Transition toward thorium fuel cycle in a molten salt reactor by using plutonium

The molten salt reactor(MSR), as one of the Generation Ⅳ advanced nuclear systems, has attracted a worldwide interest due to its excellent performances in safety, economics, sustainability, and proliferation resistance. The aim of this work is to provide and evaluate possible solutions to fissile 233 U production and further the fuel transition to thorium fuel cycle in a thermal MSR by using plutonium partitioned from light water reactors spent fuel. By using an in-house developed tool, a breeding and burning(B&B) scenario is first introduced and analyzed from the aspects of the evolution of main nuclides, net 233 U production, spectrum shift, and temperature feedback coefficient. It can be concluded that such a Th/Pu to Th/^(233)U transition can be accomplished by employing a relatively fast fuel reprocessing with a cycle time less than 60 days. At the equilibrium state, the reactor can achieve a conversion ratio of about 0.996 for the 60-day reprocessing period(RP) case and about 1.047 for the 10-day RP case.The results also show that it is difficult to accomplish such a fuel transition with limited reprocessing(RP is 180 days),and the reactor operates as a converter and burns the plutonium with the help of thorium. Meanwhile, a prebreeding and burning(PB&B) scenario is also analyzed briefly with respect to the net 233 U production and evolution of main nuclides. One can find that it is more efficient to produce 233 U under this scenario, resulting in a double time varying from about 1.96 years for the 10-day RP case to about 6.15 years for the 180-day RP case.

Transmutation of 129I in a single-fluid double-zone thorium molten salt reactor

Herein, we assess the129I transmutation capability of a 2250-MWt single-fluid double-zone thorium molten salt reactor(SD-TMSR) by considering two methods. One is realized by loading an appropriate amount of129I before the startup of the reactor, and the amount of129I during operation is kept constant by online feeding129I.The other adopts only an initial loading of129I before startup, and no other129I is fed online during operation.The investigation first focuses on the effect of the loading of I on the Th-233U isobreeding performance. The results indicate that a233U isobreeding mode can be achieved for both scenarios for a 60-year operation when the initial molar proportion of LiI is maintained within 0.40% and 0.87%, respectively. Then, the transmutation performances for the two scenarios are compared by changing the amount of injected iodine into the core. It is found that the scenario that adopts an initial loading of129I shows a slightly better transmutation performance in comparison with the scenario that adopts online feeding of129I when the net233U productions for the two scenarios are kept equal. The initial loading of129I scenario with LiI = 0.87% molar proportion is recommended for129I transmutation in the SD-TMSR,and can transmute 1.88 t of129I in the233U isobreeding mode over 60 years.

Large eddy simulation of unsteady flow in gas-liquid separator applied in thorium molten salt reactor

Axial gas-liquid separators have been adopted in fission gas removal systems for the development of thorium molten salt reactors. In our previous study, we observed an unsteady flow phenomenon in which the flow pattern is directly dependent on the backpressure in a gas-liquid separator; however, the underlying flow mechanism is still unknown. In order to move a step further in clarifying how the flow pattern evolves with a variation in backpressure, a large eddy simulation(LES) was adopted to study the flow field evolution. In the simulation, an artificial boundary was applied at the separator outlet under the assumption that the backpressure increases linearly. The numerical results indicate that the unsteady flow feature is captured by the LES approach, and the flow transition is mainly due to the axial velocity profile redistribution induced by the backpressure variation. With the increase in backpressure,the axial velocity near the downstream orifice transits from negative to positive. This change in the axial velocity sign forces the unstable spiral vortex to become a stable rectilinear vortex.

Burnup optimization of once-through molten salt reactors using enriched uranium and thorium

The advantages of once-through molten salt reactors include readily available fuel,low nuclear proliferation risk,and low technical difficulty.It is potentially the most easily commercialized fuel cycle mode for molten salt reactors.However,there are some problems in the parameter selection of once-through molten salt reactors,and the relevant burnup optimization work requires further analysis.This study examined once-through graphitemoderated molten salt reactor using enriched uranium and thorium.The fuel volume fraction(VF),initial heavy nuclei concentration(HN_(0)),feeding uranium enrichment(E_(FU)),volume of the reactor core,and fuel type were changed to obtain the optimal conditions for burnup.We found an optimal region for VF and HN_(0) in each scheme,and the location and size of the optimal region changed with the degree of E_(FU),core volume,and fuel type.The recommended core schemes provide a reference for the core design of a once-through molten salt reactor.

Design and flow field analysis for visualization experiment facility of pebble bed based on molten salt reactor

Molten salt pebble bed reactor is one of the sixth-generation Ⅳ reactor types. To investigate the mechanical behavior of the fuel pebbles in the core, a visualization experiment facility of pebble bed(VEFPB) is designed. To obtain a uniform flow field of the core and analyze the influence of the flow field on the structure of the pebble bed, computational fluid dynamics software Fluent is used to simulate the flow field distribution of the core of VEFPB. The simulation results show that the disturbance at the bottom of the pebble bed is proportional to the flow velocity of the inlet pipe, and the flow velocity close to the inlet side is more significant than that in other parts; the design of the cylinder bottom plate with holes of different sizes can effectively reduce the flow velocity and the disturbance at the bottom of the pebble bed. In addition,according to the velocity contours of the core of VEFPB, it is observed that the flow field distribution of the core is considerably uniform except at the bottom of the pebble bed. This ensures the stability of the pebble bed and verifies the rationality of the design of VEFPB. This study provides the technical support and reference for the flow field analysis of the core of molten salt pebble bed reactor.

TMSR核功率控制系统的PID设计与仿真

功率控制系统(Power Control System,PCS)是反应堆控制系统(Reactor Control System,RCS)的重要组成部分,它完成功率提升、功率保持与功率调节的作用。在钍基熔盐堆(Thorium Molten Salt Reactor,TMSR)核能系统固态堆设计方案中,功率控制器根据实测功率与设定功率值之间的偏差和偏差的变化趋势,按照经典的比例-积分-微分(Proportional Integral Derivative,PID)控制算法,给出调节控制棒的运动距离和运动方向等信号。PCS的PID算法设计与基于反应堆中子物理、热工及控制棒的传动性能构成的闭环控制系统的特性有关,其不同参数的确定与系统的静态和动态性能指标的要求相对应。本文从控制的角度出发,在已有的控制棒样机中设计的棒控棒位系统及相关中子物理的基础上对PCS的PID算法进行多层次仿真与参数分析,并对系统的可控性与可测性进行分析验证。分析及仿真结果表明两种控制模型下的系统均是完全可控及完全可测的,在合适的PID参数集下均能体现响应的快速性及系统的良好鲁棒性和抗干扰能力,具有实际的应用意义。

钍铀燃料循环专用核数据库CENDL-TMSR-V1的基准检验

中国科学院上海应用物理研究所委托中国核数据中心研制了钍铀燃料循环专用核数据库CENDL-TMSR-V1,用于钍基熔盐实验堆的临界计算及屏蔽设计。为验证该数据库的可靠性,根据钍基熔盐实验堆的特点,从国际核临界安全手册中挑选了一系列基准实验装置进行基准检验。检验结果表明,绝大部分基准装置的k eff计算结果与实验数据的相对误差在0.5%以内,证明CENDL-TMSR-V1具有较好的可靠性和适用性,可用于钍基熔盐实验堆的物理设计。

面向对象的TMSR设计数据管理平台

为提高研究堆工程的协同设计能力,通过调研核电设计院的工作流方法、协同管理方式和软件平台的功能架构,建立了钍基熔盐堆核能系统(TMSR)项目计划管理、流程控制、质量监督为核心的设计数据管理平台,实现了面向对象的设计元数据结构。提出了“从点入手、以点带线、以线促面”的设计管理方法,利用Postgres数据库、Node.js数据交互和Web端数据可视化的技术,实现了元数据的上传、共享、管控、多维度信息统计等操作,以及基于时间轴设计数据的追溯和控制。最终模拟应用在钍基熔盐堆仿真验证装置上,通过设计文件的计划清单管理、开口项管理和接口管理等活动,证明了该系统数据模型设计的可行性。

基于数字孪生的TMSR-SF0数据监控与可视化方案

目前核电运行管理系统存在数据接口不统一、设备状态数字化表达程度差等缺陷,为适应未来智慧核电的需要,将数字孪生理论与实时数据对接技术、图形可视化手段相结合,基于钍基熔盐固态仿真堆(Thorium Molten Salt Reactor-Solid Fuel,TMSR-SF0)实例,提出一套完整的数据监控与可视化技术方案。首先,建立熔盐堆反应装置数字映射模型,并完成在Unity引擎的模型对接及虚拟场景渲染;其次,基于Node-EPICS事件驱动与Socket.io套接字实现时空数据关联;最后,基于XCharts可视化框架提出集中显示实时数据的可视化方法,保证数据的可解释性,便于对数据的实时分析。经实践验证,该方案为TMSR-SF0的数据监控系统开发提供了有效技术支撑,数据更新周期为100 ms,且具备全流程数据采集、网络通信、图元动态展示等功能,有助于操作人员对核反应装置的在线监视与运行管理,为核电领域监控技术的数字化转型发展提供了参考。

The design of RMT-based IOC redundancy at RCPI experimental platform in TMSR

In the RCPI(rod control and position indication) system prototype of the TMSR(Thorium Molten Salt Reactor) project, EPICS(Experimental Physics and Industrial Control System) was adopted as instrumentation and control software platform. According to long time running, high availability and safety for the system, RMT(redundancy monitor task) software package for Input/Output Controller(IOC) redundancy was employed, and the driver for redundancy control was realized. Test shows that the system could achieve IOC redundancy switch-over quickly and ensure the IOC running with long-term stability.

GH3535合金表面渗铝层原位氧化工艺研究

高温工况下钍基熔盐堆中存在氚泄漏的风险,建立氚渗透屏障涂层有助于应对这一问题。采用包埋渗铝和原位氧化工艺,在GH3535合金表面制备了Al2O3/Ni-Al复合阻氚涂层,重点分析了氧化温度和真空度对氧化铝薄膜微观结构的影响。利用掠入射X射线衍射、扫描电子显微镜、透射电子显微镜等手段对氧化铝薄膜表面及截面的微观形貌、相构成进行了实验分析。实验结果表明:低氧分压能降低氧化铝薄膜的形成速度,促进形成更致密、表面平整的薄膜;高的氧化温度有利于形成α相氧化铝及更厚的氧化铝薄膜,但会大大增加表面缺陷。1.2 Pa真空度气氛、850℃氧化温度、72 h氧化时间是较优的原位氧化工艺参数,可以在GH3535合金基体表面获得性能较好的氧化铝薄膜,其相结构为γ和α相,厚度约为0.8μm,且表面致密无缺陷。

核裂变能钍基熔盐堆中应用的无氧六氟铝酸锂的制备方法

文章提出了一种六氟铝酸锂的制备方法,具体涉及一种应用于核裂变能-钍基熔盐堆无氧六氟铝酸锂的制备方法。克服现有湿法生产六氟铝酸锂方法存在的后处理复杂、产品纯度不达标以及干法生产六氟铝酸锂方法存在的产品含氧量及纯度不达标的技术问题。本发明先在过量的氢氟酸中加入碳酸锂,形成氟化锂,保持较高的温度;然后向其中加入氢氧化铝,通过加入氢氧化铝后瞬时产生的高温,让完全溶解的六氟铝酸一直与过量的锂离子以特定摩尔比进行反应从而以六氟铝酸锂的沉淀沉出,防止由于条件变化水解出氟化铝,多余的锂离子与氢氟酸在较高温度下过滤,热水水洗去除。真空梯度烘干防止水解与氧化,得到水氧含量满足要求的高纯度六氟铝酸锂。

高纯度钍基熔盐堆氟化盐制备工艺

氟化盐可用于核燃料钍基熔盐堆冷却剂,以降低核燃料泄露造成生态灾害的危险,四氟化锆因其性能稳定,是该领域应用最多的氟化盐之一。本文以ZrO_(2)与HF为原料,考察了二者投料比例、反应温度和反应时间。结果表明,m(ZrO_(2))∶m(HF)=1∶1.8,反应温度100℃,反应时间55 min,可制备纯度最高的ZrF_(4)。

He^+离子辐照后Hastelloy N合金的耐腐蚀性研究

高温、辐照以及强腐蚀所引起的Hastelloy N合金的失效问题是影响熔盐堆(Molten Salt Reactor,MSR)结构材料使用寿命中的关键问题。在常温下用4.5 MeV的He+离子辐照Hastelloy N合金,吸收剂量分别为:1×1015 He+·cm-2、5×1015He+·cm-2、1×1016 He+·cm-2,采用浸入法在700°C熔融氟化盐(FLiNaK)中进行300 h腐蚀试验,研究辐照剂量对合金耐腐蚀性的影响。利用扫描电子显微(Scanning Electron microscope,SEM)、同步辐射微束X射线荧光分析(Microbeam X-ray fluorescence,μ-XRF)对腐蚀后的样品进行分析测试。结果表明,随着辐照剂量的增大,合金的耐腐蚀性逐渐减弱。μ-XRF结果表明:Hastelloy N合金在熔融氟化盐中的腐蚀主要表现为合金中Cr元素的流失。

不同燃料组合在液态氟盐冷却高温堆中的物理性能研究

液态氟盐冷却高温堆是第四代反应堆中的一种具有极大优势的堆型,对其燃料的研究工作具有重要的意义。本工作采用SCALE5.1程序包,对六种不同燃料组合在高温球床堆中的物理性能进行了研究,分别比较了剩余反应性、等效满功率运行天数、燃耗和中子能谱等重要参数。结果显示,采用233U或235U启堆时,使用232Th的实际转换成裂变材料的量不如使用238U转换的多,并会消耗更多的核燃料;采用239Pu启堆时,使用232Th可使反应堆维持较长的时间,而使用238U却导致反应堆很快不能自持。研究表明,从节约核燃料和延长堆芯寿期的角度看,在不进行在线换料后处理的情况下,232Th在热堆中的表现不如238U,但在超热堆中238U的表现不如232Th。

紧急停堆棒落棒时间对熔盐堆反应性引入瞬态的影响

紧急停堆棒落棒时间是影响反应堆安全特性的重要参数,以2 MW钍基熔盐堆为研究对象,采用RELAP5-TMSR(Reactor Excursion and Leak Analysis Program-Thorium Molten Salt Reactor)程序,建立熔盐堆系统的瞬态行为分析模型,对控制棒提棒速度的敏感性进行分析,并重点分析探索紧急停堆棒落棒时间对熔盐堆反应性引入瞬态后果的影响规律。结果表明:即使紧急停堆棒落棒时间达到10 min,哈氏合金的最高温度也仅为708.2℃,燃料盐最高温度为709.2℃,均低于安全允许限值,表明该熔盐堆具有良好的应对反应性引入事件的能力。

钍增殖熔盐堆不同燃耗核数据不确定度分析

本文利用开发的耦合模块,将SCALE程序中的TRITON和TSUNAMI-3D模块相结合,针对1GWth钍增殖熔盐堆开展不同燃耗时期核数据引起的k_(eff)不确定度分析。结果表明:随着燃耗的增加,核数据引起的keff不确定度由0.49%增大到0.55%。初始时刻对k_(eff)不确定度影响最大的反应截面是232 Th(n,γ)(约0.35%),其次是233 U(n,f)和7 Li(n,γ)。随着燃耗的增加,^(135)Xe(n,γ)、^(143)Nd(n,γ)对k_(eff)不确定度的影响逐渐显著。各反应灵敏度系数分析表明,^(232)Th(n,γ)、^(233)U(n,f)和~7Li(n,γ)截面数据对k_(eff)不确定度影响较大,需重点改进。上述关键反应在0.02~0.5eV敏感性较强,需重点关注。

固态钍基熔盐堆堆芯核设计安全限值研究

钍基熔盐堆因其在安全性、经济性及可持续性方面的优势被列为第四代核能系统重点发展的堆型。我国已经开展了相关的研究,然而国内关于研究堆的法规导则不完整,相关规定和要求比较原则,必须结合钍基熔盐研究堆的特点进行具体研究。针对固态燃料钍基熔盐实验堆,从核安全评审的角度,对核安全评审相关的问题进行深入研究分析。依据NRC针对下一代先进核能系统制定的评审方法,充分调研轻水动力堆(压水堆、沸水堆)、正在运行的研究堆、以及同样被列为第四代反应堆重点发展堆型的钠冷快堆、高温气冷堆,参考国内外对上述堆型堆芯核设计限值的确定原则及相关依据,初步给出钍基熔盐堆堆芯核设计关键安全限值。

固态钍基熔盐堆概率安全评价始发事件分析研究

始发事件分析是反应堆概率安全评价的起点。本文以10 MW固态钍基熔盐堆(Thorium Molten Salt Reactor,TMSR-SF1)为研究对象,采用主逻辑图分析方法,基于TMSR-SF1的最新概念设计,在参考已有氟盐冷却高温堆、高温气冷堆和钠冷快堆的始发事件清单和始发事件分析理论的基础上,针对TMSR-SF1始发事件分析进行初步探索研究,初步确定了TMSR-SF1的始发事件清单,共得到了TMSR-SF1的37个始发事件(功率运行情况下),并按照故障类型分类的方法对这些始发事件进行分组,共分为6组。为TMSR-SF1下一步的深入分析研究始发事件及其概率安全评价(Probabilistic safety assessment,PSA)中事故序列分析奠定了重要基础,也为安全分析的完整性提供了支持。