Comparative Studies of JENDL-3.3, JENDL-3.2, JEFF-3, JEF-2.2 and ENDF/B-6.8 Data Libraries on the Monte Carlo Continuous Energy Modeling of the Gas Turbine-Modular Helium Reactor Operating with Thorium Fuels

One of the major benefits of the Gas Turbine-Modular Helium Reactor is the capability to operate with several different types of fuel; either Light Water Reactors waste, military plutonium or thorium represent valid candidates as possible types of fuel. In the present studies, we performed a comparison of various nuclear data libraries by the Monte Carlo Continuous Energy Burnup Code MCB applied to the Gas Turbine-Modular Helium Reactor operating on a thorium fuel. A thorium fuel offers valuable attractive advantages: low fuel cost, high reduction of actinides production and the possibility to enable the reactor to act as a breeder of fuel by the neutron capture of fertile ²³²Th. We evaluated the possibility to mix thorium with small quantities, about 3% in atomic composition, of ²³⁹Pu, ²³³U and ²³⁵U. The mass of thorium must be much larger than that one of plutonium or uranium because of the low capture cross section of thorium compared to the fission one of the fissile nuclides; at the same time, the quantity of the fissile isotopes must grant the criticality condition. These two simultaneous constraints force to load a huge mass of fuel in the reactor; consequently, we propose to allocate the fuel in TRISO particles with a large radius of the kernel. For each of the three different fuels we calculated the evolution of the fuel composition by the MCB code equipped with five different nuclear data libraries: JENDL-3.3, JENDL-3.2, JEFF-3, JEF-2.2 and ENDF/B.     

Comparative Studies of ENDF/B-6.8, JEF-2.2 and JENDL-3.2 Data Libraries by Monte Carlo Modeling of High Temperature Reactors on Plutonium Based Fuel Cycles

We performed a numerical comparative analysis of the burnup capability of the Gas Turbine-Modular Helium Reactor (GT-MHR) by the Monte Carlo Continuous Energy Burnup Code (MCB). The MCB code is an extension of MCNP that includes the burnup implementation; it adopts continuous energy cross sections and it evaluates the transmutation trajectories for over 2,400 decaying nuclides. We equipped the MCB code with three different nuclear data libraries: JENDL-3.2, JEF-2.2 and ENDF/B-6.8 processed for temperatures from 300 to 1,800K.

The GT-MHR model studied in this paper is fueled by actinides coming from the Light Water Reactors waste, converted into two different types of fuel: Driver Fuel and Transmutation Fuel. The Driver Fuel supplies the fissile nuclides needed to maintain the criticality of the reactor, whereas the Transmutation Fuel depletes non-fissile isotopes and controls reactivity excess. We set the refueling and shuffling period to one year and the in-core fuel residency time to three years.

The comparative analysis of the MCB code consists of accuracy and precision studies. In the accuracy studies, we performed the burnup calculation with different nuclear data libraries during the year at which the refueling and shuffling schedule set the equilibrium of the fuel composition. In the precision studies, we repeated the same simulations 20 times with a different pseudorandom number stride and the same nuclear data library.     

高温气冷堆氦气透平直接循环启动参数分析计算

开发了包括堆芯、蒸汽发生器、透平、压气机及换热器等模块在内的高温气冷堆氦气透平直接循环系统的稳态计算程序。对系统的启动过程进行了模拟分析,并对压气机的喘振问题进行了分析,考虑了换热能力、温度和压力的影响。结果表明：在变负荷过程中压气机有足够的安全裕度。     

高温气冷堆气体透平循环方式的技术评价

气体透平循环被认为是高温气冷堆发电的发展方向。循环方式包括直接循环，开式间接循环和闭式间接循环，工质包括氦气，空气和氮气，对于每一种循环都进行了热力学分析和优化计算，并对透平压气机进行了气动设计，研究结果表明，氦气直接循环是一个理想的选择，但是基于现有技术水平难度较大，氦气和氮气闭式间接循环是目前比较现实的方案，可以实现气体透平循环的设想并为将来的直接循环做技术积累。     

氦气试验回路中的氦净化

为了减少结构材料的腐蚀并验证高温气冷实验堆中的氦净化工艺，在氦气试验回路中设置了氦净化系统并进行了试验。试验结果表明，净化流量为５０ｍ＾３／ｈ的主要由分子筛和深活性炭床组成的氦净化系统，能把氦中２００００ｃｍ＾３／ｍ＾３的化学杂质净化到７６ｃｍ＾３／ｍ＾３以下。所采用的氦中痕量杂质分析测量技术达到１０＾－１ｃｍ＾３／ｍ＾３精度。     

高温气冷堆主氦风机惰转特性研究

高温气冷堆主氦风机与压水堆主泵一样,均为反应堆一回路的关键设备.在反应堆正常工况下,两者具有相同的功能要求,但在事故工况时,因反应堆的运行特性不同,其功能要求各异.目前,对压水堆主泵的惰转特性已有大量的实验研究和实际运行结果,但有关主氦风机惰转特性的实验研究与理论研究还很缺乏.本文结合风机的气动特性与高温气冷堆一回路的阻力特性,从理论上研究高温气冷堆主氦风机的惰转特性,建立主氦风机惰转时的流量与转速的预测公式,并给出其数值预测结果,为高温气冷堆设计的初步安全分析提供依据.     

Impact of Partitioning and Transmutation on High-Level Waste Disposal for the Fast Breeder Reactor Fuel Cycle

The impact of partitioning and/or transmutation (PT) technology on high-level waste management was investigated for the equilibrium state of several potential fast breeder reactor (FBR) fuel cycles. Three different fuel cycle scenarios involving PT technology were analyzed: 1) partitioning process only (separation of some fission products), 2) transmutation process only (separation and transmutation of minor actinides), and 3) both partitioning and transmutation processes. The conventional light water reactor (LWR) fuel cycle without PT technology, on which the current repository design is based, was also included for comparison. We focused on the thermal constraints in a geological repository and determined the necessary predisposal storage quantities and time periods (by defining a storage capacity index) for several predefined emplacement configurations through transient thermal analysis. The relation between this storage capacity index and the required repository emplacement area was obtained. We found that the introduction of the FBR fuel cycle without PT can yield a 35% smaller repository per unit electricity generation than the LWR fuel cycle, although the predisposal storage period is prolonged from 50 years for the LWR fuel cycle to 65 years for the FBR fuel cycle without PT. The introduction of the partitioning-only process does not result in a significant reduction of the repository emplacement area from that for the FBR fuel cycle without PT, but the introduction of the transmutation-only process can reduce the emplacement area by a factor of 5 when the storage period is extended from 65 to 95 years. When a coupled partitioning and transmutation system is introduced, the repository emplacement area can be reduced by up to two orders of magnitude by assuming a predisposal storage of 60 years for glass waste and 295 years for calcined waste containing the Sr and Cs fraction. The storage period of 295 years for the calcined waste does not require a large storage capacity because the number of waste packages produced is significantly reduced by a factor of 5 from that of the glass waste package in the FBR fuel cycle without PT.     

气冷微堆燃料设计的中子学特性影响研究

为分析气冷微堆燃料设计的中子学特性影响,基于方形燃料组件模型,利用蒙特卡罗程序RMC研究了TRISO颗粒、燃料芯块在燃料设计中的主要参量对组件中子学特性的影响。研究结果表明,燃料颗粒体积占比和包覆层厚度不变时,组件寿期随燃料核芯直径的增大先显著增大,而后趋于平稳;燃料颗粒体积占比和燃料核芯直径不变时,组件寿期随包覆层厚度的增大而减小;燃料装载量不变时,芯块直径增大,组件寿期显著增大,而芯块高度影响较小;无燃料区厚度的增加对组件中子学特性有明显的负面影响,基体材料密度、基体杂质硼当量对组件中子学特性的影响较小。研究结果将为后续气冷微堆包覆颗粒弥散燃料的设计优化提供指导。     

高温气冷堆氦气环境中电气设备绝缘设计研究

应用巴申定律研究了氦气的电气击穿特性,并与空气的绝缘特性进行比较。以高温气冷堆氦气透平发电系统电机腔室的设计参数为例,结合氦气的巴申曲线,对氦气条件下气体压力和极间距离的关系进行深入探讨,并提出氦气环境中电气设备绝缘设计需关注的问题。研究结果表明,氦气最小击穿电压为150~200V,绝缘特性较差,电气设备绝缘结构设计应考虑氦气环境压力的影响,现有针对压水堆电站电气设备绝缘结构的验收准则和试验方法并不完全适用于氦气环境。     

Analysis of the SEAFP Reactor Fuel Cycle

The safety aspects of a fusion reactor fuel cycle, which handles substantial quantities of tritium, have been assessed in the framework of the European Programme on Safety and Environmental Assessment of Fusion Power Long Term (SEAL). This study focused on the assessment of the tritium inventory that could be released from interlinked systems in accidental scenarios. A systematic review of the fuel cycle systems was performed by focusing attention on the main interfaces and to the possible propagation of accident sequences through these interfaces. For the bounding accident sequences identified, deterministic analyses were carried out to determine the accident consequences. Both process source terms (PST) and environmental source terms (EST) were estimated. Simultaneous failure of the primary and secondary containment was considered to be beyond the design basis, nevertheless a preliminary analysis has been carried out; a bounding accident sequence related to a double failure, involving a hydrogen fire, has led to a tritium environmental release of 5.3 g and the wall mechanical load deriving from the maximum hypoth-esizable hydrogen detonation has been defined. Tritium releases into the secondary containment are treated by the appropriate detritiation and by the vent detritiation system. The related EST has been estimated based on an overall tritium cleanup efficiency of 99%, deliberately chosen low to cause the EST to be overestimated. The maximum tritium environmental release is less than 11 g and corresponds to an in-vessel LOCA. For accidents initiating in the fuel cycle only, the maximum tritium release is at most 3.1 g.     

高温气冷堆气轮机循环的不同工质特性分析

高温气冷堆气轮机循环发电是今后核能发电的主要方向,在安全性和经济性有很强的竞争优势.本文主要对He、N2和CO2及其混合物的热物性、换热过程的传热系数、压力损失和汽轮机机械所需的级数作了比较分析.结果表明,采用以一定比例混合的He-CO2混合物作为高温气冷堆气轮机循环的工质,既能提高传热系数和减少气轮机机械的级数,又能使得压力损失不会过大.     

轻水堆乏燃料和钍燃料在ACR-700利用的探索

轻水堆乏燃料和钍燃料的利用是解决乏燃料后处理问题和核燃料短缺的有效途径之一。本工作以ACR-700标准燃料为参考,研究了4种不同混合比例的轻水堆乏燃料及钍燃料在ACR-700中的k_∞和燃耗。研究结果表明,将裂变产物分离后,轻水堆乏燃料的重锕系核素在ACR-700中可作为一很好的燃料;只要加入足够的启动燃料,钍燃料也可作为很好的转换燃料,使反应堆内生成²³³U的速率大于易裂变燃料的消耗速率,²³³U的生成对反应堆运行后期维持临界起重要作用。     

高温气冷堆复合联合循环特性研究

高温气冷堆具有900~1 000℃的出口温度,其能量转换系统可采用回热循环和联合循环而具有较高的效率。然而,回热循环对回热器回热度的要求较高,而联合循环的反应堆入口温度较低,不利于循环效率的提高。为进一步提高循环效率,本文提出了一种复合联合循环方案,分析了预冷器和回热器对复合循环效率的影响,并与回热和联合循环进行了比较。结果表明,当反应堆出口温度为950℃、回热器回热度为0.95时,不设置预冷器的复合循环效率可达54.2%。随回热器回热度的增大,不设置预冷器时的复合循环与设置预冷器时的效率差由1.7%逐渐减小。比较回热、联合和复合3种循环方案发现,提高透平总的做功能力较减小压气机功耗更有利于效率的提高。当反应堆出口温度为900~1 200℃时,复合循环的最佳效率较回热循环高3.6%~4.0%,较联合循环高1.9%~2.7%;当反应堆入口温度存在限值且高于500~550℃时,复合循环具有最高的效率。     

商用高温气冷堆联合循环方案研究

本文针对高温气冷堆动力转换单元设计了3种联合循环方案,并将3种循环方案在反应堆出口温度900℃的情况下与闭式Brayton循环进行比较。结果表明:闭式Brayton循环在反应堆出口温度较高时,相应反应堆入口温度也较高,这受到反应堆压力壳材料限制,且所需压气机压比较大;联合循环方案的反应堆入口温度低于370℃,反应堆压力壳可使用SA533钢材,无需内壁冷却,且所需压气机压比较小。方案比较显示,提高联合循环效率需增加下位循环出力。方案3的上位循环是简单Brayton循环,下位循环是再热Rankine循环,循环效率可达50.1%。     

钍基先进CANDU堆燃料循环方式研究

在钍基先进CANDU堆的概念设计中，钍燃料的循环利用方式是一重要问题。文章采用中心两圈为钍燃料、外面两圈为稍加浓缩铀燃料的CANFLEX燃料棒束，通过对燃料棒束栅元物理特性的研究，提出了一套切实可行的直接自身再循环的燃料棒束循环方案。     

高温气冷堆燃料球轨迹CT检测仿真研究与分析

为进一步研究高温气冷堆球流运动,清华大学核能与新能源技术研究院按照1∶5的比例,搭建10 MW高温气冷堆（HTR-10）三维模拟实验台架,采用多层螺旋CT研究燃料球在模拟堆芯中的运动规律,仿真燃料球为外观、质量一致的普通球和目标球。本文对CT球流检测过程进行了仿真,模拟了堆芯中目标球在不同时刻、不同位置的CT重建图像,对图像进行分析,寻找目标球并确定其横向坐标（x,y）和纵向坐标（z）。针对多目标球情况,建立了目标球匹配准则,匹配目标球用不同颜色（c）进行标识,最终完整还原了目标球的五维（x,y,z,t,c）历史运动轨迹信息并动态重现,本文对研究分析高温堆燃料球的行为特征具有重要的参考价值。     

高温气冷堆新型燃料球传感器的设计

对高温气冷堆中燃料球运行情况的准确监测是保障反应堆安全可靠运行的关键。针对原有探测器的不足,利用穿透式涡流检测原理提出了新型对装式燃料球传感器。运用有限元方法建立了该传感器的电磁场数值计算模型,对传感器结构参数和检测参数进行了分析和优化设计。实验结果表明,该传感器过球信号信噪比高,对连续过球具有很好的分辨率,满足反应堆现场使用要求。     

模块式高温气冷堆采用预应力混凝土压力容器的可行性研究

针对模块式高温气冷堆(MHTR)钢制压力容器存在着制造难度大、运输困难和成本高等缺点．开展了对MHTR采用预应力混凝土压力容器(PCPV)的可行性研究本文简要介绍了PCPV的发展现状与技术特点,分析研究了MHTR采用PCPV作为一回路压力边界部件的技术难点及可行性,给出了．MHTR采用PCPV的初步设计方案。对该方案的分析结果表明．将PCPV应用于MHTR在技术上是可行的．不仅能够解决多腔室PCPV的力学问题以及反应堆余热释放等关键技术问题,而且能使MHTR具有更好的安全性和经济性。     

高温气冷堆简化型联合循环特性研究

高温气冷堆具有900~1000℃的出口温度而使其具备采用联合循环发电的潜力。在典型高温气冷堆氦气 蒸汽联合循环中,氦气回路多采用预冷器等设备以降低压气机入口温度和压缩功,但这也使反应堆的入口温度过低而影响联合循环效率的进一步提高。为提高循环效率,对联合循环中的预冷器对循环效率的影响进行了热力学分析,提出了联合循环中是否需设置预冷器的判据,并根据这一判据提出了一种简化型联合循环,即无预冷器的联合循环。简化型联合循环的优化结果表明,在950 ℃的反应堆出口温度下,其循环效率可达52.2%,较典型联合循环的效率提高了1.4%,较回热循环的效率提高了1.7%。与回热循环相比,在反应堆出口温度较高而其入口温度要求较低的工况下,简化型联合循环更具有优势。     

10 MW高温气冷实验堆燃料元件装卸系统研制

根据国际上类似系统的设计和运行经验，彻底改进了10MW高温气冷实验堆(HTR-10)燃料元件装卸系统的单列器、碎球分离器、提升器、控制系统，以使系统变得更为简单、可靠。改进后的设备均在全尺寸实验装置上进行过试验。系统的调试试验和初装料运行表明：该系统的性能满足HTR-10的要求。