氟盐冷却高温堆氚输运特性瞬态分析

氚是熔盐堆运行过程中的固有产物，具有强腐蚀性和渗透性，是限制熔盐堆技术发展的瓶颈问题之一。本文围绕氟盐冷却高温堆（FHR）中氚输运特性在事故工况下的瞬态响应开展研究，主要讨论在无保护反应性引入（URI）及无保护冷却剂入口过冷（UOC）事故下，熔盐堆一回路中的氚产率、石墨吸附量、熔盐溶解量、腐蚀与沉积反应以及氚向二回路的扩散等特性。研究发现，瞬态条件下氚输运特性较稳态时更为复杂多变，呈现出强烈的动态耦合特点，这对氚控设备的性能提出了更高的要求。计算表明，在URI和UOC事故下，氚向二回路的扩散速率均降低，不需投入额外的氚控安全设施。     

固态钍基熔盐堆中~(14)C的产生及释放探讨

熔盐堆作为第四代反应堆论坛推荐的6种候选堆型之一,具有输出温度高、能量密度高、无水冷却等特点。固态钍基熔盐堆(Thorium Molten Salt Reactor with Solid Fuel,TMSR-SF1)堆芯大部分结构材料为石墨,冷却剂杂质及石墨材料中的13C和杂质N、O易被活化产生14C。14C半衰期较长,同其他稳态核素12C、13C一样广泛参与各种复杂的生物循环,在反应堆中受到关注。TMSR-SF1中的14C广泛分布于冷却剂、堆芯石墨结构材料和燃料元件。本文采用输运燃耗耦合方法,应用SCALE6.1的TRITION控制模块对反应堆各区域的14C放射性活度进行计算分析,结果表明,反应堆在正常运行工况下一回路每年产生的14C放射性活度为0.34 TBq,满足现有的压水堆、重水堆管理限值要求。向环境释放的14C主要来自于一回路熔盐中N杂质的活化。     

小型氟盐冷却高温堆耦合布雷顿循环系统分析与研究

为满足小型氟盐冷却高温堆（FHR）能量转换需求,开发与之匹配的高效、紧凑、无水冷却动力转换系统,本文对比了超临界二氧化碳（SCO₂）、空气、氩气（Ar）、氮气（N₂）、氙气（Xe）5种气体工质在不同布雷顿循环构型中的热电转换效率、?效率、?损失分布。研究发现,SCO₂布雷顿循环相比其它工质循环具有最高的热电转换效率和?效率,且结构更为紧凑,易于小型化和模块化,与小型氟盐冷却高温堆耦合更具优势;进而对SCO₂布雷顿循环进行构型优化,得出匹配小型氟盐冷却高温堆的最佳循环构型方式,构成固有安全模块化小型氟盐冷却高温堆热电转换系统,为西部能源利用提供新研究思路。      

1 GWt PB-FHR负荷跟踪运行特性分析

为研究熔盐堆系统在商业应用中的价值,分析其是否满足电网负荷的变化需求和安全运行的能力,本文以1 GWt球床式氟盐冷却高温堆（PB-FHR）为研究对象,仿真计算其在负荷跟踪模式下的瞬态行为和运行特性。以RELAP5/MOD4.0程序为研究工具,并植入相关的熔盐物性与计算关系式,建立氟盐冷却高温堆的热工水力系统与功率控制系统的仿真模型,对典型负荷工况参数变化情况下控制系统的响应特性进行仿真分析。结果表明：该氟盐冷却高温堆系统在设计的控制逻辑的调控下,展示出良好的负荷跟踪运行能力,堆芯功率能迅速响应负荷变化,功率超调和温度超调小,反应堆的运行参数始终处于合理的运行范围内。     

1 GWt PB-FHR负荷跟踪运行特性分析

为研究熔盐堆系统在商业应用中的价值,分析其是否满足电网负荷的变化需求和安全运行的能力,本文以1 GWt球床式氟盐冷却高温堆(PB-FHR)为研究对象,仿真计算其在负荷跟踪模式下的瞬态行为和运行特性。以RELAP5/MOD4.0程序为研究工具,并植入相关的熔盐物性与计算关系式,建立氟盐冷却高温堆的热工水力系统与功率控制系统的仿真模型,对典型负荷工况参数变化情况下控制系统的响应特性进行仿真分析。结果表明:该氟盐冷却高温堆系统在设计的控制逻辑的调控下,展示出良好的负荷跟踪运行能力,堆芯功率能迅速响应负荷变化,功率超调和温度超调小,反应堆的运行参数始终处于合理的运行范围内。     

采用SCALE计算氟盐冷却高温堆产氚量的一些问题

氟盐冷却高温堆(Fluoride-salt-cooled High-temperature Reactor,FHR)是以熔融的氟盐(2Li F-Be F2,FLi Be)作为冷却剂、TRISO(Tri-structural Isotropic)颗粒为燃料、石墨为慢化剂的热中子反应堆。FLi Be冷却剂中的6Li、7Li、9Be和19F等核素在中子的辐照下产生氚。氚于高温下在金属材料中具有较强的渗透性,可能对操作人员及公众造成放射性危害。氚与F原子结合生成具有强腐蚀性的TF,可能影响结构材料的力学性能和使用寿命。本文采用SCALE5.1和SCALE6(包括6.1、6.1.2和目前最新发布版本6.1.3,后文若无特别说明,SCALE6均指此三个版本)中的TRITON(Transport Rigor Implemented with Time-dependent Operation for Neutronic depletion)模块计算了典型FHR的产氚量。计算结果表明,在Li元素的产氚计算中,SCALE6的TRITON模块由于对Li元素处理不当,给计算结果带来较大差异,影响了计算的正确性。在计算9Be和19F的产氚量时,由于不同版本的SCALE采用不同的反应通道等原因,计算结果也存在差异。综合分析表明,通过SCALE计算FHR产氚量时应当优先采用SCALE6,为修正Li元素处理不当的问题,需使用TRITON中经COUPLE子模块更新整合后的截面库(TRITON默认为ft33f001)来单独运行ORIGEN-S模块。     

熔盐堆中石墨吸附氚的理论研究

为明确石墨表面的氟、氟化氚中的氟和单空位缺陷对氚吸附的影响机理,利用第一性原理模拟研究熔盐堆中石墨对氚的吸附,研究了在完整石墨和单空位石墨表面上氢原子(离子)、氟原子(离子)、氟气分子与氟化氢分子的吸附行为。研究发现:当氟和氢同时存在的时候,氟会优先和石墨表面的碳原子成键;石墨表面的单空位促使氟化氢在单空位处解离,以化学吸附形成存在;石墨表面吸附氢(或者氚)、氟后,表面会发生畸变;可预测熔盐堆运行过程中,氚会优先在反应堆中的石墨表面的单空位等反应活性强的地方吸附、积聚。     

熔盐实验堆备用停堆方案及其可行性分析

反应性控制系统的设计是反应堆物理设计的主要内容之一。熔盐堆采用熔融的氟化盐混合物作为燃料,由于核燃料的特殊性,熔盐堆在反应堆设计方面与传统固体燃料反应堆有着较大区别。本文鉴于熔盐堆的特殊性,针对2 MW液态燃料钍基熔盐堆(Thorium Molten Salt Reactor-Liquid Fuel,TMSR-LF1),提出多种停堆方式,包括排燃料盐、套管中注中子毒物、改变燃料盐成分、改变堆芯石墨栅元数,并进行了计算分析。分析结果表明:往套管中注入中子毒物是在控制棒失效的情况下很好的替换停堆方式;燃料盐成分可调,是熔盐堆本身具有的特点,因此往燃料盐中添加BF_3、LiF-BeF_2-ZrF_4、LiF-ThF_4,是调节堆芯反应性很好的方式;改变石墨栅元数也可以使反应堆停堆。本研究分析可以为熔盐堆停堆方式提供技术储备和理论参考。     

液态熔盐堆运行安全特性初步研究

液态燃料反应堆与固态燃料反应堆相比,原理上有较大不同。液态熔盐堆中由于燃料流动带走缓发中子先驱核在堆外衰变导致堆芯反应性降低,且裂变产物在堆外回路中衰变也会引起一回路发热。本文使用熔盐堆中子动力学程序Cinsf1D探讨2 MW熔盐堆的临界动力学特性和安全特性,研究零功率临界下不同熔盐流速启泵和停泵导致的缓发中子先驱核流失所需改变的控制棒棒位。同时还计算了2 MW恒定功率情况下稳态运行及降低流速时一回路温度分布,并模拟了2 MW额定功率下停泵事件。停泵后由于缓发中子损失减少反应堆功率先缓慢增加,然后迅速降低到接近余热水平。停泵后堆芯温度缓慢增加后稳定在安全值以内,说明熔盐堆具有本征安全性。     

FHR在全厂断电事故下对RVACS散热能力的要求

氟盐冷却高温堆(Fluoride salt-cooled High-temperature Reactor,FHR)是一种采用包覆颗粒燃料、高温熔融氟盐冷却剂的先进反应堆。部分FHR概念采用了反应堆容器辅助冷却系统(Reactor Vessel Auxiliary Cooling System,RVACS)导出事故下的堆芯余热。RVACS通过导热、对流换热、辐射换热等非能动过程,在事故发生时将堆芯余热排出至大气中。本文采用中国科学院上海应用物理研究所设计的10 MW FHR作为基准,利用RELAP5-MS程序,对其在全厂断电事故下的瞬态过程进行了模拟,验证了RVACS的余热导出能力。本文进一步研究了高反应堆功率情况下的全厂断电事故的瞬态过程,探讨了不同反应堆功率的FHR对RVACS散热能力的要求。     

Design and licensing strategies for the fluoride-salt-cooled,high-temperature reactor (FHR) technology

Fluoride-salt-cooled, high-temperature reactor (FHR) technology combines the robust coated-particle fuel of high-temperature, gas-cooled reactors with the single phase, high volumetric heat capacity coolant of molten salt reactors and the low-pressure pool-type reactor configuration of sodium fast reactors. FHRs have the capacity to deliver heat at high average temperature, and thus to achieve higher thermal efficiency than light water reactors. Licensing of the passive safety systems used in FHRs can use the same framework applied successfully to passive advanced light water reactors, and earlier work by the NGNP and PBMR projects provide an appropriate framework to guide the design of safety-relevant FHR systems. This paper provides a historical review of the development of FHR technology, describes ongoing development efforts, and presents design and licensing strategies for FHRs. A companion review article describes the phenomenology, methods and experimental program in support of FHR.     

Diffusion analysis on the thermal release of tritium from a neutron absorbing material

Tritium released from neutron irradiated borosilicate glass was determined by a specially designed sampling system and a liquid scintillation counter at temperatures in the range of 200–700°C. It was found that the chemical form of tritium released was tritiated water (HTO, T₂O) for the most part. Tritium produced in the glass would react with oxygen to form OT and diffuse out by a similar mechanism as the molecular diffusion of water in glasses. The diffusion coefficient of tritiated water in borosilicate glass obtained is expressed by D (cm²/s) = 5.3 × 10⁻⁴ exp( −128 kJ/mol)/RT). It is concluded from the diffusion analysis that the greater part of tritium produced in a neutron absorber, which is made of borosilicate glass, would remain in the glass for a few years of irradiation.     

Tritium diffusivity in crystal grain of Li2TiO3 and tritium release behavior under several purge gas conditions

It has been reported by the present authors that behavior of tritium release from solid breeder grain is consisted of diffusion in grain, tritium transfer at surface layer and surface reactions on grain surface such as adsorption or isotope exchange reactions. Tritium release curves estimated using the tritium release model gave good agreement with observed tritium release curves from Li₄SiO₄, Li₂ZrO₃ or LiAlO₂.

Tritium release behavior from Li₂TiO₃ under humid purge gas, dry purge gas and dry purge gas with hydrogen conditions is discussed in this study, tritium release curves using the release model that we proposed previously give a good agreements with experimental tritium release curves. Tritium effective diffusivity in the crystal grain of Li₂TiO₃ is also estimated in this study using a curve-fitting method applied to the release curves obtained under the humid purge gas condition. It is discussed that change of color of Li₂TiO₃ surface under hydrogen purge gas condition is observed and this phenomenon might affect tritium release behavior from Li₂TiO₃.     

LaNi_(4.25)Al_(0.75)吸氕、氘、氚热力学同位素效应

采用自制的全金属氢化物吸放氢实验装置,恒温等容条件下测定LaNi_(4.25)Al_(0.75)材料吸氕、氘、氚单质气体的压力-组成等温线(P-C-T曲线),并根据Van’t Hoff(范特霍夫)方程得到LaNi_(4.25)Al_(0.75)吸氕、氘、氚形成氢化物相的热力学参数焓变ΔH分别为:-44.5、-45.0、-47.1kJ·mol-1,熵变ΔS分别为:-118.0、-121.8、-127.5J·mol~(-1)·K~(-1)。结果表明:LaNi_(4.25)Al_(0.75)材料吸收氕、氘、氚单质气体,在温度较低时,同位素效应不明显;温度高于100℃时,热力学同位素效应显著。相同温度、吸气容量条件下,吸气平衡压力从低到高依次是氕、氘和氚,其反应焓变和熵变从小到大依次是氚、氘和氕。结果表明,LaNiAl合金吸氢的热力学同位素效应依赖于温度的变化。     

Neutronic Comparison of Tritium－Breeding Performance of Candidate Tritium－Breeding Materials

Tritium self-sustainment, which will meet the fuel requirement of fusion reactor, is one of the key issues of fusion power development. The tritium breeding performances of various tritium-breeding materials are compared based on a series of neutronics calculations using three-dimensional Monte Carlo neutron-photon transport code MCNP/4C with the IAEA FENDL-2 data library. The effects of the dimensions of the tritium-breeding zone and the enrichment of 6Li on Tritium Breeding Ratio (TBR) are analyzed. The effects of Be as a neutron multiplier on TBR are also calculated.     

The soret effect and its implications for fusion reactors

Tritium permeation through and retention in fusion reactor structures may be strongly influenced by the heat load carried by the structures through the Soret effect. After a short discussion suggestive of a heuristic model for predicting the associated energy and the heat of transport, data from several experiments are analyzed to show that the simplistic model works reasonably well with endothermic materials such as Fe and Ni, but is less successful with hydride formers. The implications of the model for tritium permeation and retention are discussed, and sample calculations are presented to illustrate the importance of properly accounting for the Soret effect in predicting tritium permeation and retention in fusion reactor structures. Neglecting the Soret effect may result in order of magnitude errors in estimating permeation and retention, while accounting for temperature sensitivity in the heat of transport will result in less significant corrections. An Appendix summarizes the development of transport equations from non-equilibrium thermodynamics to clarify the relationships between the various transport parameters involved.     

Tritium issues to be solved for establishment of a fusion reactor

In order to establish a D–T fusion reactor as an energy source, economical conversion of fusion energy to electricity and/or heat, attaining enough margins in tritium breeding, and insuring tritium safety must be simultaneously achieved. Scientists and researchers working on Tritium in Japan are now tackling with T related problems. Their research subjects can be categorized into two, i.e. researches on “Science and technology” to establish safe and economic Tritium fuel cycle for fusion reactors and “Tritium safety”. Many researchers from various universities, and institutes such as NIFS, JAEA and IEA (Inst. Environmental Science) in Japan are involved in various research programs. In this report, after brief introduction on Tritium related researches in Japan, important T issues to be solved for establishment of a fusion reactor will be summarized considering the handling of large amount of tritium, i.e. fuelling, D–T burning, T inventory, exhausting, refinement, confinement, permeation, leakage, contamination, regulation and tritium accountancy.     

CLAM钢中氕氘的渗透与去除

低活化铁素体/马氏体钢（RAFM钢）是聚变堆产氚包层的优选结构材料。氢同位素在结构材料中的扩散渗透特性关系到产氚回收率、燃料循环及运行安全。本工作对国内研发RAFM钢之一的CLAM钢进行了气体驱动的氘渗透实验,得到573～873 K温度范围内氘的宏观溶解度S（mol/(m³•Pa^0.5)）为0.264exp(-22 447/RT),扩散系数D（m²/s）为1.38×10^-7exp(-17 271/RT),渗透率Φ（mol/(m•s•Pa^0.5)）为3.64×10^-8exp(-39 718/RT)。还进行了氕氘气体混合物的渗透实验,确认了渗透同位素效应;探索了钢中溶解氘的真空热释放去除。     

PdY合金管束的氢渗透性能

PdY合金膜因其具有良好的透氢性能与机械性能,有望应用于聚变堆氢同位素纯化工艺。基于PdY合金膜的服役参数及氚安全要求,有必要研究在低氢压下PdY合金膜的氢同位素渗透特性,为后续设计氢同位素纯化组件提供数据支撑。本工作基于直管外压式PdY扩散器,研究了低氢压（<50 kPa）、工作温度为350~450 ℃条件下,厚度为80 μm的PdY合金薄膜管的氢渗透速率与膜两侧压力、工作温度的关系。结果表明,低氢压下,PdY合金膜的氢渗透规律符合J=Φpⁿ_H-pⁿ_Ld,且压力指数n等于0.9,渗透速率控制机制主要表现为表面过程控速;提高工作温度使得合金膜的渗透通量增大,且温度对扩散过程的影响更大,使渗透过程更加趋于表面控速。此外,计算了该工作温度范围下的渗透系数,并通过阿伦尼乌斯公式推导求得渗透活化能约为24.54 kJ/mol,渗透常数Φ₀为5.86×10^-6 mol/(m·s·kPa^0.9)。低氢压下,该厚度膜的渗透系数可由5.86×10^-6e^-24.54/^(RT) mol/(m·s·kPa^0.9)进行计算。