三种堆型核电厂经济性评价

本文分析了核电投资的特点、建立了考虑价格浮动和通货膨胀等因素影响的核电厂建成价和核燃料成本的计算模型。对压水堆、高温气冷堆和快堆三种堆型的经济性进行了研究。结果表明，当高温气冷堆和快堆两种先进堆型实现商用概念设计后，其商业竞争能力可与现有的压水堆相媲美。     

10MW高温气冷实验堆的堆体结构特点

模块式高温气冷堆是当今世界上公认的先进反应堆堆型之一。固有安全性是它的最突出的优点。本文对10MW高温气冷堆的堆体布置进行了详细描述,并对10MW高温气冷堆的结构设计特点进行了分析。根据10MW高温气冷堆的特点,本文对该堆的固有安全性、制造工艺等方面的优点进行了论述。     

美国高温气冷堆及其近期发展计划

简要地说明了美国高温气冷堆的发展过程,评价了美国已建成的高温气冷堆,介绍了美国目前正在实施的发展模块式高温气冷堆的三项计划。     

模块式高温气冷堆核电站设计特点

模块式高温气冷堆具有安全、灵活、可靠、经济性好的优点,受到核技术先进国家的重视。本文着重介绍了美国新近推出的模块式高温气冷堆核电站的设计特点和安全特性。     

球床式高温气冷堆MOX燃料循环优化

与压水堆相比,球床式高温气冷堆能在堆芯结构不做明显改变的情况下采用全堆芯装载混合氧化物（MOX）燃料元件。基于250 MW球床模块式高温气冷堆堆芯结构,设计了4种球床式高温气冷堆下MOX燃料循环方式,包括铀钚混合的燃料球和独立的钚球与铀球混合装载的等效方式,采用高温气冷堆设计程序VSOP进行分析,比较了初装堆的有效增殖因数、燃料元件在堆芯内滞留时间、卸料燃耗、温度系数等主要物理特性。结果表明：采用纯铀和纯钚两种分离燃料球且铀燃料球循环时间更长的方案,平均卸料燃耗较高,总体性能较其他循环方式优越。     

高温气冷堆和轻水堆超铀元素管理特性研究

乏燃料中长寿命锕系元素对环境造成长期潜在危害,本文研究球床高温气冷堆不同燃料循环中超铀元素的产生和焚烧特性。在250 MW球床模块式高温气冷堆示范电站HTR-PM铀钚循环的乏燃料中提取铀和钚作为核燃料,设计了PuO₂和MOX燃料元件,将新设计的燃料元件重新装入与HTR-PM相同结构和尺寸的堆芯,分别形成纯钚燃料循环和MOX燃料循环。采用高温气冷堆物理设计程序VSOP,研究了高温气冷堆一次通过燃料循环和不同闭式燃料循环的超铀元素焚烧特性,并与轻水堆燃料循环结果进行比较和分析。结果表明：高温气冷堆一次通过燃料循环超铀元素生成率约为轻水堆的1/2;高温气冷堆闭式燃料循环能有效嬗变超铀元素。     

HTR—10石墨球与燃料球均匀混合装料初装堆方案研究

分析了球床式高温气冷堆几种可能的初装堆方案的特点，选取石墨球与燃料球均匀混合作为１０ＭＷ高温气冷实验堆的初装堆方案。利用高温气冷堆物理设计程序ＶＳＯＰ进行计算，分析屯ＨＴＲ－１０从初始装料向平衡态过渡过程中的倒换料方式，最大单球功率及最大燃耗变化情况。     

高温气冷堆核电技术产业化思考

根据国外高温气冷堆核电技术发展历程和我国高温气冷堆核电技术产业化的成就、现状与发展方向,从有利于发挥高温气冷堆核电技术多用途特性和加快高温气冷堆核电技术"走出去"步伐的战略角度考虑,本文提出了我国高温气冷堆核电技术产业化的相关建议,供政府有关部门及行业相关单位参考。     

高温气冷堆核电站在我国的商业化前景

分析了发展高温气冷堆核电站的必要性,介绍了高温气冷堆发电技术的特点、现状和我国高温气冷堆核电站示范工程的建设情况,提出了充分发挥高温气冷堆发电技术特点、进一步开发高温气冷堆核电站的设想,认为条件成熟情况下,高温气冷堆核电站将有很大的商业推广价值。     

美国能源部选用固有安全动力堆和高温气冷堆

【英国《国际核工程》1988年9月号第2页报道】通用原子公司(GA)的模块式高温气冷堆(HTGR)和通用电气公司(GE)的固有安全反应堆(PRISM)在由美国能源部赞助的竞赛设计中分别获胜。美国能源部选择HTGR作为它两座军用生产堆的一种,该堆计划将于90年代晚些时候投入运行。两堆战略需要一座建在南卡罗来纳的萨凡纳河工厂的常规重水堆和一座建在爱达荷国家工程实验室的HTGR堆。     

高温气冷堆堆芯流场高效全局求解方法研究

高温气冷堆堆芯流场计算是高温气冷堆系统分析模拟的重要部分。原堆芯流场计算程序在采用分块迭代方法计算耦合流场时,部分工况收敛速度慢、耗时长,甚至出现不收敛问题。基于此,本文分析了分块迭代方法的收敛性,从全局求解的思路出发,并根据堆芯流场模型的特点使用直接求解法,开发了一套用于计算高温气冷堆堆芯流场的高效全局求解方法。与分块迭代方法相比,全局求解方法的收敛性和计算效率均得到很大提高。     

Small high temperature gas-cooled reactors with innovative nuclear burning

Since the innovative concept of CANDLE (Constant Axial shape of Neutron Flux, nuclide densities and power shape During Life of Energy producing reactor) burning strategy was proposed, intensive research works have been continuously conducted to evaluate the feasibility and the performance of the burning strategy on both fast and thermal reactors. We learned that one potential application of the burning strategy for thermal reactors is for the High Temperature Gas-Cooled Reactors (HTGR) with prismatic/block-type fuel elements. Several characteristics of CANDLE burning strategy such as constant reactor characteristics during burn-up, no need for burn-up reactivity control mechanism, proportionality of core height with core lifetime, sub-criticality of fresh fuel elements, etc. enable us to design small sized HTGR with a high degree of safety, easiness of operation and maintenance, and long core lifetime which are required for introducing the reactors into remote areas or developing countries with limited infrastructures and resources. In the present work, we report our evaluation results on small sized block-type HTGR designs with CANDLE burning strategy and compared with other existing small HTGR designs including the ones with pebble fuel elements, under both uranium and thorium fuel cycles.     

Crossflow characteristics of HTGR fuel blocks

Leakage crossflow characteristics in an HTGR core have been studied experimentally and numerically. Two-block crossflow experiments were carried out and the crossflow rate and the pressure difference were measured for various interface gap configurations. A numerical model has been proposed to predict crossflow rates, and the numerical results using the finite element method agreed well with experimental ones. In addition, empirical crossflow equations, which apply to various fuel blocks, were derived for the analysis of the flow distribution in an HTGR core.     

Nonlinear dynamic analysis of prismatic elements for high-temperature gas-cooled reactor cores

The high-temperature gas-cooled reactor (HTGR) core consists of several thousand prismatic graphite fuel elements arranged in columns within a prestressed concrete vessel. A major research and development effort was initiated in 1970 at General Atomic Company to study the dynamic response of the HTGR core arrangement to seismic excitation.This paper presents a discussion of the history and some of the results of this effort, with respect to advances made in the development of analytical methods. The computer programs developed to perform the analysis are described, along with certain techniques and the modeling required to utilize them. The purpose is to describe the nonlinear dynamic analysis techniques employed to analyze the HTGR core. Correlation of the codes is beyond the scope of the paper and will be discussed in subsequent publications.Each fuel column in the HTGR core is composed of stacked elements doweled together to ensure alignment of the coolant channels. Gaps exist between columns, allowing the elements to impact during a seismic disturbance. Analysis of this type of structure by standard structural dynamics techniques is not possible since both nonlinearities and discontinuities exist. One- and two-dimensional models of the three-dimensional core have been developed with explicit time integration methods. Various methods to treat the impact between elements are discussed.Three computer codes were developed. CRUNCH-1D models a one-dimensional horizontal strip through the core; CRUNCH-2D, a two-dimensional horizontal planar section; and MCOCO, a two-dimensional vertical planar section. The dynamic characteristics of these three representations of the full core structure are compared and the methods evaluated in the text. Plans for additional development and work to improve the techniques are also discussed.     

HTGR plant safety design bases

The high temperature gas-cooled reactor (HTGR) has inherent and design safety features that are sifnificant and unique, requiring a number of safety criteria and approaches that differ markedly from other reactor types. This paper briefly reviews the design of HTGR plants that have been built and are being offered in the United States. It then reviews the safety considerations involved in the design of the plants being offered. The unique features, their development, and their effects on safety criteria are described. The design bases of the prestressed concrete reactor vessel (PCRV) are given particular attention. Operating characteristics of the HTGR and plant response to transient conditions are discussed. The design-basis depressurization accident evolution and related HTGR safety requirements are discussed. Characteristics of the HTGR with respect to technical specifications are discussed, with particular emphasis on the PCRV and the core safety limit.     

HTGR plant dynamics analysis under accident conditions

Important features of high temperature gas-cooled reactor (HTGR) systems related to plant dynamics and accident analysis are discussed. Because of the basic simplicity of the HTGR system, it is possible to analyze the full reactor plant (core, helium circulators, steam generators and reheaters, feedwater controls, turbine controls, and plant protective action) in a single computer code. Representative dynamics analysis is presented for the Fort St. Vrain Power Station.     

A seismic analysis method for a block column gas-cooled reactor core

An analytical method for predicting the behavior of a prismatic high-temperature gas-cooled reactor (HTGR) core under seismic excitation has been developed. In this analytical method, blocks are treated as rigid bodies, are constrained by dowel pins which restrict relative horizontal movement but allow vertical and rocking motions. Coulomb friction between blocks and between dowel holes and pins is also considered. A spring dashpot model is used for the collision process between adjacent blocks and between blocks and boundary walls.Analytical results are compared with experimental results and are found to be in good agreement. The analytical method can be used to predict the behavior of the HTGR core under seismic excitation.     

The HTGR gas turbine plant with dry air cooling

The development of the HTGR gas turbine power plant as a future evolution of the HTGR is one of the most promising solutions to the interrelated power generation and environmental problems. The HTGR gas turbine can make dry air cooling economical and can make possible increased flexibility and economy in power plant siting. The simplification and size reduction of the overall plant imply lower capital costs. Cycle parameters and plant layout for a typical HTGR direct-cycle gas turbine plant of 1100 MW(e) output are described. For safety reasons all the primary equipment is integrated inside the prestressed concrete reactor vessel. Four parallel loops are contained in eight vertical PCRV cavities located around the core cavity. Alternative design configurations and parameter choices are discussed. The advantages and the development potential of the direct cycle with regard to heat rejection and cost are discussed. The possibility of profitably using the gas turbine thermal discharge for operating a seawater distillation plant is pointed out.     

安全经济高效的先进能源

具有第四代安全经济特性的核电应该是人们期待的先进的清洁低碳能源。高温气冷堆是当今研发的第四代核电堆型之一,但现有的设计还存在需要排除的严重的安全隐患。堆芯不熔化,不等于说不会有严重事故发生。需要吸取国外球床高温堆和柱状高温堆两种实验堆型运行的经验教训、扩展安全观念和应对安全低概率事件,确保反应堆不出现后果极其严重的放射性释放事故。当热电转换系统采用与燃气蒸汽联合循环耦合应用的技术以后,会发挥高温堆所长,更大地提升转换效率,形成一种高安全低投资和高效率的双燃料清洁能源,可用于大堆或小堆的应用环境,可满足电力系统基本负荷和调锋负荷的需要。在工程设计上采取一系列改进和创新措施,包括釆用规则床模块化及地下反应堆设计以后,可在提高反应堆核心部位安全防卫能力的同时,防范低概率事件,成为一种新的安全经济高效的先进能源。