钠冷快堆燃料破损定位方法综述

反应堆燃料元件破损后,为了确保运行安全和及时处理破损事故,必须对破损的燃料组件进行定位操作。本文广泛调研了国际上钠冷快堆中采用的各种燃料破损定位方法,介绍了定位原理,评析各种方法的应用范围和缺陷,展望了燃料破损定位技术的研发趋势,为我国钠冷快堆燃料破损定位方法的选择提供了参考依据。     

钠冷快堆燃料元件性能分析程序的开发与验证

为了评估钠冷快堆氧化物燃料元件稳态、瞬态和事故条件下的性能和行为演化,开发了钠冷快堆燃料元件性能分析程序FIBER。程序采用有限体积法实现燃料元件温度的计算,用有限元方法实现力学、裂变气体释放的计算,并通过时间步长控制模块控制程序的稳定运行。为验证程序的准确性,通过调研得到俄罗斯BN600反应堆辐照数据,与FIBER程序的裂变气体释放、柱状晶粒等计算结果进行对比分析。结果表明,FIBER程序对最大燃耗11.8at%、最大辐照损伤78 dpa的快堆燃料元件的辐照变形、柱状晶区、裂变气体释放性能评价是有效的。     

钠冷快堆的金属燃料元件

使用铀-钚-锆金属合金燃料的钠冷快堆具有良好的固有安全性。采用小堆组合的模块化设计使这类金属燃料快堆电站具有很好的固有安全性、经济性、增殖性并可实现燃料的现场后处理。金属燃料的加工及后处理都采用高温冶金方法,因而制造方便,造成的放射性废物量少。金属型快堆燃料已重新受到世界上的重视。     

基于超临界二氧化碳布雷顿循环的钠冷快堆方案研究

超临界二氧化碳布雷顿循环因热效率高、布置紧凑等特点受到了广泛的关注,多种核能系统将其列为备选的动力循环系统。为研究基于超临界二氧化碳布雷顿循环的钠冷快堆系统的特点,本文在调研和分析的基础上,从反应堆回路数目、动力循环方式、系统参数选取及设备材料选型等方面开展了分析与对比,针对给定的系统配置方式初步分析了系统主要参数特点,并对应用于钠冷快堆的超临界二氧化碳动力循环系统的发展提出了建议。     

钠冷快堆燃料组件热工水力特性数值模拟与分析

利用CFD程序CFX,分别对7、19、37、61根棒组成的三角形排列螺旋绕丝定位的钠冷快堆燃料组件棒束通道进行了热工水力特性的分析研究,并将结果与子通道程序SuperEnergy进行了对比验证。重点考察了棒束通道轴向流动分布、横向流交混效应及子通道轴向温升,分析了定位绕丝的影响。结果表明,绕丝对棒束通道的横向流交混效应、轴向流动分布及子通道温升有着重要影响,且随棒束的增多,通道内的流动趋向复杂化,轴向流动不均匀性有升高趋势。     

钠冷快堆棒状燃料堆芯子通道分析程序开发及验证

热工水力分析软件的验证是安全审查重点关注的问题。为了实现不同设计软件间的对比验证,本工作开发出具有自主知识产权的钠冷快堆堆芯子通道分析程序SSCFR,进行中国实验快堆（CEFR）全堆芯稳态分析、子通道稳态分析及全堆芯瞬态分析,并将分析结果与CEFR运行和设计值进行对比。结果表明,SSCFR程序的计算结果与CEFR运行值及安全分析报告中的设计计算值符合较好,可用于钠冷快堆后续的软件对比验证及设计计算工作。     

钠冷快堆自然循环组件子通道程序初步研发

本文为计算和分析钠冷快堆自然循环组件的热工水力特性,开发了钠冷快堆堆芯自然循环冷却组件子通道分析程序。基于61棒单组件模型,通过将本程序的结果与COBRA程序进行比较,验证了钠冷快堆堆芯自然循环冷却组件子通道分析程序对自然循环冷却组件的适用性。基于多盒组件模型,初步验证了本程序具备自然循环冷却组件的流量分配和盒间换热计算的功能。本程序能为池式快堆自然循环冷却组件提供有效的设计和分析工具。     

超临界二氧化碳动力循环在钠冷快堆中的应用综述

超临界二氧化碳循环系统在气冷快堆、铅冷快堆、钠冷快堆中极具应用前景。综述了应用于钠冷快堆的超临界二氧化碳动力循环系统及其样机关键部件研究现状和有关进展,结合钠冷快堆的热源特征,分别就典型超临界二氧化碳动力循环结构、印刷电路板式换热器换热特征系数、不同功率等级S-CO_2压缩机与透平类型选择以及轴承与密封关键特征进行了总结与分析,分析结果为后续开展适用于钠冷快堆的S-CO_2布雷顿循环设计及样机开发提供可借鉴的参考与依据。     

钠冷快堆中池式钠火的计算分析

文章论述了根据池式钠火的特点建立了理论模型 ,编制了SPOOL程序。该程序模拟钠燃烧过程中钠和氧气的化学反应 ,钠燃烧热在各种介质中不同方式的传递 ,钠气溶胶的产生、沉积 ,以及在各种通风条件下多种介质的质量和能量交换等瞬态过程 ,描述了钠燃烧过程中各种特征参数随时间的变化。其主要的计算参数包括房间内气体的压力和温度、房间建筑结构的温度、钠气溶胶质量浓度等等。用俄罗斯别洛雅尔斯克核电站实验和法国卡桑德拉 3号实验的数据 ,对SPOOL程序进行验证的结果表明 ,该程序的计算结果可信。该程序为国内钠冷快堆中池式钠火事故的安全分析提供了分析方法     

钠冷快堆中喷雾钠火的计算分析

根据钠冷快堆中喷雾钠火的特点建立了理论模型,编制了SSPRAY程序。该程序模拟钠喷雾燃料过程中钠滴的运动、钠和氧气的燃烧反应、热量传递和质量传递等瞬态过程。用该程序计算了气体和墙壁温度、气体压力、氧气摩尔份额、喷雾流燃烧速率和热量热递速率等主要参数。利用AI实验数据和美国SPRAY－3A的计算结果对程序进行了验证,结果符合较好。     

Co-processing of uranium and plutonium for sodium-cooled fast reactor fuel reprocessing by acid split method for plutonium partitioning without reductant

A solvent extraction flowsheet for Pu partitioning, based on the acid split method without reductant, originally proposed by the Oak Ridge National Laboratory (ORNL), was tested for sodium-cooled fast reactor fuel reprocessing. To enhance resistance to nuclear proliferation, a flowsheet for co-processing was developed that controls Pu content in the products while avoiding Pu polymerization and formation of a third phase during extraction. In this method, Pu is partitioned using the difference in distribution coefficients of U and Pu. It is effective for selective Pu stripping from U at low temperatures and HNO₃ concentrations. The flowsheet with a supply of 0.15 mol/dm³ HNO₃ solution at 21°C for Pu partitioning was tested experimentally using miniature centrifugal contactors and a highly radioactive solution. Neither a Pu(IV) polymer nor a third phase was observed during the experiment. The Pu content in the U/Pu product increased to 2.28 times that in the feed solution. The leakage ratio of Pu to the U product was slightly less in the U stripping section. Some fission products (FPs) were effectively decontaminated; e.g., decontamination factors (DFs) of Cs in U/Pu and U products were 4.51×10⁵ and 2.42×10⁵, respectively.     

Innovative nuclear technology based on modular multi-purpose lead–bismuth cooled fast reactors

Today's nuclear power is in the state of an intrinsic conflict between economic and safety requirements. This fact makes difficult its sustainable development.

One of the ways of finding the solution to the problem can be the use of modular fast reactors SVBR-75/100 cooled by lead–bismuth coolant that has been mastered in conditions of operating reactors of Russian nuclear submarines.

The inherent self-protection and passive safety properties are peculiar to that reactor due to physical features of small power fast reactors (100 MWe), chemical inertness and high boiling point of lead–bismuth coolant, integral design of the pool type primary circuit equipment.

Due to small power of the reactor, it is possible to fabricate the whole reactor at the factory and to deliver it to the NPP site in practical readiness by using any kind of transport including the railway.

Substantiation of the high level of reactor safety, adaptability of the SVBR-75/100 reactor relative to the fuel type and fuel cycle, issues of non-proliferation of nuclear fissile materials, opportunities of multi-purpose usage of the standard SVBR-75/100 reactors have been viewed in the paper.     

Computational fluid dynamic investigations of partial blockage detection by core-temperature monitoring system of a sodium cooled fast reactor

Thermal hydraulic studies have been carried out to understand temperature dilution suffered by core-temperature monitoring system of a sodium cooled fast reactor. The three-dimensional computational model is validated against experimental results of a water model. Jet mixing phenomenon as predicted by different turbulence models is compared and RNG k–? model is found to be better than other models. A comprehensive parametric study considering: (i) effects of construction/manufacturing tolerances on thermocouple positions with respect to subassembly positions, (ii) thermal/irradiation bowing of subassemblies, and (iii) changes in core power profile during reactor operation cycles has been carried out. The studies indicate the maximum possible dilution in fuel and blanket subassemblies to be 2.63 K and 46.84 K, respectively. Shifting of thermocouple positions radially outward by 20 mm with respect to subassembly centers leads to an overall improvement in accuracy of thermocouple readings. It is also seen that subassembly blockage that leads to 7% flow reduction in fuel subassembly and 12% flow reduction in blanket subassembly can be detected effectively by the core-temperature monitoring system.     

Plutonium and Minor Actinides incineration options using innovative Na-cooled fast reactors: Impacting on phasing-out and on-going fuel cycles

The flexibility of innovative Na-cooled fast reactors for burning Pu and/or Minor Actinides (MA) is investigated with respect to different fuel cycle strategies. Under phasing-out conditions, the burner systems are used for reducing to a minimum level the accumulated TRansUranic (TRU) inventory, whereas when continuous use of nuclear energy is envisaged (on-going case), burner systems may be dedicated to MA management only.As an example of a phasing-out case, the accumulated German TRU inventory (at 2022) is assumed to be transmuted in a chosen time period of 150 years. For this purpose, two different burner fast reactors concepts, developed at KIT, are deployed in a Partitioning and Transmutation based fuel cycle. The effects are analyzed in order to confirm the behavior expected by the neutronics studies and to provide a basis for further optimization of the scenarios with respect to a number of reactors, deployment paces and fuel compositions.Additionally the performance of the MA burner is assessed to provide an effective MA mass stabilization in case of a continuous use of nuclear energy. Preliminary results are compared with those of past studies based on the European Sodium-cooled Fast Reactor.     

MOX燃料在轻水堆核电站中的应用

目前MOX燃料已成为一种可用于轻水堆核电站成熟的核燃料。简要介绍了国外该领域的发展状况以及MOX燃料对反应堆性能的主要影响和应对措施。探讨了MOX燃料在国内压水堆核电站中的应用问题。     

Development of the residual sodium quantification method for a fuel assembly of SFRs

ABSTRACT

In a fuel handling system of sodium-cooled fast reactors (SFRs), it is necessary to remove the sodium remaining on spent fuel assemblies (FAs) before storing them in a spent fuel water pool (SFP). A next-generation SFR in Japan has adopted an advanced dry cleaning process which consists of argon gas blowing to remove the metallic residual sodium on the FA, which increases economic competitiveness and reduces waste products thanks to a waterless process. In this R&D work, performance of the dry cleaning process has been investigated. This paper describes experimental and analytical work focusing on the amount of residual sodium remaining on FA components, for instance the handling head, the wrapper tube, the upper shielding, and the entrance nozzle. The tests, using water and sodium, investigated the amount of residual liquid remaining on laboratory scale specimens representing three fundamental shapes: narrow gaps, horizontal holes, and corners. On the basis of the experimental results, the residual sodium quantification method for FA was constructed. The constructed method enables quantitative estimation of the amount of residual sodium on the entire FA before and after the argon gas blowing with 95% reliability.     

Visual study of ex-pin phenomena for SFR with metal fuel under initial phase of severe accidents by using simulants

In the present Korean sodium-cooled fast reactor (SFR) program, early dispersion of the molten metal fuel within a subchannel is suggested as an inherent safety strategy in the initiating phase of a hypothetical core disruptive accident (HCDA). This safety strategy provides a negative reactivity driven by the melt dispersion; therefore, it could reduce the possibility of occurrence of a severe recriticality event. In the initiating phase, the melt could be injected into the subchannel horizontally by the internal pressure of the fuel pin. Complex phenomena occur during intermixing of the melt with the coolant after the horizontal injection of the melt. It is rather difficult to understand the several combined mechanisms that occur that are related to the dispersion and fragmentation of the melt. Thus, it seems worthwhile to study the horizontal injection of melt at lower temperatures, which could help to observe the dispersion phenomenon and understand the fragmentation mechanism. In this work, for a parametric study, tests were performed under structural conditions, coolant void conditions, and boiling conditions. As a result, in some cases, the injected molten materials were stuck around the injection hole. On the other hand, the molten materials were dispersed upward sufficiently well under the boiling condition when R123 was used as the coolant. The built-up vapor pressure was found to be one of the driving forces for the upward dispersion of the molten materials.     

聚变裂变混合乏燃料焚烧堆FDS-SFB燃料循环方案设计与分析

针对聚变裂变混合乏燃料焚烧堆FDS-SFB(Spent Fuel Burner),基于湿法和干法两种后处理技术途径提出了不同的燃料循环方案。并分别对FDS-SFB燃料循环所需的初装资源量、燃料制备和乏燃料后处理能力进行初步质量流分析和可行性初步评估。基于较好嬗变和增殖性能的FDS-SFB典型中子学方案的质量流初步分析表明:两种方案燃料循环其所需的初装资源量、燃料制备、乏燃料后处理能力具有初步的可行性。     

Neutronic and safety aspects of inert matrix fuel utilization in fast reactors for plutonium and minor actinides transmutation

The solving of ecological problems of future nuclear power is connected with the solving of long-lived radioactive waste utilization problems. It concerns primarily plutonium and minor actinides (MAs), accumulated in the spent fuel of nuclear reactors. One of the ways this can be solved is to use a fast reactor with uranium-free or inert matrix fuel (IMF). The physics of this type of reactor was widely investigated during last year for BN-800 reactors. The solution of the most important problems was: a decrease in non-uniformity of power distribution and an increase of the Doppler effect. The next stage of such core investigations is an evaluation of self-protection to beyond design accidents. Preliminary results show a high safety level of BN-800 reactors with IMF in the event of unprotected loss of coolant flow (ULOF) accident.     

Neutronics studies on the feasibility of developing fast breeder reactor with flexible breeding ratio

This paper investigates the feasibility of designing a flexible fast breeder reactor from the view of neutronics. It requires that the variable breeding ratio can be achieved in operating a fast reactor without significant changes of the core, including the minimum change of fuel assembly design, the minimum change of the core configuration and the same control system arrangement in the core. The sodium cooled fast reactor is investigated. Two difficulties are overcome: (1) the different excess reactivity is well controlled for different cores, especially for the one with small breeding ratio; (2) the maximum linear power density is well controlled while the breeding ratio changes. The optimizations are done to meet the requirements. The U–Pu–Zr alloy is applied to enhance the breeding. The enrichment-zoning technique with unfixed blanket assembly loading position is searched to get acceptable power distributions when the breeding ratio changes. And the control system is designed redundantly to fulfill the control needs. Then, the achieved breeding ratio can be adjusted from 1.1 to 1.4. The reactivity coefficients, temperature distributions and preliminary safety performances are evaluated to investigate the feasibility of the new concept. All the results show that it is feasible to develop the fast reactor with flexible breeding ratios, although it still highly relies on the advancement of the coolant flow control technology.