完全自动运行的快堆

【日本《原子能产业新闻》2001年7月19日报道】 日本原子能研究所受电力中央研究所的委托,为其开发研究的可完全自动运行的超安全、超小型锂冷却快堆(RAPID-L)现已具备了开发条件。 该项研究是基于原研的公开提案招标式原子能基础研究促进制度、作为1999年度的研究课题实施的。该项研究的负责人是电力中央研究所的研究员神户满上席,参加研究的单位有三菱综合研究所、东北大学、京都大学等。 RAPID-L是“整堆芯换料反应堆”的英文名称。这种反应堆在设计研究上,以确保运行的高度安全性为目标,排除了人为因素,在世界上首次实现了传统快堆…     

热管冷却空间反应堆系统启动特性研究

为研究热管冷却空间堆系统从冷态零功率到满功率的启动特性,以典型热管冷却空间反应堆电源系统SAIRS为对象,开发了热管冷却空间反应堆系统瞬态程序TAPIRS。该程序的模型主要包括中子动力学模型、堆芯传热模型、热管模型、碱金属热电转换装置（AMTEC）能量转换模型和散热板模型。将TAPIRS程序各模块和系统稳态计算结果与参考文献计算值比较分析,验证了本文模型和求解方法的合理性。启动特性研究表明,通过控制控制鼓的转动速率,可实现反应堆从次临界到满功率、热管和AMTEC从固态到正常运行状态的启动过程。热管冷却空间堆依靠核热启动具有可行性,热管最高温度不超过1 250 K,满功率参数与相关文献的最大相对误差不超过6%。     

热管式锂冷空间快堆中子学计算分析

采用MCNP程序对锂冷热管式锂冷空间快堆建立中子学计算模型,对其中子通量密度和功率分布、有效增殖因子等进行了计算,采用分区燃料布置,得到满足长寿命运行要求的分区装载方案,利用ORIGEN2程序进行燃耗校核,计算了转鼓的价值和转鼓转角随运行时间的变化情况。模型分析结果表明:分区装料后的堆芯满足临界安全设计和不均匀系数要求;堆芯的过剩反应性足够7年不换料满功率运行;意外发射失败掉入湿沙或海水中,由于有谱移吸收体铼,堆芯仍然保持足够的次临界度;转鼓的价值可以保证堆芯在整个寿期内安全的停堆和正常的启动;热管式锂冷空间快堆基本物理特性合理,满足设计要求。     

先进空间快堆安全特性分析

以200kW空间快堆RAPID－L为对象，建立瞬态分析模型，分析了在无保护超功率事故UTOP和无保护失流事故ULOF下的瞬态特性。计算结果表明:快速型锂膨胀模块（LEM）可以随着冷却剂温度变化自动快速的响应，能够在不停堆的情况下保证反应堆的安全，因此，RAPID－L具有固有安全特性。     

堆芯换料设计模块化和图表报告标准化研制

堆芯换料设计模块化和图表报告标准化软件包括三个软件包：自动计算模块软件包、数据搜索和处理软件包和图自动生成软件包。自动计算模块软件包用于计算设计报告中所要求的内容，生成编制图表所需的直接或间接的数据文件；数据搜索和处理软件包用于从自动计算模块软件包生成的文件中读取设计报告图表所需的数据和进一步处理还不能绘制图表的间接数据，并按照用户要求的标准格式输出含有最终数据的文件；图自动生成软件包根据最终数据文件自动绘制图。     

基于顺序控制技术的反应堆自动启动系统

在深入分析反应堆启动过程各项技术操作的基础上,提出利用顺序控制技术实现反应堆自动启动的构想.对基于顺序控制技术的反应堆自动启动系统的设计思路、层次结构、系统组成和系统设计的关键技术进行了研究,并初步进行了系统设计.所设计的顺序控制系统能够实现反应堆的自动启动,减轻运行人员工作负担,从而提高了反应堆启动运行的安全性和经济性.     

高温气冷堆启动过程的模拟与分析

本文利用工程模拟机可实时计算、多系统耦合求解的技术优势,对球床模块式高温气冷堆核电站(HTR-PM)的启动过程进行模拟。分析并掌握了HTR-PM主要运行参数的变化规律和调节手段,研究了模块式高温气冷堆在启动过程中的运行特性,特别是不同模块间所体现出的耦合效应对整个系统运行特性的影响,为电厂运行方式的确定和具体运行规程的制定提供了重要参考。     

完全自动运行的超安全快堆RAPID

【日本《原子能视野》2002年特集专刊报道】 出于灵活应对将来电力需求的考虑,电力业界逐渐看好适用于分散发电的小型反应堆。它不仅可作为发达国家在临近需求地的电源,而且在发展中国家今后也将受到重视。另一方面,由于受日本JCO事故及美国“9·11”恐怖事件等的影响,人们对能够排除人为失误、发生任何故意或过失性错误都能安全运行的超安全反应堆寄予很大期望。在此背景下,可完全自动运行的、世界首座超安全快堆概念RAPID-L便应运而生了。该项研究是日本推进原子能基础研究体制的一个环节,由电力中央研究所等组成的集团受日本原子能研…     

模块式高温气冷堆核电站的运行特性研究

球床模块式高温气冷堆核电站(HTR-PM)由于具有多个模块，运行特性比单堆电站更复杂。利用集总参数方法建立了HTR-PM的动态模型，并利用该模型对电站的运行过程进行了仿真。升功率运行的仿真结果表明，蒸汽温度严重偏离了正常允许值。设计了1个基于蒸汽温度的简单控制器，仿真结果表明，该控制器能很好地对电站进行运行控制，结果令人满意。     

基于RELAP5的双功能液态锂铅实验包层模块安全分析

利用嵌入了液态锂铅（LiPb）的热工水力子模块的系统程序RELAP5/MOD3,对双功能液态锂铅（DFLL）实验包层模块（TBM）的安全特性进行评价。对DFLL-TBM及其辅助冷却系统的稳态运行工况、预期运行事件和相关事故工况进行了建模、计算和分析。计算结果表明,稳态运行时第一壁（FW）结构材料表面最高温度低于允许值550 ℃。事故工况下氦气泄漏引起的ITER真空室（VV）、窗口设备室（port cell）以及托卡马克冷却水系统大厅拱顶（TCWS vault）的增压均低于ITER要求的限值0.2 MPa。实验包层钢结构不会熔化且可通过辐射换热有效地导出衰变余热。DFLL-TBM的设计可满足ITER对其热工水力安全方面的要求。     

Overview of TBM R&D activities in India

In India, development of Lead–Lithium Ceramic Breeder (LLCB) blanket is being performed as the primary candidate of Test Blanket Module (TBM) towards DEMO reactor. The LLCB TBM will be tested from the first phase of ITER operation (H-H phase) in one-half of an ITER port no. 2. The Indian TBM R&D program is focused on the development of blanket materials and critical technologies: structural material (IN-RAFMS), breeding materials (Pb–Li, Li₂TiO₃), development of technologies for Lead–Lithium cooling system (LLCS), helium cooling system (HCS), tritium extraction system (TES) and TBM related fabrication technologies. This paper will provide an overview of LLCB TBM R&D activities under progress in India.     

ITER中国液态锂铅实验包层模块设计研究与实验策略

在广泛调研和深入分析国际聚变堆包层发展状况的基础上,根据液态锂铅包层一般特点和中国发展的系列液态锂铅包层概念设计,提出了一个具有演示氦气单冷却剂和氦气/锂铅双冷却剂包层技术的双功能包层模块实验系统方案,对其性能进行了分析研究,作为中国向ITER实验包层工作组(TBWG)提交的液态包层实验模块最终设计描述文件的内容框架。总结了该工作主要内容,包括基本设计思想和方案描述、性能分析概况、对辅助系统的要求和实验策略与关键技术等。     

RAPID Operator-Free Fast Reactor Concept without Any Control Rods Reactor Concept and Plant Dynamics Analyses

The 1,000kWe metal fueled sodium-cooled fast reactor concept “RAPID” to achieve highly automated reactor operation has been demonstrated. RAPID (Refueling by All Pins Integrated Design) is designed for a terrestrial power system which enables quick and simplified refueling. It is one of the successors of the RAPID-L, the operator-free fast reactor concept designed for lunar base power system. The essential feature of the RAPID concept is that the reactor core consists of an integrated fuel assembly instead of conventional fuel subassemblies. In this small-size reactor core, 14,000 fuel pins are integrated and encased in a fuel cartridge. Refueling is conducted by replacing a fuel cartridge. The reactor can be operated without refueling for up to 10 years.

Unique challenges in reactivity control systems design have been addressed in the RAPID concept. The reactor has no control rod but involves the following innovative reactivity control systems: lithium expansion modules (LEM) for inherent reactivity feedback, lithium injection modules (LIM) for inherent ultimate shutdown, and lithium release modules (LRM) for automated reactor startup. All these systems adopt ⁶Li as a liquid poison instead of B₄C rods. In combination with LEMs, LIMs and LRMs, RAPID can be operated without an operator. In this paper, the RAPID reactor concept and its transient characteristics are presented.     

Analysis of the reference accidental sequence for safety assessment of LLCB TBM system

Safety analysis of the reference accidental sequence has been carried out for Lead Lithium cooled Ceramic Breeder (LLCB) Test Blanket Module (TBM) system; India's prototype of DEMO blanket concept for testing in International Thermonuclear Experimental Reactor (ITER). The accidental event analyzed starts with a Postulated Initiating Event (PIE) of ex-vessel loss of first wall helium coolant due to guillotine rupture of coolant pipe with simultaneous assumed failure of plasma shutdown system. Three different variants of the sequences analyzed include simultaneous additional failures of TBM and ITER first wall, failure of TBM box resulting in to spilling of lead lithium liquid metal in to vacuum vessel and reactor trip on Loss of Coolant Accident (LOCA) signal from TBM system. The analysis address specific reactor safety concerns, such as pressurization of confinement buildings, vacuum vessel pressurization, release of activated products and tritium during these accidental events and hydrogen production from chemical reactions between lead–lithium liquid metal and beryllium with water. An in-house customized computer code is developed and through these deterministic safety analyses the prescribed safety limits are shown to be well within limits for Indian LLCB-TBM design and it also meets overall safety goal for ITER. This paper reports transient analysis results of the safety assessment.     

Normal operation and maintenance safety lessons from the ITER US PbLi test blanket module program for a US FNSF and DEMO

A leading power reactor breeding blanket candidate for a fusion demonstration power plant (DEMO) being pursued by the US Fusion Community is the Dual Coolant Lead Lithium (DCLL) concept. The safety hazards associated with the DCLL concept as a reactor blanket have been examined in several US design studies. These studies identify the largest radiological hazards as those associated with the dust generation by plasma erosion of plasma blanket module first walls, oxidation of blanket structures at high temperature in air or steam, inventories of tritium bred in or permeating through the ferritic steel structures of the blanket module and blanket support systems, and the ²¹⁰Po and ²⁰³Hg produced in the PbLi breeder/coolant. What these studies lack is the scrutiny associated with a licensing review of the DCLL concept. An insight into this process was gained during the US participation in the ITER Test Blanket Module (TBM) Program. In this paper we discuss the lessons learned during this activity and make safety proposals for the design of a Fusion Nuclear Science Facility (FNSF) or a DEMO that employs a lead lithium breeding blanket.     

ITER中国液态锂铅实验包层模块结构热应力数值模拟

使用有限元程序对中国向国际热核实验堆ITER实验包层工作组提交的双功能锂铅实验包层模块(DFLL-TBM)的两种结构设计方案即双冷LiPb包层DLL和单冷准静态LiPb包层SLL进行热应力数值模拟,在包层结构设计、热工水力学设计和中子学计算基础上,给出包层结构温度场和应力场分布,依据ITER高温结构设计标准,进一步对包层高温部件进行力学性能分析.根据这些模拟结果,分析两种结构基本设计方案的合理性和可行性,并作为进一步优化分析的基础.     

ITER双功能液态锂铅实验包层系统PIE-PIT分析

ITER要求各参与国的实验包层模块在实验前必须提交安全分析报告(含确定论分析和概率论分析),进而获取安全许可证.结合中国双功能锂铅实验包层模块的具体特点,采用了假设始发事件-潜在影响表(PIE-PIT)分析方法对DFLL-TBM进行了安全评估与分析,已验证确定论安全分析所选择的三个参考事件是否可包络PIE-PIT分析得到的严重事故序列.     

Performance evaluation of ultra-long lithium heat pipe using an improved lumped parameter model

Lithium heat pipes have broad applications in heat pipe cooling reactors and hypersonic vehicles owing to their ultra-high working temperature.In particular,when the length of the lithium heat pipe is ultra-long,the flow and heat transfer characteristics are more complex.In this study,an improved lumped parameter model that considers the Marangoni effect,bending effect,and different vapor flow patterns and Mach numbers was developed.There-after,the proposed model was verified using the University of New Mexico's Heat Pipe and HTPIPE models.Finally,the verified model was applied to simulate the steady-state operation of an ultra-long lithium heat pipe in a Heat Pipe-Segmented Thermoelectric Module Converters space reactor.Based on the results:(1)Vapor thermal resistance was dominant at low heating power and decreased with increasing heating power.The vapor flow inside the heat pipe developed from the laminar to the turbulent phase,whereas the liquid phase in the heat pipe was always laminar.(2)The vapor pressure drop caused by bending was approximately 22-23％of the total,and the bending effect on the liquid pressure drop could be ignored.(3)The Marangoni effect reduced the capillary limit by hindering the liquid reflux,especially at low vapor temperatures.Without considering the Marangoni effect,the capillary limit of the lithium heat pipe was overestimated by 9％when the vapor temperature was 1400 K.(4)The total thermal resistance of the heat pipe significantly increased with increasing adiabatic length when the vapor tempera-ture was low.Further,the wick dryness increased with increasing adiabatic length at any vapor temperature.Such findings improve on current knowledge for the optimal design and safety analysis of a heat pipe reactor,which adopts ultra-long lithium heat pipes.     

The TFTR Lithium Blanket Module program

The Lithium Blanket Module (LBM) is an approximately 80×80×80 cm cubic module, representative of a helium-cooled lithium oxide fusion reactor blanket module, that will be installed on the TFTR (Tokamak Fusion Test Reactor) in late 1986. The principal objective of the LBM Program is to perform a series of neutron transport and tritium-breeding measurements throughout the LBM when it is exposed to the TFTR toroidal fusion neutron source, and to compare these data with the predictions of Monte Carlo (MCNP) neutronics codes. The LBM consists of 920 2.5-cm diameter breeder rods constructed of lithium oxide (Li₂O) pellets housed in thin-walled stainless steel tubes. Procedures for mass-producing 25,000 Li₂O pellets with satisfactory reproducibility were developed using purified Li₂O powder, and fabrication of all the breeder rods was completed in early 1985. Tritium assay methods were investigated experimentally using both small lithium metal samples and LBM-type pellets. This work demonstrated that the thermal extraction method will be satisfactory for accurate evaluation of the minute concentrations of tritium expected in the LBM pellets (0.1–1 nCi/g).