Progress in Nuclear Energy

Published by Elsevier
Print ISSN: 0149-1970
Publications
In the framework of joint effort between the Nuclear Energy Agency (NEA) of OECD, the United States Department of Energy (US DOE), and the Commissariat a l'Energie Atomique (CEA), France a coupled three-dimensional (3D) thermal-hydraulics/neutron kinetics benchmark for VVER-1000 was defined. The benchmark consists of calculation of a pump start-up experiment labelled V1000CT-1 (Phase 1), as well as a vessel mixing experiment and main steam line break (MSLB) transient labelled V1000CT-2 (Phase 2), respectively. The reference nuclear plant is Kozloduy-6 in Bulgaria. The overall objective is to assess computer codes used in the analysis of VVER-1000 reactivity transients. A specific objective is to assess the vessel mixing models used in system codes. Plant data are available for code validation consisting of one experiment of pump start-up (V1000CT-1) and one experiment of steam generator isolation (V1000CT-2). The validated codes can be used to calculate asymmetric MSLB transients involving similar mixing patterns. This paper summarizes a comparison of CATHARE and TRAC-PF1 system code results for V1000CT-1, Exercise 1, which is a full plant point kinetics simulation of a reactor coolant system (RCS) pump start-up experiment. The reference plant data include integral and sector average parameters. The comparison is made from the point of view of vessel mixing and full system simulation. CATHARE used a six-sector multiple 1D vessel thermal-hydraulic model with cross flows and TRAC used a six-sector, 18-channel coarse-mesh 3D vessel model. Good agreement in terms of integral parameters and inter-loop mixing is observed.
 
This paper presents the results of performance and radiological analyses of a space reactor power system to support space-based, radar satellites in a 1000–3000 km orbit for global civilian air and ocean traffic control. The power system with six primary and secondary loops to avoid single point failures in reactor cooling and energy conversion employs a sectored, liquid NaK-78 cooled fission reactor that has a negative temperature reactivity feedback and SiGe thermoelectric energy conversion modules, nominally generating 37.1 kWe for up to 6 years. The reactor nominally generates 1183.6 kWth at an exit temperature of ∼992 K. Thermoelectric–electromagnetic (TEM) pumps circulate the liquid NaK-78 in the primary and secondary loops and passively remove the decay heat from the reactor core after shutdown. The system parameters during the startup and shutdown transients and nominal steady-state operation are calculated. The effects of the period of incrementally inserting external reactivity on the system parameters during nominal operation are also investigated. The radioactivity buildup in the reactor during nominal operation up to 6 years and the decay in radioactivity after shutdown and as a function of storage time in orbit up to 1000 years are calculated.
 
In the framework of joint effort between the Nuclear Energy Agency (NEA) of OECD, the United States Department of Energy (US DOE), and the Commissariat a l'Enerige Atomique (CEA), France a coupled three-dimensional (3D) thermal-hydraulics/neutron kinetics benchmark was defined. The overall objective of OECD/NEA V1000CT benchmark is to assess computer codes used in analysis of VVER-1000 reactivity transients where mixing phenomena (mass flow and temperature) in the reactor pressure vessel are complex. Original data from the Kozloduy-6 Nuclear Power Plant are available for the validation of computer codes: one experiment of pump start-up (V1000CT-1) and one experiment of steam generator isolation (V1000CT-2). Additional scenarios are defined for code-to-code comparison. As a 3D core model is necessary for a best-estimate computation of all the scenarios of the V1000CT benchmark, all participants were asked to develop their own core coupled 3D thermal-hydraulics/neutron kinetics models using the data available in the benchmark specifications and a common cross-section library. The first code-to-code comparisons based on the V1000CT-1 Exercise 2 specifications exhibited unacceptable discrepancies between two sets of results. The present paper focuses on the analysis of the observed discrepancies. The VVER-1000 3D neutron kinetics models are based on cross-section data homogenized on the assembly level. The cross-section library, provided as part of the benchmark specifications, thus consists in a set of parameterized two group cross sections representing the different assemblies and the reflectors. The origin of the observed large discrepancies was found mainly to lie in the methods used to solve the diffusion equation. The VVER reflector properties were also found to enhance discrepancies by increasing flux gradients at the core/reflector interface thus highlighting more the difficulties in some codes to handle high exponential flux gradients. This paper summarizes the different steps applied to analyze the neutronic codes and their predictions as well as the impact of cross-section generation procedures.
 
This paper presents the development and application of methodology used in analytical validation of emergency operating procedures (EOPs) for Kozloduy Nuclear Power Plant (KNPP), VVER-1000/V320 units. EOPs provide generic guidance to a reactor plant operator in maneuvering the plant to a safe, stable condition in the event of an unexpected plant transient or emergency. These procedures have been analytically validated in order to provide technical justification that the prescribed operator actions are reasonable, effective and prudent. This evaluation is accomplished by systematically evaluating the procedures using specialized thermal-hydraulic computer codes designed for nuclear reactor plant simulation. Thermal-hydraulic computer code calculations have been performed to simulate the symptoms presented to the operator to diagnose challenges to the critical safety functions (CSFs). The accomplished emergency operating procedures' (EOPs') analyses are designed to provide the response of monitored plant parameters to identify symptoms available to the operators, timing of the loss of critical safety functions and timing of operator actions to avoid the loss of critical safety functions or core damage. The principal acceptance criteria for EOPs are averting the onset of core damage.
 
The purpose of this research was to determine the effect of moderated heterogeneous subassemblies located in the core of a sodium-cooled fast reactor on minor actinide (MA) destruction rates over the lifecycle of the core. Additionally, particular emphasis was placed on the power peaking of the pins and the assemblies with the moderated targets as compared to standard unmoderated heterogeneous targets and a core without MA targets present. Power peaking analysis was performed on the target assemblies and on the fuel assemblies adjacent to the targets. The moderated subassemblies had a marked improvement in the overall destruction of heavy metals in the targets. The design with acceptable power peaking results had a 12.25% greater destruction of heavy metals than a similar ex-core unmoderated assembly. The increase in minor actinide destruction was most evident with americium where the moderated assemblies reduced the initial amount to less than 3% of the initial loading over a period of five years core residency. In order to take advantage of the high minor actinide destruction and minimize the power peaking effects, a hybrid scenario was devised where the targets resided ex-core in a moderated assembly for the first 506.9 effective full power days (EFPDs) and were moved to an in-core arrangement with the moderated targets removed for the remainder of the lifecycle. The hybrid model had an assembly and pin power peaking of less than 2.0 and a higher heavy metal and minor actinide destruction rate than the standard unmoderated heterogeneous targets either in-core or ex-core. The hybrid model has a 54.5% greater Am reduction over the standard ex-core model. It also had a 27.8% greater production of Cm and a 41.5% greater production of Pu than the standard ex-core model. The radiotoxicity of the targets in the hybrid design was 20% less than the discharged standard ex-core targets.
 
In the light of the sustained development in computer technology, the possibilities of code capabilities have been enlarged substantially. Consequently, advanced safety evaluations and design optimizations, which were not possible few years ago, can now be performed. Nowadays, it becomes possible to switch to new generation of computational tools, namely, coupled code (CC) technique. The application of such method is mandatory for the analysis of transient events where strong coupling between the core neutronics and the primary circuit thermal-hydraulics exits, and more especially when asymmetrical processes take place in the core leading to local space-dependent power generation. Through the current study, a demonstration of the maturity level achieved in the calculation of 3-D core performance during complex accident scenarios is emphasized. The study is followed by a typical application through which the main features and limitations of this technique are discussed.
 
In this paper the safety performance of 25–100 MWe Pb–Bi cooled long life fast reactors based on three types of fuels: MOX, nitride and metal is compared and discussed. In the fourth generation NPP paradigm, especially for Pb–Bi cooled fast reactors, inherent safety capability is necessary against some standard accidents such as unprotected loss of flow (ULOF), unprotected rod run-out transient over power (UTOP), unprotected loss of heat sink (ULOHS). Selection of fuel type will have important impact on the overall system safety performance.The results of safety analysis of long life Pb–Bi cooled fast reactors without on-site fuelling using nitride, MOX and metal fuel have been performed. The reactors show the inherent safety pattern with enough safety margins during ULOF and UTOP accidents. For MOX fuelled reactors, ULOF accident is more severe than UTOP accident while for nitride fuelled cores UTOP accident may push power much higher than that comparable MOX fuelled cores.
 
With its long half-life (5730 years) and high mobility in the environment, 14C is a radionuclide of considerable interest in nuclear power production. Carbon-14 is present in virtually all parts of nuclear reactor primary system and has a high production rate. It is released to the environment through gaseous and liquid discharges and though the disposal of solid radioactive waste. This paper summarizes existing scientific understanding of 14C issues surrounding nuclear power production. Two main purposes of the paper are: (1) To provide the basic/up-to-date understanding of the life cycle of 14C, starting from its production in reactors, to eventually its transport and its potential incorporation in natural cycles; (2) To present the technical issues in current 14C waste management. The emphasis of the paper is on Light Water Reactors (LWRs, which include Pressured Water Reactors-PWRs, and Boiling Water Reactors-BWRs) and Heavy Water Reactors (HWRs-CANDU type reactors). Major issues with 14C in HTGR are also addressed.
 
NUREG-1465 was a major change to the preceding figures defining in-containment source term. Since then, the most important research venture in the arena of severe accidents has been the PHEBUS-FP project. Experimental data and interpretations brought in along the course of the project have highlighted similarities and discrepancies with those insights given in the NUREG-1465.This paper sets comparisons between NUREG-1465 and PHEBUS-FP in three key aspects: the release of radionuclides into containment, the in-containment aerosol behaviour and, finally, iodine chemical behaviour in the containment. The experimental basis of discussions is the FPT0, FPT1, FPT2 and FPT3 series, although the latter cannot be openly addressed yet. Similarities have been found regarding qualitative gap and early in-vessel releases, quantitative net release of noble gases and iodine releases, dominance of sedimentation as natural removal mechanism for in-containment aerosols, etc. Nonetheless, PHEBUS-FP data have also drawn attention to major discrepancies with respect to NUREG-1465. Examples are cesium and tellurium releases and possible massive iodine release under specific conditions. But, in addition, PHEBUS-FP has brought new insights, such as potential formation of in-sump iodine precipitates or the need of revisiting NUREG-1465 element grouping.
 
An experimental campaign was conducted to assess the neutron dose field inside the biological shield of the PETrace cyclotron at the S. Orsola-Malpighi hospital in Bologna. The results of the survey were analyzed to confirm that neutrons are emitted isotropically, thus validating the hypothesis of nuclear evaporation, as opposed to cascade, as the dominant mechanism for neutron emission.
 
In 1942, a sub-group of scientists collaborated to develop a nuclear reactor theory. The members of the group comprised an international mix of Canadian, Britons and Americans. The group worked with utmost secrecy. The results of their work were issued as Canadian National Research Council reports with the prefix MT (Montreal Theory). Between 1943 and 1946, about 80 reports were written. The theory was the fundamental in the later design and construction of the Canadian NRX reactor, which was a very successful research tool.
 
The method of characteristics (MOC) code CRX solves the three-dimensional transport problem by the 2D/1D fusion method, in which MOC is used in radial 2D calculation and SN-like methods are used in axial 1D calculation. The CRX code was used to provide the solutions for the three-dimensional OECD benchmark problem C5G7 MOX and the results were submitted to OECD and documented in the OECD report. This paper provides a description of the 2D/1D fusion method for three-dimensional transport calculation and presents additional work implementing several improved features performed since the submittal of benchmark solutions, with comparison of the results.
 
In the year 2002 and 2003 the Japanese Ministry of Education, Culture, Sports, Science and Technology (MEXT) started the “Priority Assistance for the Formation of Worldwide Renowned Centers of Research — The 21st Century Center of Excellence (COE) Program”, which is planned to continue for 5 years.A program proposed by Tokyo Institute of Technology “Innovative Nuclear Energy Systems for Sustainable Development of the World” simply called as COE-INES was selected as only one program in nuclear engineering field.The program consists of research, education and international collaboration. The research will be performed on the innovative nuclear energy systems, which include innovative nuclear reactors and innovative fuel cycles. The research on innovative nuclear reactors does not cover only reactor design studies but also its utilization systems such as hydrogen production. Both free thinking and overall vision are taken on the research. They are stressed on education also.In the education program (COE-INES Captainship Program) by integrating research with education, we will foster creative researchers and engineers. The program also provides lectures at the professional engineer level, and also various opportunities to cultivate internationalism.We believe these ideas are occupied by many scientists in various countries. Then we have a plan to promote the international collaboration for research and education on innovative nuclear energy systems.
 
Energy security is vital for the steady growth of the world's welfare and economy and although many novel non-nuclear energy sources are being explored, much less attention is given to nuclear energy. In developing this source, safety, environment protection, and non-proliferation are essential considerations. I have proposed establishing deep underground nuclear parks where not only energy production, but also the processing and transportation of fuel can be carried out in a well protected small area; in the near future they might be operated under international supervision to ensure non-proliferation. The quantum physics on which modern technologies such as nanotechnology, and biotechnology are based, offer a sound foundation. The mathematical and physical technologies developed in the fields of nuclear engineering will provide the fundamental educational basis for such 21st century science and technology.
 
Nuclear energy must compete against other energy technologies in the 21st century. It must be economical and it must be proven that it fulfills the conditions for sutainability. This means that the requirements of — no short term depletion of resources — extremely low emission of noxious or radioactive substances to the environment — extremely low release of radioactivity from a nuclear plant in case of the most severe accidents and — the present very long term problem of high active waste must be transformed into a few hundred years problem through destruction of plutonium, transmutation of the minor actinides and the most important very long lived fission products.
 
The economic growth of recent Asia is rapid, and the GDP and the energy consumption growth rate are about 8–10% in China and India. The energy consumption forecast of Asia in this century was estimated based on the GDP growth rate by Goldman Sachs. As a result, about twice in India and Association of South East Asian Nations (ASEAN) and about 1.5 times in China of SRES B (Special Report on Emission Scenarios) are forecasted. The simulation was done by Grape Code to analyze the impact of energy increase in Asia. As for the nuclear plant in Asia, it is expected 1500 GWe in 2050 and 2000 GWe in 2100, in the case of the environmental constrain. To achieve this nuclear utilization, there are two important aspects, technically and institutionally.A.Development of the CANDLE core and/or the Breed and Burn core.B.The establishment of the stable nuclear fuel supply system like “Asian nuclear fuel supply organization”.
 
The paper is addressed to the problem how to impart nuclear materials the properties of inherent protective barriers, which are able to impede substantially or even make it practically impossible to extract and use highly enriched uranium fuel for non-energy production purposes. It is demonstrated that highly enriched uranium, doped with 1% 232U, acquires inherent proliferation resistance, at least, at the level of uranium enriched up to 20% 235U. This property is well supplemented with previous results (Shmelev, A.N., Kulikov, G.G., 1997. About neutron-physical features of the modified (denatured) fuel cycles. Communications of Higher Schools, Nuclear Power Engineering, No. 6, pp. 42–48.) demonstrating improvement of neutron-multiplying properties in such a fuel. Another important circumstance consists in the fact that application of uranium enrichment technologies for so denatured uranium fuel is substantially hampered.
 
A survey is made of integral experiments useful for testing thorium and 233U nuclear data in thermal reactor applications. Emphasis is on homogeneous 233UH2O criticals and simple, water-moderated 233U-thorium and 235U-thorium lattice experiments. Thorium-233U-graphite experiments are also discussed briefly. Although the available experiments provide a fairly consistent test of important nuclear data, their accuracy and scope leave much to be desired. In detailed Monte Carlo analyses, ENDF/B-IV data are found to perform reasonably well. Adequate (though partly fortuitous) agreement is found with integral measurements of thorium resonance capture in lattices. A new, harder fission spectrum for 233U can correct the principal discrepancy observed with ENDF/B-IV, a bias trend in Keff attributed to an underprediction of leakage.
 
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Today there is no well-established theoretical model to predict the fission delayed neutron yield vd with the required accuracy. In this field the recommended data result from the rare experimental data analysis or from purely phenomenological or semi-phenomenological models. There is another source of valuable information: the related integral data or βeff- data. In this report we demonstrate, via a careful analysis of the experimental methods leading to revisited experimental βeff values and associated uncertainties, that for the major nuclei the vd evaluated data are of acceptable quality. For U-235 U-238 and Pu-239 we recommend vd values for the thermal and the fast reactor ranges which have been obtained from a statistical consistent adjustment to the βeff data. In the course of this study we show that the energy dependence of vd, suspected from a physics point of view, probably exists with a different magnitude according to the nucleus. Concerning the major nuclei it is of negligible importance for the applications. The improvement of the higher Pu isotopes and minor actinides data is the main motivation for developing the theoretical investigations of the delayed neutron generation mechanism at the same level as the necessary experimental activity.
 
The energy dependence of the relative abundances and periods has been measured for neutron induced fission of the main fuel nuclides, 235U and 239Pu in the energy range from epi-thermal to 5 MeV and 238U in the energy range from 1 to 5 MeV. The efforts undertaken in improving the experimental and data processing procedures made it possible to improve the accuracy of the delayed neutron parameters determining their time-dependent behavior. In terms of the average half-life of delayed neutron precursors — the value that unequivocally characterizes the particular fissioning system — the real scale of the changes in the relative abundances and periods of delayed neutrons were determined.
 
Benchmarking work was recently performed for the issue of molten corium concrete interaction (MCCI). A synthesis is given here. It concerns first the 2D CCI-2 test with a homogeneous pool and a limestone concrete, which was used for a blind benchmark. Secondly, the COMET-L2 and COMET-L3 2D experiments in a stratified configuration were used as a post-test (L2) and a blind-test (L3) benchmark. More details are given here for the recent benchmark considering a matrix of four reactor cases, with both a homogeneous and a stratified configuration, and with both a limestone and a siliceous concrete. A short overview is given on the different models used in the codes, and the consistency between the benchmark actions on experiments and reactor situations is discussed. Finally, the major uncertainties concerning MCCI are also pointed out.
 
The 3-D extension C5G7 MOX benchmark problems have been solved by CHAPLET-3D code which is based on the idea of dynamic linkage of the multi-plane method of characteristics solutions. The benchmark results show that CHAPLET-3D code gives quite accurate solutions for eigenvalue and pin power distribution in comparison with those of the reference Monte Carlo calculations.
 
A new numerical approach for modeling of multiphase mixing during melt jet/droplet fragmentation process is developed. Melt or debris movements are simulated by a particle transport model in a Lagrangian formulation, while thermohydraulic conditions of the surrounding medium are obtained from solution of the Navier-Stokes and energy-conservation equations written in an Eulerian formulation. The Lagrangian and the Eulerian solutions are coupled and advanced in time, with source terms included to model the interactions between the particle and the continuum phases. The method is validated against isothermal solid-sphere, and drop fragmentation experiments. It is found that the model is capable of describing the evolution of the melt-coolant multiphase mixing process with reasonable accuracy. The method is then applied to investigate fragmentation of a continuous jet. Effects of variations in jet/coolant velocities, and of coolant thermophysical properties are analyzed, with particular emphasis on their implications for the fragmentation and mixing processes.
 
Three-dimensional (3D) transport benchmark problems for simple geometries with void region were proposed at the OECD/NEA in order to check the accuracy of deterministic 3D transport programs. The exact total fluxes by the analytical method are given for the pure absorber cases, and Monte Carlo values are given for the 50% scattering cases as the reference values. The total fluxes of all contributions to the present benchmark problems calculated by the 3D transport programs, TORT, TORT with FNSUNCL3, PARTISN, PENTRAN, IDT, MCCG3D, EVENT and ARDRA are compared in figures.
 
We present the solutions for the set of three-dimensional radiation transport Benchmark problems obtained with the TORT transport code using its three optional methods: Theta Weighted (θW), Linear Nodal (LN), and Linear Characteristic (LC). Only the cases with 50% scattering are presented in this paper since the nonscattering cases are bound to suffer severe ray effects. By solving the problems on a sequence of refine meshes we illustrate that for some points defined in the benchmarks the solution converges with mesh refinement. However, the solution at most points does not converge with mesh refinement, and we illustrate that this is a consequence of ray effects in the void region. Also, we compare TORT's solutions to the Monte Carlo reference solution and observe that even when TORT's solution converges with mesh refinement, it usually does not converge to the Monte Carlo reference. This behavior also results from ray effects, and therefore we conjecture it will appear in varying degrees in all discrete ordinates accurate solutions because ray effects exist in the exact solution of the discrete ordinates equations. While this result is disappointing from the benchmarking point of view, it bodes well for TORT's ability to produce highly accurate solutions to the discrete ordinates approximation. Eliminating ray effects requires extensions of the solution algorithm, e.g. via a first collision source, while preserving the desirable features of the discrete ordinates methodology.
 
The best-estimate coupled neutronic/thermal-hydraulics code, SIMULATE-3K (S3K), is used by many utilities, research institutes, and regulatory authorities in Europe for performing BWR stability analysis. Analysis of many measured BWR stability tests (often performed in European BWRs) provides the basis for the validation for stability parameter calculations (decay ratio and natural frequency) with S3K. This paper summarizes part of the extensive validation database for the code, and discusses the influence of fuel pin model parameters on the stability results.
 
On the basis of pressure fluctuation measurements in some primary circuit loops at 2nd Unit of Kola NPP with VVER-440 type reactors, the shapes of acoustic standing waves (ASW) were determined at frequencies corresponding to four minimal oscillation eigenfrequencies in the primary circuit coolant. On identification of the ASW modes and properties, experimental results based on six circulating loops in symmetric arrangement allowed determination of the three-dimensional space structure of the wave nodes and antinodes inside and outside of the reactor vessel (RV). As part of this analysis, the geometric features of the primary circuit that caused the formation of these standing waves were identified. Differences in each ASW shape were shown to cause different individual effects on the neutron field in the reactor core and on fuel assembly vibration. This has been partially confirmed by ex-core neutron ionization chamber noise analysis. One type of ASW, possessing an antinode inside the RV, can be used for measurement of the pressure coefficient of reactivity. However, this must be done with care to avoid the potential for incorrect results in some cases. The results presented in this paper can be readily extended to other VVER type reactors with both odd and even number of loops.
 
KUES '95 continuously monitors the noise level picked up by its sensors and records any signal burst that occurs. A new technique enables the system to identify and differentiate automatically between noise associated with normal operation and signals generated by loose or detached parts, etc. This makes it possible for false alarms to be essentially ruled out. Following automatic detection, the system also performs functions which previously required the efforts of specialist personnel. These include not only locating of loose and detached parts and determination of their mass, but also regular generation of status reporst as well as the printing out of comprehensive noise signal documentation and trend plots.The result is a significant improvement in performance combined with a considerable reduction in the time and effort involved in operating the system.
 
The “zero dynamic material buckling method” of measurement of the thermal neutron macroscopic absorption cross section of small samples is presented. Theoretical principles of the method have been elaborated in the one-velocity diffusion approach in which the thermal neutron parameters used have been averaged over a modified Maxwellian. In consecutive measurements the inestigated sample is enveloped in shells of a known moderator of varying thickness and irradiated with a pulsed beam of fast neutrons. The neutrons are slowed-down in the system and a die-away rate of escaping thermal neutrons is measured. The decay constant vs. thickness of the moderator creates the experimental curve. The absorption cross section of the unknown sample is found from the intersection of this curve with the theoretical one. A theoretical curve is calculated for the two-region bounded system (sample-moderator). For the inner medium (sample) the dynamic material buckling is assumed to zero. The method does not use any reference absorption standard and is independent of the transport cross section of the measured sample.Another faster and cheaper procedure was developed in the frame of the method for its routine applications. This method was established through a careful consideration of all experimental results obtained till now. It requires a single die-away measurement performed with only one size of the Plexiglas moderator. The volume of the sample (fluid or crushed material) in both methods is about 170 cm3. The standard deviation for the measured mass absorption cross section of rock samples is in the range of 4 to 20 per cent of the measured value and for brines is of the order of 0.5 per cent.
 
It is important to understand the heat transfer deterioration (HTD) phenomenon for specifying cladding temperature limits in the fuel assembly design of supercritical water-cooled reactor (SCWR). In this study, a numerical investigation of heat transfer in supercritical water flowing through vertical tube with high mass flux and high heat flux is performed by using six low-Reynolds number turbulence models. The capabilities of the addressed models in predicting the observed phenomena of experimental study are shortly analyzed. Mechanisms of the effect of flow structures and fluid properties on heat transfer deterioration phenomenon are also discussed. Numerical results have shown that the turbulence is significantly suppressed when the large-property-variation region spreads to the buffer layer near the wall region, resulting in heat transfer deterioration phenomenon. The property variations of dynamic viscosity and specific heat capacity in supercritical water can impair the deterioration in heat transfer, while the decrease of thermal conductivity contributes to the deterioration.
 
An overview of the most significant studies in the last 35 years of partitioning and transmutation of commercial light water reactor spent fuel is given. Recent Accelerator-based Transmutation of Waste (ATW) systems are compared with liquid-fuel thermal reactor systems that accomplish the same objectives. If no long-lived fission products (e.g., 99Tc and 129I) are to be burned, under ideal circumstances the neutron balance in an ATW system becomes identical to that for a thermal reactor system. However, such a reactor would need extraordinarily rapid removal of internally-generated fission products to remain critical at equilibrium without enriched feed. The accelerator beam thus has two main purposes (1) the burning of long-lived fission products that could not be burned in a comparable reactor's margin (2) a relaxing of on-line chemical processing requirements without which a reactor-based system cannot maintain criticality. Fast systems would require a parallel, thermal ATW system for long-lived fission product transmutation. The actinide-burning part of a thermal ATW system is compared with the Advanced Liquid Metal Reactor (ALMR) using the well-known Pigford-Choi model. It is shown that the ATW produces superior inventory reduction factors for any near-term time scale.
 
The accelerator-driven transmutation system has been studied at the Japan Atomic Energy Research Institute. This system is a hydrid system which consists of a high intensity accelerator, a spallation target and a subcritical core region. In the conceptual design study, two types of system concepts, sodium cooled and lead-bismuth cooled system, are being studied. In this study, we fucus on our lead-bismuth cooled accelerator-driven transmutation system to investigate basic characteristics. The fuel compositions were optimized for efficient transmutation of minor actinide. The transmutation of long-lived fission products was also considered.
 
The Impact of Accelerator Based Technologies on Nuclear Fission Safety (IABAT) project was initiated to perform a preliminary assessment of the potential of Accelerator-Driven Systems (ADS) for transmutation of nuclear waste and for nuclear energy production with minimum waste generation. Four ADSs were studied for different fuel/coolant combinations: liquid metal coolant and solid fuel, liquid metal coolant and dispersed fuel, and fast and thermal molten salt systems. This paper presents a summary of the results obtained to date.
 
Japan Atomic Energy Research Institute (JAERI) performs Research and Development (R&D) for accelerator-driven systems (ADS) for transmutation of long-lived nuclides. To study the basic characteristics of ADS, Transmutation Experimental Facility is proposed within the framework of the J-PARC project. The facility consists of two buildings, Transmutation Physics Experimental Facility to research the neutronics and the controllability of ADS and ADS Target Test Facility for material irradiation and partial mockup of beam window. A comprehensive R&D program for future ADS plant is also underway in three technical fields, 1) accelerator, 2) lead-bismuth target/coolant and 3) subcritical core.
 
The present paper focuses on analysis of radiological hazard of spallation products that appears to be an important factor that might affect the choice of beam/target performance in designing the accelerator-driven systems. The analysis is done in terms of toxicity expressed in units of Annual Limit on Intake (ALI). It reveals the significant contribution of alpha emitting rare earths (Sm-146, Gd-148, Gd-150, Dy-154) into overall toxicity of spallation targets. (C) 2002 Published by Elsevier Science Ltd.
 
We review recent evaluations of neutron and proton reaction cross sections up to 150 MeV in the LA150 Library, for use in computer code simulations of accelerator-driven systems. An overview is provided of the nuclear theory together with measured cross section data used in the evaluations. The possible use of bismuth activation foils for high-energy neutron spectrometry is also discussed. We describe recent developments to the MCNPX radiation transport code, which merges MCNP and LAHET in one code and uses the LA150 evaluated data. A number of benchmark comparisons against integral experiments are described, for thick-target neutron production, neutron transmission through macroscopic slabs, and neutron kerma coefficients. The benchmarks help validate the transport code and the evaluated data for use in ADS simulations of neutron production in a spallation target (n/p), radiation shielding, heating, and damage. A brief summary is also given of future data needs for subcritical transmuters and spallation targets, in accelerator transmutation of waste technologies.
 
At the Japan Atomic Energy Research Institute (JAERI), active and comprehensive studies on partitioning and transmutation (P&T) of long-lived nuclear waste from the reprocessing processes of spent fuel has been carried out under the OMEGA program. Studies at JAERI include a design study of dedicated transmutation systems both of an MA burner fast reactor (ABR) and an accelerator-driven subcritical system (ADS), and the development of a high intensity proton accelerator as well as the development of partitioning process, nitride fuel fabrication/dry separation process technologies and nuclear data studies.During the course of studies, JAERI developed the concept of the double-strata fuel cycle, in which a dedicated system is used for transmutation. Comparing the various transmutation systems, such as thermal neutron spectrum or fast neutron spectrum systems, power reactors or dedicated systems, from the viewpoints of reactor physics, nuclear fuel cycle and socio-technical issues, it was concluded that the ADS is the best option for transmutation of minor actinide(MA). JAERI, therefore, decided to concentrate its R&D efforts on the development of ADS and related technologies.One of the goals of R&D is to provide a basis for designing demonstration facilities of ADS, aqueous partitioning process and nitride fuel fabrication and dry separation technologies. As the initial step toward this purpose, the construction of an ADS experimental facility is planned under the High-Intensity Proton Accelerator Project which JAERI and the High Energy Accelerator Research Organization (KEK) are jointly proposing since 1998.The paper discusses the some of the results of P&T studies and the outline of the High-Intensity Proton Accelerator Project under which ADS experimental facility will be constructed.
 
A basic study on the nuclear characteristics in the accelerator driven subcritical reactor (ADSR) was performed through a series of neutronics design calculations and reactor physics experiments. Calculations were executed mainly by the MCNPX code, and experiments were performed at the Kyoto University Critical Assembly (KUCA). Some nuclear features of the research reactor type ADSR were revealed through the present study. The following facts were found: 1) Further studies are necessary concerning the nuclear data in the high energy region and the generated neutrons through the spallation reactions especially by the light nuclei and the lower energy protons. 2) The adjustment of subcriticality by the control rod significantly affects the reactor power of ADSR because of the distortion in the neutron flux distribution caused by the control rod insertion. 3) An accurate calculation is essential to evaluate the neutron multiplication in the ADSR. 4) The neutronics behavior after a pulse injection can be approximately simulated by the calculation.
 
European R&D for ADS design and fuel development is driven in the 6th FP of the EU by the EUROTRANS Programme. In EUROTRANS two ADS design routes are followed, the XT-ADS and the EFIT. The XT-ADS is designed to provide the experimental demonstration of transmutation. The EFIT, the European Facility for Industrial Transmutation, aims at a conceptual design of a full transmuter. A key R&D issue is the choice of an adequate fuel. Various fuel forms have been assessed and CERCER and CERMET fuels, specifically the matrices MgO and Mo, have finally been selected. Within EUROTRANS, the domain ‘AFTRA’ is responsible to more deeply assess the behavior of these dedicated fuels and to provide the fuel database for the EFIT. The EFIT is optimized towards: a good transmutation efficiency, high burnup, low reactivity swing, low power peaking, adequate subcriticality, reasonable beam requirements and a high safety level. In the current paper the fuels under investigation are described, including their production route and their safety limits. First core designs of CERCER and CERMET fuelled 400 MWth EFITs have been developed within AFTRA. The trends found in the design studies and first safety analyses are presented.
 
XY cut (core median plane) of MASURCA reactor in the SC0 1108 fuel cells configuration with the detector locations.
MUSE-4 1108 fuel cells configuration: reactivity and the corrective spatial factors evaluated by means of the application of the area method and the 'explicit' Source-Jerk calculation procedures in each detector position
One of the operative problems in the Accelerator Driven Systems (ADS) is to develop a strategy for inferring the reactivity level. In this frame the zero-power MUSE experimental program, carried out at the CEA-Cadarache MASURCA facility, indicated the Pulsed Neutron Source (PNS) area method to be very reliable at large subcriticalities. All the methods used to measure the reactivity level of a system are based on the point kinetics assumption. Depending on the reactivity level and on the presence of spatial effects, inferring the subcriticality level of the actual systems generally needs at least corrective spatial factors, evaluated by means of calculations. In this paper two static calculation procedures reproducing the application of the Source-Jerk method and PNS area method are proposed in order to evaluate the corrective spatial factors to be applied to the experimental results obtained by means of the corresponding experimental techniques. In this paper a MUSE experimental configuration at a meaningful subcritical level has been analyzed. Results show that the calculation procedures seem to be capable to predict the spatial reactivity spread of the experimental results.
 
A review is given of the technological status of accelerator-driven nuclear systems for transmutation and energy production based on information given at the 2nd International Conference on Accelerator-Driven Transmutation Technologies and Applications held on June 3-7, 1996 at Kalmar, Sweden. The review includes the present status of accelerator, target/blanket and separation techniques. Aspects on safety and proliferation issues of accelerator-driven systems are also given. (C) 1997 Published by Elsevier Science Ltd.
 
The present study is a content analysis of references and abstracts from three international CDROM databases: INIS, INSPEC and Chemical Abstracts (CA). A total of 2336 bibliographic records on Accelerator Driven Systems for Energy Production and Waste Transmutation were downloaded and analyzed. These records were grouped under six separate categories (1) target systems. (2) Blanket/fuel systems. (3) Materials studies. (4) Experiments and computer simulation codes. (5) Chemical separation and fuel processing. (6) Accelerator systems and development.
 
The U.S. Program to evaluate accelerator-driven systems for transmuting problematic, long-lived nuclear waste stream components was initiated during fiscal year 2000, based largely on the Accelerator-driven Transmutation of Waste (ATW) Technology Development Road Map developed during 1999. The Road Map (DOE/RW-0519) effort provided a long-range plan, involving technology development, demonstration, and deployment, as well as a recommended initial effort to evaluate the technology options for five or six years. This paper summarizes the ATW Research and Development Plan currently in draft form. Technology Readiness Levels (TRLs), which are based in part on a system used by the U.S. National Aeronautics and Space Administration in determining levels of flight readiness, was developed for use in assessing and advancing technologies relevant to waste transmutation. Based on TRLs and other considerations, the Program is screening technology options and prioritizing the long-term research and development effort. A top-level schedule illustrates the efforts planned to advance the important technology options in preparation for integrated system tests.
 
A basic study on the nuclear characteristics in the accelerator driven subcritical reactor (ADSR) was performed through a series of neutronics calculations in view of a future neutron source in Kyoto University Research Reactor Institute (KURRI) for the joint use program among researchers of Japanese universities. In this series of calculations, it was assumed that three kinds of monoenergetic neutrons were isotropically generated at the center of spherical and homogeneous cores with different moderator-to-fuel volume ratios in order to examine the spectrum mismatching effect between injected neutrons and fission neutrons born in the subcritical core. The results of calculations clearly showed the spectrum mismatching effect on the neutron multiplication in the ADSR.
 
A postulated steam generator tube rupture (SGTR) accident in a lead cooled accelerator driven transmuter (ADT) is investigated. The design of the ADT without intermediate loops bears the risk of water/steam blasting into the primary coolant. As a consequence a nuclear power excursion could be triggered by steam ingress into the ADT core which has a significant positive void worth. A thermal coolant–coolant interaction (CCI) might initiate a local core voiding too and additionally could lead to sloshing of the lead pool with mechanical impact of the heavy liquid on structures. The steam formation will also lead to a pressurization of the cover gas. The problems related to an SGTR are identified and investigated with the SIMMER-III accident code.
 
The neutronics and burnup analyses of an accelerator-based transmutation system with tungsten target and TRU-nitride fuel were performed with a newly developed code system named ATRAS (Accelerator-based Transmutation Reactor Analysis System). The ATRAS code is an integrated code system which can perform the hadronic cascade process above 20 MeV and neutron transport and core burnup process below 20 MeV with the spallation neutron source.The specifications of the transmutation system are investigated. The core consists of the central spallation target region and the surrounding TRU-mononitride fuel region. The core is driven by protons at an energy of 1.0 GeV. This system was also proposed as a benchmark problem in the “OECD NEA/NSC Benchmark on Physics aspects of Different Transmutation Concepts”.According to the calculation results by the ATRAS code, higher power density and transmutation rate were achieved with nitride fuel, and the neutron spectrum was slightly harder than that of the metallic fuel system. The burnup calculation for thermal power 800 MW was also performed with the ATRAS code. It is shown that about 300 kg of TRU are transmuted annually.
 
The trends of the high power accelerators development are outlined. The natural examples of their applications in nuclear physics and technology are discussed: muon physics, physics of rare decays, intense 14 MeV - neutron source based on muon catalyzed fusion (INS - MCF) and accelerator driven system (ADS) for nuclear waste incineration. The accelerator with power similar to 10 MW and particle energy similar to1 GeV/nucl is considered as the best candidate for these purposes. (C) 2002 Published by Elsevier Science Ltd.
 
Since ALWRs are anticipated to provide an enhanced margin of safety and newer nuclear power plants have a higher standard of severe accident safety performance than the existing plant designs, recent efforts aimed at simplifying the concept of EPZ for ALWR standard plant designs had been made by a group of utility companies through the EPRI in the U.S. In Korea, similar efforts have been made for the APR1400, which is a 1,400 MW-sized PWR of an advanced concept developed in Korea. This is possible due to a large enhancement in the performance of various accident mitigation features adopted in its design. The technical assessment shows that radius of the EPZ area of the APR1400 can be reduced to 700 in. However, even though the reduction in the EPZ area is well accepted through an extensive technical assessment process, it has been found that public acceptance appears to be a completely different matter. Therefore, this study performed a public poll to assess the degree of public acceptance to a reduction in the EPZ area and to identify the means of implementing the simplification of EPZ that would be most acceptable to the public.
 
In this study, we have developed a thermo-hydraulic and safety analysis code named TSAC1.0 with Visual Fortran 6.5 to analyze the thermal-hydraulic characteristics of the China advanced research reactor (CARR) under reactivity insertion accident (RIA) which was induced by unexpected control rod withdrawal in full power condition. The neutron kinetic model depended on the point kinetics with six groups of delayed neutrons including reactivity feedback effects and it was adopted for the solution of reactor power. Furthermore, a new simple and convenient model was adopted for the solution of the transient behaviors of main pump instead of the complicated four-quadrant model. Visual input, real-time processing and dynamic visualization output were achieved using Microsoft Visual Studio.NET 2003 to make the application of TSAC1.0 much more convenient in the engineering. The simulated results of TSAC1.0 were found to be in reasonable agreement with those of RELAP5/MOD3 and showed that the parameters, including the peak coolant temperature, the peak heat structure temperature, and MDNBR, were in the acceptable range of design safety limit under RIA.
 
The objectives of the SARNET network are to define common research programmes in the field of severe accidents and to develop common computer tools and methodologies for safety assessment in this field. To reach these objectives, one of the work packages, named “Severe Accident Research Priorities” (SARP), aimed at reviewing and reassessing the priorities of research issues as a basis to harmonize and to re-orient research programmes, to define new ones, and to close – if possible – resolved issues on a common basis. The work was performed in close collaboration with 8 participating institutions, led by GRS, representing technical safety organisations, industry and utilities (IRSN, CEA, EDF, FZK, GRS, KTH, TUS, VTT). This action made use notably of (1) the outcomes of the EURSAFE project in the 5th Framework Programme, i.e. the Phenomena Identification and Ranking Tables (PIRT) on severe accidents, (2) the results of the validation and benchmarking activities on ASTEC, (3) the results of reactor calculations carried out in the other SARNET tasks, and (4) the outcome of the research performed in the three thematic sub-domains of SARNET (corium, containment and source term).
 
Top-cited authors
Guanghui Su
  • Xi'an Jiaotong University
S.Z. Qiu
  • Xi'an Jiaotong University
XI WEN Tian
  • Xi'an Jiaotong University
Elsa Merle
  • Institut National de Physique Nucléaire et de Physique des Particules
Lelio Luzzi
  • Politecnico di Milano