Challenges and Innovative Technologies On Fuel Handling Systems for Future Sodium-Cooled Fast Reactors

The reactor refuelling system provides the means of transporting, storing, and handling reactor core subassemblies. The system consists of the facilities and equipment needed to accomplish the scheduled refuelling operations. The choice of a FHS impacts directly on the general design of the reactor vessel (primary vessel, storage, and final cooling before going to reprocessing), its construction cost, and its availability factor. Fuel handling design must take into account various items and in particular operating strategies such as core design and management and core configuration. Moreover, the FHS will have to cope with safety assessments: a permanent cooling strategy to prevent fuel clad rupture, plus provisions to handle short-cooled fuel and criteria to ensure safety during handling. In addition, the handling and elimination of residual sodium must be investigated; it implies specific cleaning treatment to prevent chemical risks such as corrosion or excess hydrogen production. The objective of this study is to identify the challenges of a SFR fuel handling system. It will then present the range of technical options incorporating innovative technologies under development to answer the GENERATION IV SFR requirements.     

Uncertainty Analysis for Unprotected Accidents in Sodium-Cooled Fast Reactors

Reactor safety analyses often utilize a deterministic approach where in addition to performing best estimate calculations, uncertainty is accommodated by performing calculations with pessimistic values for input parameters that are important to safety. Here, a stochastic approach is considered for explicitly including uncertainty in safety parameters by applying Monte Carlo sampling coupled with established deterministic reactor safety analysis tools. The Monte Carlo approach yields frequency distributions for reactor safety metrics (e.g., peak temperatures) that can be compared to performance limits, allowing for an improved determination of the safety margin and a clear determination of which safety parameters are most important to the transient response. Because the approach enables the estimation of probabilities for violating safety boundaries, it should be useful in a risk-based regulatory environment. It has the advantage of not requiring any substantial rewriting of existing safety analysis computer codes.     

Development and Testing of Fuel Elements with Vibrationally Compacted Oxide Fuel for Fast Reactors

The main results of a series of scientific-research and technological studies performed at the State Science Center of the Russian Federation – Scientific-Research Institute of Nuclear Reactors to substantiate the use of fuel elements with vibrationally compacted oxide fuel in fast reactors are presented. In the course of this work, the physical-mechanical and technological characteristics of granular UO₂ and UPuO₂ fuel were studied; radiation tests and materials-engineering investigations of experimental and test fuel elements were performed in BOR-60, BN-350, and -600 reactors. More than 30,000 fuel elements were fabricated. Maximum burnup 30% heavy atoms was attained in BOR-60 using fuel assemblies with the standard construction and 32.3% heavy atoms was obtained using experimental fuel elements with a collapsible fuel assembly. In testing fuel elements with vibrationally compacted UPuO₂ in BN-600, maximum burnup of 9.6% (10.8% heavy atoms for individual fuel elements) was achieved. Postreactor investigations showed that the service life of the fuel elements is determined only by the choice of the cladding material. In accordance with the concept developed at the Ministry of Atomic Energy of Russia for the utilization of weapons plutonium, the Institute set about to implement in practice a technology for converting the metallic weapons-grade plutonium into mixed uranium–plutonium oxide fuel on the basis of pyroelectrochemistry and vibrational compaction.     

Mononitride Fuel for Fast Reactors

Substantiation is given for the development of nuclear power based on inherently-safe fast reactors with a mononitride core. Fundamental studies and design work on the development of such reactors with lead (BREST-OD-300), lead–bismuth (SVBR-75/100), and sodium coolant (BN-800) are being performed. The development of nuclear power in our country is based on organizing a closed fuel cycle. The results of experimental investigations of the properties of mononitride fuel are correlated. Mononitride fuel meets all requirements for fast-reactor fuel.     

快堆燃料组件热工流体力学计算研究

对于钠冷快堆,在燃料和包壳最高温度等设计限值下,为获得较高的堆芯出口温度,需深入分析燃料组件内的热工流体力学问题,准确预测组件内的冷却剂温度分布。本文在CRT模型和F.C.Engel等人工作的基础上,提出了ICRT压降关系式,用以计算冷却剂在湍流区、过渡流区和层流区的棒束压降;引入CRT模型和WEST对流传热模型,改进了SUPERENERGY子通道分析程序,并将改进程序与原程序计算结果进行了对比,结果表明：最热子通道出口温度略有降低,液膜温压略有增加;并用计算流体力学软件CFX对中国实验快堆单盒燃料组件活性段进行了三维数值模拟,将计算结果用CRT模型、ICRT压降关系式及改进后的SUPERENERGY子通道分析程序进行了验证,相互符合较好。     

Safety Approach and R&D Program for Future French Sodium-Cooled Fast Reactors

This paper presents briefly the safety approach as well as the R&D program that is underway to support the deployment of future French Sodium-Cooled fast Reactors (SFRs): A) Safety objectives and principles for future reactors. The content of the first section reflects the works of AREVA, CEA, and EDF concerning the safety orientations for the future reactors. The availability of such orientations and requirements for the SFRs has to allow introducing and managing the process that will lead to the detailed definition of the safety approach, to the selection of the corresponding safety options, and to the identification and motivation of the supporting R&D. B) Strategy and roadmap in support of the R&D for future SFRs. This section describes the R&D program led jointly by CEA, EDF, and AREVA, which has been developed with the objectives to be able to preliminarily define, by 2012, the safety orientations for the future SFRs, and to deduce from them the characteristics of the ASTRID prototype.     

Inherent Prevention and Mitigation of Severe Accident Consequences in Sodium-Cooled Fast Reactors

Safety challenges for sodium-cooled fast reactors include maintaining core temperatures within design limits and assuring the geometry and integrity of the reactor core. Due to the high power density in the reactor core, heat removal requirements encourage the use of high-heat-transfer coolants such as liquid sodium. The variation of power across the core requires ducted assemblies to control fuel and coolant temperatures, which are also used to constrain core geometry. In a fast reactor, the fuel is not in the most neutronically reactive configuration during normal operation. Accidents leading to fuel melting, fuel pin failure, and fuel relocation can result in positive reactivity, increasing power, and possibly resulting in severe accident consequences including recriticalities that could threaten reactor and containment integrity. Inherent safety concepts, including favorable reactivity feedback, natural circulation cooling, and design choices resulting in favorable dispersive characteristics for failed fuel, can be used to increase the level of safety to the point where it is highly unlikely, or perhaps even not credible, for such severe accident consequences to occur.     

Analysis of Mixed Oxide Fuel Critical Experiments with Neutronics ANalysis Codes for Boiling Water Reactors

Critical experiments of UO₂ and full mixed oxide (MOX) fuel cores conducted at the Tank-type Critical Assembly (TCA) were aNalyzed using BWR design-purpose codes HINES and CERES with ENDF/B files and Monte Carlo fine analysis codes VMONT and MVP with the JENDL-3.2 library.

The averaged values of the multiplication factors calculated with HINES/CERES, VMONT and MVP agreed with those of experiments within 0.3%ΔAk. The values by the design-purpose codes showed a small difference of 0.1%Δk between UO₂ and MOX cores. Monte Carlo code results showed that the JENDL-3.2 library had a tendency to overestimate the multiplication factors of UO₂ cores by about 0.3%Δk compared with those values of MOX cores. The root mean square errors of calculated power distributions were less than 1% for HINES/CERES and VMONT.

These results showed that (1) the accuracy of these codes when applied to full MOX cores was almost the same as their accuracy for UO₂ cores, which confirmed the accuracy of present core design codes for full MOX cores; and (2) the accuracy of the 190-energy-group Monte Carlo calculation code VMONT was almost the same as that of the continuous-energy Monte Carlo calculation code MVP.     

Fast Reactors. Status and Prospects

The stages of the development of fast reactors in the world are analyzed. It is shown that substantial progress has been made in the development and operation of sodium-cooled fast reactors and accident-free operation of the main liquid-metal equipment, equal to the performance of general industrial equipment. Ways to make nuclear-power-plant units of this type competitive are discussed.The status of the work on fast reactors with other coolants – gas, steam, and heavy metals – is briefly reviewed. The main problems which must be solved to implement these directions of the development of fast reactors are indicated.     

BN-800 as a New Stage in the Development of Fast Sodium-Cooled Reactors

The role of fast reactors in a strategy for developing nuclear power in Russia because of the inevitable exhaustion of natural uranium deposits in the foreseeable future is discussed. The BN-800 reactor, which is under construction and incorporates unique solutions – greatly enhancing the safety of the reactor – to technical and constructional problems, is examined. Cost assessments taking account of the complete life cycle show that fast reactors could be no more expensive than the most widely reactors in the world – water-moderated water-cooled reactors.Closing the BN-800 nuclear fuel cycle will make it possible to solve the problem of utilizing plutonium and actinides. This makes fast reactors safer for the environment.     

Prospects for the Development of Fast Breeder Reactors

Possible ways to improve nuclear power systems with fast breeder reactors and conditions for ensuring that such systems are competitive are discussed. Certain questions concerning schematic and structural improvements are examined. The results of a comparative analysis of sodium- and lead-cooled breeder reactors are presented. It is pointed out that for sodium-cooled reactors the corresponding informtion is due to many years of experience in developing, investigating, and operating experimental, test, and commercial reactors. There is no experience in developing lead-cooled reactors. A comparative analysis does not confirm that there are any advantages with respect to technical or economic performance for lead-cooled breeder reactors.     

池式钠冷快堆热分层现象模型开发及瞬态分析

针对池式钠冷快堆特点,建立了三维系统分析模型,并结合热分层现象演化机制,提出了准确模拟热分层的关键处理方法,包括能量源项处理、三维动量方程对流项处理及三维空间进口效应处理。在此基础上,采用KALIMER及MONJU热分层实验对所开发的三维系统分析模型进行验证。结果表明模型有效解决了池式钠冷快堆三维热工水力分析的难题,实现了对钠池内温度场瞬态变化及热分层现象演化进程的快速准确模拟,同时也能够确定热分层过程中池式结构表面热应力最大位置,为池式快堆安全设计提供参考。      

Prospects for the BN-1800 Sodium-Cooled Fast Reactor Satisfying 21st Century Nuclear Power Requirements

The BN-1800 power-generating unit is designed to meet the requirements of the strategy for developing atomic energy in Russia in the first half of the 21st century. The development time is the next 15 years and construction could start after 2020. The design is innovative and includes the development of key new technical solutions as compared with the BN-800 reactor which is now under construction.The new technical solutions are based on the substantial positive experience in operating fast reactors in Russia (~125 reactor·years), specifically the BN-600 reactor. The innovations make it possible not only to solve strategic problems, such as increasing safety, improving ecology (including by burning actinides), and nonproliferation but also to make large improvements in economic performance.     

Prospects for Using High-Nickel Alloys in Power Reactors with Supercritical-Pressure Water

Certain high-nickel alloys possess attractive properties: they swell very little under neutron irradiation and possess comparatively high heat and corrosion resistance. An assessment is made of the possibility of using such alloys as structural materials for fast-reactor cores with supercritical pressure water from the standpoint of radiation resistance and expected high corrosion resitance of the alloys in water with high parameters. The information on the influence of neutron irradiation on the swelling, creep, and mechanical properties of domestic and foreign high-nickel alloys is analyzed.     

Thermohydraulic Aspects of Deep Burnup of Nuclear Fuel in Fast Reactors

The results of thermomechanical and thermohydraulic studies showing the relative effect of the deformation of fuel-element claddings and lattices in fast-reactor fuel assemblies on their temperature regimes are presented. It is shown that the temperature nonuniformities in fuel assemblies largely determine the deformation of fuel assemblies and, in turn, the operating efficiency and, correspondingly, the degree of burnup of nuclear fuel in fast reactors. The increase in the efficiency of the fuel assemblies is largely due to temperature smoothing, including smoothing of local temperature nonuniformities. Various solutions to technical and structural problems can accomplish this.