 to the concrete and certain slowing down of neutrons were analysed. Experiments on a bare spallation targets were conducted and the production of radionuclides in the lead target were measured and compared with calculations.
05•02039
Nuclear design
Fujimoto, N. et al. Nuclear Engineering and Design, 2004, 233, (1 3), 23 36. The high-temperature engineering test reactor (HTTR) has been designed for an outlet temperature of 950~'C. That is the highest temperature in the world for a block-type high-temperature gas-cooled reactor (HTGR). The functions of the reactivity control system are determined considering the operational conditions, and the reactivity balance is planned so that the design requirements are fully satisfied. Moreover, the reactivity coefficients are evaluated to confirm the safety characteristics of the reactor. The power distribution in the core was optimized by changing the uranium enrichment to maintain the fuel temperature at less than the limit (1600~'C). Deviation from the optimized distribution due to the burnup of fissile materials was avoided by flattening time-dependent changes in local reactivities. Flattening was achieved by optimizing the specifications of the burnable poisons. The original nuclear design model had to be modified based on the first critical experiments. The Monte Carlo code MVP was also used to predict criticality of the initial core. The predicted excess reactivities are now in good agreement with the experimental results.
05•02040
Overview of HTTR design features
Shiozawa, S. et al. Nuclear Engineering and Design, 2004, 233, (1 3), 11 21. The Japan Atomic Energy Research Institute (JAERI) designed and constructed the high temperature engineering test reactor (HTTR) in order to establish and upgrade the technology basis for the high temperature gas-cooled reactor (HTGR) and develop the technology for high temperature heat applications. The HTTR is a helium-cooled and graphite-moderated HTGR with a thermal power of 30 MW and a maximum reactor outlet coolant temperature of 950c'C. The first criticality was attained on November 10, 1998 and the rated power operation with the reactor outlet coolant temperature of 850~'C was achieved on December 7, 2001. Since 2002, safety demonstration tests simulating anticipated operational occurrences such as decrease of primary coolant flow-rate and reactivity insertion have been carried out to ensure safety during off-normal conditions. From 2005, various irradiation tests for fuels and materials will start. In addition, a hydrogen production test facility will be coupled with the HTTR by
2015 to produce hydrogen by nuclear. The history and future plan, major design features and R&D programs of the HTTR are summarized in this paper.
05•02041 Parameter estimation during a transient application to BWR stability Tambouratzis, T. and Antonopoulos-Domis, M. Annals o f Nuclear Energy, 2004, 31, (18), 2077 2092. The estimation of system parameters is of obvious practical interest. During transient operation, these parameters are expected to change, whereby the system is rendered time-varying and classical signal processing techniques are not applicable. A novel methodology is proposed here, which combines wavelet multi-resolution analysis and selective wavelet coefficient removal with classical signal processing techniques in order to provide short-term estimates of the system parameters of interest. The use of highly overlapping time-windows further monitors the gradual changes in system parameter values. The potential of the proposed methodology is demonstrated with numerical experiments for the problem of stability evaluation of boiling water reactors during a transient.
05•02042
Performance test of HTTR
Nakagawa, S. et al. Nuclear Engineering and Design, 2004, 233, (1 3), 291 300. The high temperature gas-cooled reactor (HTGR) is particularly attractive due to its capability of producing high temperature helium gas and due to its inherent safety characteristics. The high temperature engineering test reactor (HTTR), which is the first HTGR in Japan, was successfully constructed at the Oarai Research Establishment of the Japan Atomic Energy Research Institute. The HTTR achieved full power of 30 MW at a reactor outlet coolant temperature of about 850c'C on 7 December 2001 during the 'rise-to-power tests'. Two kinds of tests were carried out during the 'rise-to-power tests'. One is commissioning test to get operation permit by the government and another is test to confirm a performance of the reactor, heat exchanger, and control system. From the test results of the 'rise-to-power tests' up to 30 MW, the functionality of the reactor and the cooling system were confirmed, and it was also confirmed that an operation of reactor facility could be performed safely.
05•02043 Potential of thorium molten salt reactorsdetailed calculations and concept evolution with a view to large scale energy production Nuttin, A. eta[. Progress in Nuclear Energy, 2005, 46, (1), 77 99. The paper discusses the concept of a thorium molten salt reactor dedicated to future nuclear energy production. The fuel of such reactors being liquid, it can be easily reprocessed to overcome neutronic limits. In the late sixties, the MSBR project showed that breeding is possible with thorium in a thermal spectrum, provided that an efficient pyrochemical reprocessing is added. With tools developed around the Monte Carlo MCNP code, the performance of a MSBR-like reference system with 232Th / 232U fuel was re-evaluated. An important reduction of inventories and induced radiotoxicities was found at equilibrium compared to other fuel cycles, with a doubling time of about 30 years. The study then considered how to start this interesting reference system with the plutonium from PWR spent fuel. Such a transition appears slow and difficult, since it is very sensitive to the fissile quality of the plutonium used. Deployment scenarios of Z3ZTh/ 232U MSBR-like systems from the existing French PWRs demonstrate the advantage of an upstream 232U production in other reactors, allowing a direct start of the MSBR-like systems with 232U. This finally leads to the exploration of alternatives to some MSBR features, for energy production with 232Th/ 232U fuel from the start. Different options were then tested, especially in terms of core neutronics optimization and reprocessing unit adaptation.
05•02044 sheet
Purex co-processing of spent LWR fuels: flow
Zabuno~lu, O. H. and Ozdemir, L. Annals of Nuclear Energy, 2005, 32, (2), 151 162. Purex co-processing of spent LWR fuel is investigated. In purex coprocessing, uranium and plutonium in spent fuel are processed and recovered together as a single stream, while in standard purex reprocessing uranium and plutonium are obtained as separate streams. A two-step (co-decontamination and co-stripping) flow sheet for purex co-processing is devised; concentrations, recoveries and decontamination factors are calculated; and methods to co-convert uranium plutonium nitrate to mixed oxide are reviewed. A closed nuclear fuel cycle in which at no point uranium and plutonium are separated from each other is reached.
05•02045
R&D on core seismic design
Iyoku, T. et al. Nuclear Engineering and Design, 2004, 233, (1 3), 225 234.
Fuel and Energy Abstracts
September 2005
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