池式钠冷快堆二回路钠比活度研究

池式钠冷快堆中的二回路钠比活度影响二回路钠工艺间中的辐射剂量及分区、二次钠泄漏事故的环境影响和运行维修期间工作人员接受的辐射剂量。二回路钠比活度水平是由堆本体屏蔽决定的。本文以中国实验快堆(CEFR)为研究对象,通过研究它们之间的关系,找到设计池式钠冷快堆时确定二回路比活度设计值的方法。研究表明,二回路工艺间中冷阱间的剂量当量率决定了二回路钠比活度值的上限;而从堆容器材料损伤和二次钠事故后果等方面限制推算出的比活度限值远高于上限值,可以据此规律来确定中国示范快堆的二回路比活度设计值,并通过堆本体的屏蔽达到此设计值。     

池式快堆系统分析软件稳态功能开发

针对目前我国快堆系统分析软件主要采用国外引进方式而导致难以掌握核心物理模型的现状,以中国实验快堆(CEFR)为研究和建模对象,基于中子动力学模型、堆芯及其热钠池模型、中间热交换器模型、一回路和中间回路热量传输系统模型、三回路模型等,自主开发了基于CompaqVisualFortran(CVF)的适用于稳态计算的池式快堆系统分析软件SAC-CFR。通过与中国实验快堆安全分析报告中数据进行对比,验证了所开发模型的精度,为下一步瞬态模型的开发及控制和保护系统的开发做准备。     

中国实验快堆2018年度运行年报

正中国实验快堆(CEFR)热功率为65 MW,试验发电功率为20 MW,首炉燃料使用UO2,采用堆本体池式结构和钠-钠-水三回路热传输系统,并首次设立独立的非能动事故余热排出系统。2018年反应堆处于冷停堆运行状态,继续进行大修遗留工作及大修调试工作,完成系统恢复与功能鉴定,完成3次开堆前检查和常规岛热态运行,对开堆强相关项进行处理和验证,实现冷停堆运行工况下的全厂安全稳定运行。     

基于集总参数法的池式钠冷快堆无保护失流工况分析

快堆(Fast Reactor)具有燃料利用率高、可嬗变核废料的优势,是目前较为理想的先进堆型之一。快堆广泛采用池式回路布置,因此对池式快堆(Pool-type Fast Reactor)进行安全分析具有重要意义。本文采用集总参数法建立池式快堆的一回路模型,基于MATLAB编写核热耦合程序并对其进行无保护失流事故工况的安全分析,并将计算结果与实验值及其他机构计算结果进行了对比。结果显示:集总参数法的计算结果与实验和其他机构计算值均符合较好,验证了程序的可靠性。使用该程序可对池式钠冷快堆在无保护失流事故中的堆芯行为与固有安全性做出较为准确的预估计算。     

钠冷快堆载热系统设计

本文概述钠冷快堆的载热系统,综合快堆冷却系统的设计原则和方法,提供某些重要数据。对钠冷快堆一次钠回路系统的“池式”和“管式”布置方案作了分析比较,认为实验快堆采用“管式”双回路载热系统比较合适。     

中国先进研究堆回路系统设计总结

目前，研究堆的类型曰趋多样化，有重水堆、轻水堆、气冷堆和正在研制的核聚变堆，不同的堆型，回路系统的配置相差很大，如101重水堆有与重水氦气有关的几个回路系统，49．2游泳池式堆也有与轻水有关的几条回路。但采用轻水作冷却剂，重水作反射层的堆，至少需设置十几条回路。CARR是一座轻水作冷却剂、重水作反射层的研究堆，回路系统设计时主要参考了国内外一些研究堆，如HWRR、ORPHEE堆，HANARO堆、FRM—Ⅱ等。     

日本池式快堆通过地震试验

【英国《国际核工程》1984年10月号第3页报道】继日本电力工业中心研究所肯定了池式快堆之后,日本联合电力公司正计划进入快堆长期计划的下一阶段。这一阶段从1984年开始到1986年结束,在此期间,将进行回路式和池式快堆的概念设计,以便从中做出选择。     

CARR重水冷却系统的优化设计

中国先进研究堆(CARR)是轻水冷却和慢化、重水反射的池式反应堆。本文着重介绍了重水冷却系统自可行性报告以来对该系统所进行的优化设计,包括系统工艺流程及设备设计等的优化。     

中国实验快堆2019年度运行年报

<正>中国实验快堆(CEFR),热功率为65 MW,试验发电功率为20 MW,首炉燃料使用UO2,采用堆本体池式结构和钠-钠-水三回路主热传输系统,并首次设立独立的非能动事故余热排出系统。2019年度CEFR未发生上级别运行事件,未造成设备损坏和放射性释放,实现全年安全稳定运行。     

池式快堆三维热工流体力学设计研究

随着计算机软硬件技术的发展,三维数值分析技术已经成为池式快堆堆芯和钠池热工设计和计算分析的重要组成部分,并在其中发挥着不可替代的作用.通过对池式快堆几个典型热工现象的分析,展示了我国第一座池式快堆(中国实验快堆)热工设计和安全分析中所拥有的设计手段和工具,总结了三维数值分析技术在快堆工程中的应用,并指出了其对今后快堆热工设计的重要意义.     

The effect of high pressure test loop (HPTL) on the Reactivity of ET-RR-2 reactor

The second Egyptian Research Reactor ET-RR-2 is a multipurpose research reactor. It is an open pool type, with nominal power of 22 MW water-cooled. The reactor pool is designed to accommodate two fuel test loops mainly 500 and 20 KW loop in the reactor reflector to enable performing experiments on the behavior of fuel rods for nuclear reactors under their operating conditions. For that, inserted high-pressure test loop (HPTL) loaded with suggested CANDU type fuel element in the reactor core is important to achieve the above reason. From the neutronic safety point of view, it is necessary to study the mutual neutronic and reactivity effect between the reactor core and HPTL. This paper aimed at the study of the temperature coefficients of fuel and moderator of the CANDU type fuel element at different ²³⁵U enrichments, and the effect of HPTL on the reactor core reactivity. The effect of flooding the contact second shut down system (SSS) chamber with water and gadolinium nitrate on the reactor core reactivity in the presence of HPTL. All analysis was performed with the WIMSD4 and DIXY2 codes. This study shows that, an unacceptable change of reactor core reactivity was found due to the presence of the HPTL and the maximum inserted reactivity does not exceed 527 pcm at high possible ²³⁵U enrichment (10%).     

SADHANA facility for simulation of natural convection in the SGDHR system of PFBR

The Prototype Fast Breeder Reactor (PFBR) is a 500 MWe sodium cooled pool type fast reactor being constructed at Kalpakkam, India. PFBR has all the reactor components immersed in the pool of sodium and the fission heat generated in the core, is removed by the sodium circulating in the pool. During normal operation this fission heat is transferred by primary sodium to secondary sodium, which in turn transfers the heat to water in the steam generator for producing steam. The removal of the decay heat generated in the reactor core after the reactor shutdown is also very important to maintain the structural integrity of reactor core components. PFBR employs two independent systems namely, Operational Grade Decay Heat Removal system (OGDHRS) and Safety Grade Decay Heat Removal System (SGDHRS) for decay heat removal. SGDHR system is a passive system working on natural convection to ensure the core coolability even under station blackout condition. It is very important to study the thermal hydraulic behavior of Safety Grade Decay Heat Removal system of PFBR to ensure its reliable operation. A scaled down model of the circuit, named SADHANA has been modeled, designed, constructed and commissioned for demonstration and evaluation of these systems. The facility has completed around 2000 h of high temperature operation. The performance of the experimental system is satisfactory and it meets all the design requirements. At 550 °C sodium pool temperature in test vessel the secondary sodium loop generated a sodium flow of 6.7 m³/h. These experiments have revealed the adequacy and capability of SGDHR system to remove the decay heat from the fast breeder reactor core after its shutdown.     

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。     

中国先进研究堆(CARR)回路系统总体设计特点

研究堆设计不同于核电站,因需要满足不同的使用要求,而具有各自的特点,CARR也不例外。CARR是一座轻水作冷却剂、重水作慢化剂的研究堆,因此CARR回路设计围绕堆本体以反应堆冷却剂系统和重水冷却系统为主体共设置了18条回路系统,本文就CARR回路系统的总体设计特点进行了总结与介绍。     

池式钠冷快堆电厂运行方案仿真研究

以Matlab软件Simulink为仿真软件平台,通过理论分析、推导,建立了池式钠冷快堆电厂各主要系统的模型,包括:堆芯物理模型、堆芯热工模型、冷、热钠池模型、栅板联箱模型、中间热交换器模型、管道、泵模型及蒸汽发生器模型;同时建立了以步进电机作为驱动电机的功率调节系统模型并采用闭环控制来控制步进电机运行.基于这些模型构...     

Natural circulation characteristics of lead-based reactor under long-term decay heat removal

The natural circulation of primary coolant plays an important role in removing the decay heat in Station-Black-out (SBO) accident from reactor core to decay heat removal systems, such as RVACS and PHXS cooling, for lead-based reactor. In order to study the natural circulation characteristics of primary coolant under Reactor Vessel Air Cooling System (RVACS) and primary heat exchangers (PHXs) cooling, which are crucial to the safety of lead-based reactors. A three-dimensional CFD model for the China Lead-based Research Reactor (CLEAR-I) has been built to analyze the thermal-hydraulics characteristics of primary coolant system and the cooling capability of the two systems. The abilities of the two cooling systems with different decay heat powers were discussed as well. The results demonstrated that the decay heat could be removed effectively only relying on either of the two systems. However, RVACS appeared the obvious thermal stratification phenomenon in the cold pool. Besides, with the increase in decay heat power, the natural circulation capacity of primary coolant between the two systems had a significant difference. The PHXs cooling system was stronger than the RVACS, with respect to the mass flow of primary coolant and the average temperature difference between cold pool and hot pool.     

Experimental investigation of weir instability in main vessel cooling system of 1/4 FBR model

The 500 MWe Prototype Fast Breeder Reactor (PFBR) is under construction at Kalpakkam, India. The main vessel of this pool type reactor acts as the primary containment in the reactor assembly. In order to keep the main vessel temperature below creep range and to reduce high temperature embrittlement and also to ensure its healthiness for 40 years of reactor life, a small fraction of core flow (0.5 m³/s) is sent through an annular space formed between the main vessel and a cylindrical baffle (primary thermal baffle) to cool the vessel. The sodium after cooling the main vessel overflows the primary baffle (weir shell) and falls into another concentric pool of sodium separated from the cold pool by the secondary thermal baffle and then returned to cold pool. The weir shell, where the overflow of liquid sodium takes place, is a thin shell prone to flow induced vibrations due to instability caused by sloshing and fluid-structure interaction. A similar vibration phenomenon was first observed during the commissioning of Super-Phenix reactor. In order to understand the phenomenon and provide necessary experimental back up to validate the analytical models, weir instability experiments were conducted in a 1:4 scale stainless steel (SS) model installed in a water loop. The experiments were conducted with flow rate and fall height as the varying parameters. The experimental results showed that the instability of the weir shell was caused due to fluid structure interaction. This paper discusses the details of the model, the modeling laws, similitude criteria adopted, analytical prediction, the experimental results and conclusion.     

Reactor core power measurement using Cherenkov radiation and its application in Tehran Research Reactor

Cherenkov radiation is a process that could be used as an extra channel for power measurement to enhance redundancy and diversity of a reactor. This is especially easy to establish in a pool type research reactor. A simple photo diode array is used in Tehran Research Reactor to measure and display power in parallel with the existing conventional detectors. Experimental measurements on this channel showed that a good linearity exists above 100 kW range. The system has been in use for more than a year and has shown reliability and precision. Nevertheless, the system is subject to further modifications, in particular for application to lower power ranges.     

Assessment of flow induced vibration in a sodium-sodium heat exchanger

The 500 MWe Prototype Fast Breeder Reactor (PFBR) is under construction at Kalpakkam. It is a liquid metal sodium cooled pool type fast reactor with all primary components located inside a sodium pool. The heat produced due to fission in the core is transported by primary sodium to the secondary sodium in a sodium to sodium Intermediate Heat Exchanger (IHX), which in turn is transferred to water in the steam generator. PFBR IHX is a shell and tube type heat exchanger with primary sodium on shell side and secondary sodium in the tube side. Since IHX is one of the critical components placed inside the radioactive primary sodium, trouble-free operation of the IHX is very much essential for power plant availability. To validate the design and the adequacy of the support system provided for the IHX, flow induced vibration (FIV) experiments were carried out in a water test loop on a 60° sector model. This paper discusses the flow induced vibration measurements carried out in 60° sector model of IHX, the modeling criteria, the results and conclusion.