基于FR-Sdaso的法国凤凰堆寿期末自然循环实验分析

系统分析程序是开展反应堆安全分析的重要工具之一,也可用于开展系统瞬态实验过程的分析。法国凤凰堆（Phenix）在停运之前开展的自然循环实验是钠冷快堆领域非常重要的系统瞬态实验,为研究钠冷快堆的瞬态特点提供了很好的参考。为分析此实验过程,利用自主研发的系统分析程序FR-Sdaso对凤凰堆进行建模,对其自然循环实验开展计算分析,并将主要参数的计算值与实验值进行了对比分析。结果表明,FR-Sdaso可较好地模拟此实验的瞬态过程,可用于开展钠冷快堆此类瞬态的安全分析。     

池式钠冷快堆放射性释放风险概率安全评价事件树分析

池式钠冷快堆的安全特性和放射性释放机制与压水堆有着显著不同,在核安全新要求下,亟待开展放射性释放风险概率安全评价（PSA）研究。本文以池式钠冷快堆为研究对象,通过分析放射性来源、包容边界及破坏包容边界完整性的严重事故现象,确定了池式钠冷快堆大量放射性释放的主要位置和释放模式,构建分析了放射性释放事件树。本文分析结果可为进一步开展池式钠冷快堆放射性释放风险PSA提供参考。     

ATHLET程序的钠冷快堆应用扩展及其验证

系统分析程序是对钠冷快堆的冷却剂回路系统进行全局模拟、瞬态及事故安全分析的重要工具。本工作对德国核设施与反应堆安全机构（GRS）开发的轻水堆最佳估算系统程序ATHLET进行修改,增加了钠的物性公式和传热关系式,将其适用范围扩展到钠冷快堆。为验证修改过的ATHLET程序,对法国凤凰（Phenix）反应堆系统建模,并对其自然对流实验进行模拟,将计算结果与实验数据进行比较。结果显示,ATHLET程序的钠冷快堆应用扩展具有良好的适用性。     

Plutonium management and Generation IV systems

This paper discusses the potential role of Generation IV nuclear energy systems in managing plutonium. It briefly reviews the Generation IV goals and their relevance to plutonium management. Each of the six selected Generation IV systems [very high temperature reactor (VHTR), gas-cooled fast reactor (GFR), sodium-cooled fast reactor (SFR), super-critical-water-cooled reactor (SCWR), lead-cooled fast reactor (LFR), molten salt reactor (MSR)] is briefly discussed. The main characteristics of each system are summarised and the capability for plutonium management indicated. The potential for the management of plutonium using Generation IV systems is briefly reviewed from a complete fuel cycle perspective to illustrate the issues in the context of a fleet of reactor and fuel cycle facilities.     

钠冷快堆系统程序FR-Sdaso开发

为解决600 MW示范快堆（CFR600）事故分析和工况设计中的实际问题,自主开发了钠冷快堆系统程序FR-Sdaso,其建模范围包括堆芯、一回路、二回路、三回路、四回路和事故余热排出系统,主要物理模型包括点堆模型、单通道堆芯热工模型、多区钠池模型、四区蒸汽发生器模型等核岛设备或部件分析模型,汽轮机、凝汽器、给水加热器、除氧器等常规岛设备采用集总参数模型,泵、阀门、管道及控制体等采用通用模型。对程序进行了初步验证,结果表明,FR-Sdaso程序可用于分析全厂瞬态工况及超功率、失流、失热阱等典型事故过程。目前,FR-Sdaso程序已用于CFR600的设计和安全分析。     

Development of System Code FR-Sdaso for Sodium cooled Fast Reactor

To solve actual problems in the accident analysis and working condition design of the 600 MW demenstration fast reactor (CFR600), the sodium-cooled fast reactor (SFR) system code FR-Sdaso was developed, which could be used to model the reactor core, primary system, secondary system, tertiary system, quadruple system and the decay heat removal system of the SFR. The physical models can be divided into three categories: The models for nuclear island equipment including point reactor model, single-channel core thermal model, multi-zone sodium pool model and four-zone steam generator model, etc., the lump parameter models for conventional island equipment, including turbine, condenser, feed water heater, deaerator, etc., and the general models for pump, valve, pipe and control volume. Preliminary V&V work for FR-Sdaso was conducted, and the results show that FR-Sdaso can be used to analyze the transient conditions of the whole plant and typical SFR accidents such as overpower, loss of flow, and loss of heat sink. FR-Sdaso was used in the design and safety analysis of the CFR600.     

Characterization of a sodium-cooled fast reactor in an MHR–SFR synergy for TRU transmutation

In the task of destroying the light water reactor (LWR) transuranics (TRUs), we consider the concept of a synergistic combination of a deep-burn (DB) gas-cooled reactor followed by a sodium-cooled fast reactor (SFR), as an alternative way to the direct feeding of the LWR TRUs to the SFR. In the synergy concept, TRUs from LWR are first deeply incinerated in a graphite-moderated DB-MHR (modular helium reactor) and then the spent fuels of DB-MHR are recycled into the closed-cycle SFR. The DB-MHR core is 100% TRU-loaded and a deep-burning (50–65%) is achieved in a safe manner (as discussed in our previous work). In this analysis, the SFR fuel cycle is closed with a pyro-processing technology to minimize the waste stream to a final repository. Neutronic characteristics of the SFR core in the MHR–SFR synergy have been evaluated from the core physics point of view. Also, we have compared core characteristics of the synergy SFR with those of a stand-alone SFR transuranic burner. For a consistent comparison, the two SFRs are designed to have the same TRU consumption rate of ∼250 kg/GW EFPY that corresponds to the TRU discharge rate from three 600 MW DB-MHRs. The results of our work show that the synergy SFR, fed with TRUs from DB-MHR, has a much smaller burnup reactivity swing, a slightly greater delayed neutron fraction (both positive features) but also a higher sodium void worth and a less negative Doppler coefficients than the conventional SFR, fed with TRUs directly from the LWRs. In addition, several design measures have been considered to reduce the sodium void worth in the synergy SFR core.     

钠冷快堆二回路系统热工水力瞬态分析程序SELTAC的开发与应用

针对钠冷快堆二回路系统的具体结构和运行特点,对中间热交换器、直流蒸汽发生器、钠缓冲罐以及泵、管道等设备和部件建立模型,采用FORTRAN语言自主编制了二回路系统热工水力瞬态分析程序SELTAC。利用中国实验快堆的停堆试验数据对所编制程序进行了初步验证。结果表明,程序计算值与试验值趋势一致,最大相对偏差不超过4.34%,吻合程度较好。将验证后的程序与一回路系统程序耦合,分析了某600 MW钠冷快堆在主热传输系统保持排热能力时的紧急停堆工况,得到了二回路系统的瞬态特性,为大型商用快堆电站的设计提供了参考。     

基于RELAP5 MOD3.2的钠冷快堆热工水力系统分析程序开发及验证

对大型核反应堆热工水力分析程序RELAP5 MOD3.2进行了改造,使之适用于钠冷快堆系统安全分析。在不影响原程序功能的基础上添加了气液两相钠物性和液态金属对流换热模型,并改造了相应的初始化模块和计算模块。改造后的程序可正确模拟钠的流体力学特性和热物性,搭建钠冷快堆热工水力流体网络进行分析计算。对EBR-Ⅱ试验堆基准题进行了稳态模拟和失流事故分析,其中稳态计算主要参数与实验值相对偏差小于1%,瞬态计算相对偏差小于10%,各参数变化趋势与实验值相符良好,初步验证了改造程序的可靠性。     

TRU burning by dual tier system of LWR-SFR

This paper presents a concept of the dual tier system consisting of the existing light water reactor (LWR) plants and sodium-cooled fast reactor (SFR) for transuranics (TRU) burning for the purpose of downsizing the required SFR. In this system, Pu is combusted by the LWR at first and then the remaining Np, Am, and Pu are destructed by the SFR. The iteration number of Pu combustion by the LWR is chosen to be twice owing to the sodium void reactivity limitation of $6. As a result of combustion calculation, the twice Pu burning of LWR lessens the TRU amount by 27% and changes the composition significantly. Moderator pins of zirconium hydride are deployed to the SFR fuel subassembly so as to enhance TRU burning and reduce the sodium void reactivity. The nuclear calculation found that the core characteristics become similar to the conventional SFR due to the moderator: the sodium void reactivity remains still $4 and the Doppler coefficient becomes −6 × 10⁻³ Tdk/dT. This study concludes that this dual tier strategy can downsize the required SFR to approximately 40% of the single tier system of SFR with TRU conversion ratio of 0.6.     

钠冷快堆丧失蒸汽发生器模块后的最佳工况研究

钠冷快堆通过采用模块式蒸汽发生器的设计方案以提高核电厂的负荷因子。核电厂运行中若发生丧失蒸汽发生器模块事件,核电厂工况将发生变化,应进行适当的调节,调节的目标工况可通过设计与研究给出。本工作对某典型池式钠冷快堆丧失1个蒸汽发生器模块后的最佳工况进行研究,主要研究内容包括对其主热传输系统进行建模,开展主热参数匹配计算,根据相关运行限值来筛选方案并分析关键参数,最终给出较为合适的运行工况。本工作为钠冷快堆在丧失蒸汽发生器模块后的工况设计提供了重要依据。     

Development of an evaluation methodology for the natural circulation decay heat removal system in a sodium cooled fast reactor

A natural circulation evaluation methodology has been developed to ensure the safety of a sodium-cooled fast reactor (SFR) of 1500 MW adopting the natural circulation decay heat removal system (NC-DHRS). The methodology consists of a one-dimensional safety analysis which can evaluate the core hot spot temperature taking into account the temperature flattening effect in the core, a three-dimensional fluid flow analysis which can evaluate the thermal-hydraulics for local convections and thermal stratifications in the primary system and DHRS, and a statistical safety evaluation method for the hot spot temperature in the core. The safety analysis method and the three-dimensional analysis method have been validated using results of a 1/10 scaled water test simulating the primary system of the SFR and a sodium test simulating a part of the primary system and the DHRS with about a 1/7 scale, and the applicability of the safety analysis for the SFR has been confirmed by comparing with the three-dimensional analysis adopting the turbulence model. Finally, a statistical safety evaluation has been performed for the SFR using the safety analysis method.     

大型MOX燃料快堆钠空泡反应性微扰理论研究

钠空泡反应性效应是钠冷快堆核设计和安全分析的重要内容。本文基于多群节块扩散法,采用微扰理论推导出钠空泡反应性的计算方法,对1 000 MWe钠冷快堆MOX燃料堆芯的总钠空泡反应性、空间分布、物理分项进行了计算。结果表明,钠空泡反应性主要来源于中子泄漏的增加和能谱的硬化,两者一正一负,且空间分布规律相反,导致钠空泡反应性具有强烈的空间依赖性;对于所计算的MOX燃料堆芯钠空泡反应性高达3 $左右。计算和分析结果阐明了钠空泡反应性的产生机理和分布规律,可为低钠空泡的设计提供参考。     

A comparative study of the MATRA-LMR-FB calculation with the SABRE result for the flow blockage accident in the sodium cooled fast reactor

A sub-channel flow blockage may be initiated by an ingression of damaged fuel debris or foreign obstacles into a core subassembly for the sodium cooled fast reactor (SFR) due to the compact design of the fuel arrangement. Since local coolant temperature could go up high enough to reach a safety limit by the blockage disturbance in the subassembly, the MATRA-LMR-FB code was developed to analyze such blockage effect. An effort has been undergoing to enhance its reliability.In this study, a code-to-code comparison analysis with another code, SABRE4, was performed to supplement a qualification of the MATRA-LMR-FB. The two codes were applied to the analysis of partial sub-channel blockage accidents in a subassembly of the KALIMER-150, which is a conceptual design of a sodium-cooled fast reactor with an electric output of 150 MW. The analyses were carried out not only for radially different blockage positions but also for different blockage sizes in the subassembly.In result, the two code results were generally agreed both in magnitude and trend within a range. Therefore, it was concluded that the comparison results could support complementarily the applicability of the MATRA-LMR-FB to the partial flow blockage accident in the subassembly of the SFR.     

Fundamental safety strategy against severe accidents on prototype sodium-cooled fast reactor

The accident categories of severe accidents (SAs) for prototype sodium-cooled fast reactor (SFR) which need proper measures were investigated through the internal event probabilistic risk assessment (PRA) and event tree analysis for the external event and six accident categories, unprotected loss of flow (ULOF), unprotected transient over power (UTOP), unprotected loss of heat sink (ULOHS), loss of reactor sodium level (LORL), protected loss of heat sink (PLOHS) and station blackout (SBO), were identified. Fundamental safety strategy against these accidents is studied and clearly stated considering the characteristics and existing accident measures of prototype SFR, and concrete measures based on this safety strategy are investigated and organized. The sufficiency of these SA measures is confirmed by comparing the evaluated core damage frequency (CDF) and containment failure frequency (CFF) to the target value, 1×10^?5 and 1×10^?6 per plant operating year, respectively, which were selected based on the IAEA's safety target. However, the target value of CDF and CFF should be satisfied considering all the SAs caused by both internal and external events. External event PRA for prototype SFR is now under evaluation and we set out to satisfy the target value of CDF and CFF considering both internal and external events.     

SARAX程序系统在钠冷快堆瞬态分析中的应用

无保护事故下的瞬态分析是钠冷快堆安全分析的重要内容。基于OECD/NEA发布的MOX-3600和MET-1000基准题,本文利用SARAX程序系统对不同钠冷快堆进行了瞬态计算,分析了堆内各种反应性反馈效应,并计算了无保护失流(ULOF)事故和无保护超功率运行(UTOP)事故下燃料温度和冷却剂温度的变化。计算结果表明：SARAX程序系统在快堆瞬态分析中可给出合理的参数预测结果;ULOF事故对于钠冷快堆是更为严重的事故瞬态,会导致堆内的钠沸腾进而发生严重事故。     

钠冷快堆瞬态热工水力及安全分析程序开发

钠冷快堆是第4代核反应堆的主力堆型,瞬态热工水力及安全特性是其设计研发和安全评审的重要工作,需要专用的分析工具。本文基于模块化建模思想,建立了钠冷快堆系统关键部件的热工水力模型和辅助模型,采用具有高稳定性和自动变步长能力的Gear算法,开发了钠冷快堆瞬态热工水力及安全分析软件THACS,并通过了国际基准题EBR-Ⅱ的有保护失流事故实验SHRT-17的初步验证。结果表明,THACS程序能较好模拟此实验的瞬态过程,具备钠冷快堆瞬态热工水力及安全分析的能力,可为我国钠冷快堆研发提供支持。     

CFD modeling and thermal-hydraulic analysis for the passive decay heat removal of a sodium-cooled fast reactor

In this study, a pool-typed design similar to sodium-cooled fast reactor (SFR) of the fourth generation reactors has been modeled using CFD simulations to investigate the characteristics of a passive mechanism of Shutdown Heat Removal System (SHRS). The main aim is to refine the reactor pool design in terms of temperature safety margin of the sodium pool. Thus, an appropriate protection mechanism is maintained in order to ensure the safety and integrity of the reactor system during a shutdown mode without using any active heat removal system. The impacts on the pool temperature are evaluated based on the following considerations: (1) the aspect ratio of pool diameter to depth, (2) the values of thermal emissivity of the surface materials of reactor and guard vessels, and (3) innerpool liner and core periphery structures. The computational results show that an optimal pool design in geometry can reduce the maximum pool temperature down to ∼551 °C which is substantially lower than ∼627 °C as calculated for the reference case. It is also concluded that the passive Reactor Air Cooling System (RACS) is effective in removing decay heat after shutdown. Furthermore, thermal radiation from the surface of the reactor vessel is found to be important; and thus, the selection of the vessel surface materials with a high emissivity would be a crucial factor for consideration in safety design. This study provides future researchers with a guideline on designing safety measures for the fourth generation of the fast reactors with no particular reference to any specific manufacturer.     

Design and preliminary test of an annular linear induction electromagnetic pump for a sodium-cooled fast reactor thermal hydraulic experiment

An annular linear induction electromagnetic pump (ALIP) with a flow rate of 2265 L/min and a developed pressure of 4 bar was designed and fabricated to test the performance of the components of a sodium-cooled fast reactor (SFR) in a sodium thermal hydraulic experimental loop. The design characteristic of the ALIP was calculated using the electrical equivalent circuit method typically used for analyzing linear induction machines. Preliminary tests, such as verification of the moving function using an annular Al pipe, were carried out. The linearity between the input voltage, current, and magnetic flux density was verified. The developed force demonstrated an increase proportional to the square of the input current, whereas the velocity was linearly proportional to the input current. The main design variables of the pump were calculated theoretically for the SFR thermal hydraulic experimental loop. The pump was optimized for the design variables including input frequency, and the characteristics of the optimized pump were compared with those of the pump at the commercially used frequency of 60 Hz.     

Visualization of the ultrasound-induced behaviour of gas pockets entrapped on a patterned surface. Application to inspection of sodium-cooled fast reactors

The presence of microscopic gas pockets trapped on a non-wetting solid/liquid interface induces a significant decrease of the transmitted ultrasonic energy. This phenomenon can reduce the performance of ultrasonic inspection in sodium-cooled fast reactors (SFR) for instance. To explain this event, a hypothesis was formulated in a previous study: under the effect of ultrasound, the gas pockets could coalesce, forming a gas film which would prevent the passage of the ultrasound. This coalescence hypothesis can be studied by visual observation of the phenomenon. This paper is presenting an original experiment which simulates this phenomenon in water. This experiment consists in observing the ultrasound-induced behaviour of gas pockets over various time scales. The results allow dismissing the coalescence hypothesis. Our conclusion indicates how future works are reoriented to improve the design of ultrasonic transducers dedicated to SFR inspection.