2022年国外新型核电反应堆发展动向

<正>在能源危机凸显以及气候问题严峻的背景下，核电成为各国发展安全独立能源的焦点。2022年，小型模块化反应堆、先进压水堆、热管型反应堆以及第四代反应堆等新型核电反应堆技术发展迅速，应用进展显著。多个小型轻水堆和先进大型压水堆项目发展活跃，钠冷快堆、铅冷快堆等第四代反应堆的研究和建设也有多项突破。     

第四代核电 高温气冷堆

正2021年《政府工作报告》提出"大力发展新能源,在确保安全的前提下积极有序发展核电"。面对碳达峰、碳中和的目标,核电以其巨大的能量以及清洁高效的特征备受瞩目,步入优化产业结构和能源结构的主战场。由中国华能集团有限公司牵头组织实施的国家科技重大专项高温气冷堆示范工程,是全球首座球床模块式高温气冷堆项目,更是我国具有自主知识产权、具备第四代核能系统安全特性的核电机组。项目的建成对于实现全球第四代核电技术的"中国引领"具有重要的意义,也将为核电技术在产业结构和能源结构调整中发挥更大的作用奠定基础。     

我国近期核电发展可选择的先进堆型

核电经过四十多年的发展，形成特点不同的四代核电堆型。目前全球正在运行的400多台核电机组的主力堆型基本都属于第二代核反应堆，而更加安全和先进的第三代先进反应堆目前还未被厂泛应用，第四代核电堆型还处于概念设计阶段。本文重点介绍丁几种主要的第三代先进核电堆型，对其技术特点、安全性和经济性进行了分析比较，为我国的核电发展提供参考。     

高温气冷堆技术的研究及发展

自1954年前苏联第一座SMW试验性核电站投运以来,核电在一些国家的电力工业中保持着重要作用。从世界核电下一阶段发展来看,重点仍是提高安全性和降低造价,主要发展的是先进的水堆技术和其他先进的反应堆技术,可以预测,高温气冷堆技术作为一种先进反应堆技术在未来的10～15年必将取得长足的发展。 高温气冷堆技术的发展和现状 气冷堆是国际上反应堆发展中最早的一种堆型,这种反应堆初期被用来生产军用钚,20世纪50年代中期以后发展成为商用核电站的堆型之一。气冷堆的发展大致可以分为四个阶段:即早期气冷堆(Magnox)、改进型气冷堆(AGR)、…     

多模块高温气冷堆地震概率安全分析关键技术研究

高温气冷堆核电厂采取多个反应堆模块匹配1个汽轮机的设计方式,即1台高温气冷堆机组会包含多个反应堆模块,这使多个高温气冷堆模块在地震外部事件下存在明显的相关性,因此在利用概率风险分析方法来全面地识别和评价高温气冷堆的地震风险时,需要从机组的角度充分考虑和模化机组内多个反应堆模块间的相关性。高温气冷堆示范电站已完成了较为完整的单模块地震概率安全分析,本文将以该分析结果为基础梳理出高温气冷堆多模块地震概率安全分析的关键技术要素并进行研究,研究内容包括多模块事件序列建模和地震相关性失效评价等关键技术,并针对多模块高温气冷堆提出了应用策略。然后以双模块设计的高温气冷堆示范电站为对象,以地震导致丧失厂外电始发事件为代表,对多模块高温气冷堆地震概率安全分析进行了实例分析获得远低于概率安全目标的释放类频率,且分析得到了高温气冷堆多模块事件序列建模策略与地震相关性失效的评价路线可行这一重要结论。     

世界核电发展趋势与高温气冷堆

核能的发展面临经济竞争力、核安全、核废物的最终处置及防止核武器材料扩散的挑战。为改善公众的可接受性 ,核电厂的安全性进一步改进。电力市场体制的非管制化改革加剧了电力技术的竞争。环境保护意识增强使核废物的处置倍受关注。 80年代中期以来发展的先进轻水堆核电厂如ABWR ,System 80 ,EPR ,AP60 0等是今后一段时期内商用核电的主力堆型。进入 2 0 0 0年之际 ,美国能源部正在规划发展第四代先进核能系统 ,目标是在 2 0 2 0年或之前 ,向市场提供经过验证的成熟的第四代核电厂技术 ,以替代美国退役的核电容量。球床高温气冷堆被认为是第四代先进核能系统的优选技术。南非ESKOM电力公司选择了球床高温气冷堆作为今后核电发展的堆型。清华大学承担设计和建设的 10MW高温气冷实验堆计划在 2 0 0 0年内临界。通过10MW高温气冷堆的建造 ,我国已形成了高温气冷堆技术的自主知识产权 ,初步具备了自主设计、制造和建造的能力     

法国第四代反应堆原型堆将在2020年之前建成

【法新社巴黎2006年1月5日电】法国总统希拉克2006年1月5日宣布,法国将在2020年之前建成一座第四代核反应堆的原型堆,并在未来几十年内减少对石油的依赖。希拉克说,作为世界上仅次于美国的第二大核电生产国,法国应当“在核能领域保持其领先地位。”在当天对法国工商界人士的新年讲话中,希拉克宣布“已经决定立即启动由法国原子能委员会(CEA)主导的第四代核反应堆原型堆的研制工作,并计划于2020年将原型堆投入运营”。希拉克表示,法国将与希望参加该项目的工业界伙伴或国际伙伴共同努力,开发出一种更安全、更清洁且造价更低的反应堆,以满足…     

PSA风险重要度分析在高温堆调试监督中的应用探索

高温堆因具有良好的安全特性、较强的经济竞争能力、广阔的应用前景成为第四代先进核能系统的优选技术。为做好高温堆调试监督工作,华东监督站应用概率安全分析结果开展了高温堆调试监督研究工作,分析了高温堆调试试验项目,选取安全重要的调试试验和见证点,形成了风险指引型高温堆调试监督项目清单,为合理配置调试监督资源,提高监督效率和效能提供了参考。     

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。     

谈世界核电站发展趋势

本文简述了世界核电发展趋势情况，重点描述了AP1000、模块高温气冷堆（MHTGR）和国际合作的先进安全反应堆（IRIS）等核电站的主要性能和研究情况。     

核能制氢技术的发展

氢是清洁能源,有非常好的应用前景.但氢是二次能源,需要利用一次能源来生产.以可持续的方式(原料来源丰富、无温室气体排放)实现氢的大规模生产是实现氢广泛利用的前提.核能是清洁的一次能源,核电已经成为世界电力生产的主要方式之一.正在研发的第四代核能系统除了要使核电生产更经济和更安全之外,还要为实现核能在发电之外的领域的应用...     

我国高温气冷堆的发展

模块化高温气冷堆具有的固有安全特性、建造周期短和相处容量小等优势正好符合电力系统非管制化（Ｄｅｒｅｇｕｌａｔｉｏｎ）发展趋势对于发电厂的要求,清华大学核能设计研究院正在建造一座１０ＭＷ高温气冷实验堆。本文着重分析了高温气冷堆的安全特性和提高发电效率的氦循环方式。     

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.     

Large eddy simulation in pebble bed gas cooled core reactors

A high temperature gas-cooled reactor (HTGR) is one of the renewed reactor designs to play a role in nuclear power generation. This reactor design concept is currently under consideration and development worldwide. The combination of coated particle fuel, inert helium gas as coolant and graphite moderated reactor makes possible to operate at high temperature yielding a high efficiency. In this study the simulation of turbulent transport for the gas through the gaps of the spherical fuel elements (fuel pebbles) was performed using the large eddy simulation. This would help in understanding the highly three-dimensional, complex flow phenomena caused by flow curvature in the pebble bed. Resolving all the scales of a turbulent flow is too costly, while employing highly empirical turbulence models to complex problems could give inaccurate simulation results. The large eddy simulation (LES) method would overcome these shortcomings. An attempt to obtain experimental velocity flow patterns using particle image velocimetry technique combined with matched refractive index liquid was pursued.     

ANTARES: The HTR/VHTR project at Framatome ANP

Framatome ANP is developing a very high temperature reactor (VHTR), relying on its previous experience with high temperature reactor concepts, from its participation in the MODUL and the GT-MHR designs. While being a major actor in the nuclear reactor business with proven light water technology, AREVA wishes to be ready to meet the new challenges calling for small grid requirements, high temperature process heat and cogeneration. The Framatome ANP VHTR design for electricity production is based on an indirect cycle coupled to an “off-the-shelf” combined cycle gas turbine. Although direct cycle HTRs are being promoted for their high efficiency, preliminary evaluations show that the Framatome ANP design efficiency is on par with a direct cycle while avoiding power generation system (PGS) developments and keeping the PGS contamination free. Moreover, the nuclear heat source of the indirect cycle could also be used to meet the heat supplies from a standard design for multiple applications.     

Future HTGR developments in China after the criticality of the HTR-10

Nuclear power has a great potential to develop in China because of China's fast economic increase. HTGR will be the most promising nuclear reactor to apply in the future Chinese market. After the initial criticality of the HTR-10, subsequent research and validation of the HTGR performance is by hot commissioning tests and power operation, safety demonstration experiments, R&D of gas turbine and process heat application technologies, and promotion of industrial application of HTGR technologies. The commercial prototype HTR-PM is under study and conceptual design has started. These activities will result in the safe and economic development of HTGR technologies in China.     

高温和超高温气冷堆动力转换方案研究

动力转换单元是高温和超高温气冷堆的重要组成部分。本文对高温和超高温气冷堆的动力转换单元进行研究。从4个关键参数（反应堆出口温度、反应堆入口温度、压缩比和主蒸汽参数）入手,对5个循环方案进行比较分析。综合考虑各种工程因素,上位循环为简单氦气透平循环、下位循环为有再热的蒸汽轮机循环的联合循环方案是具有竞争力的,其中下位循环在高温气冷堆范围是亚临界参数循环,在超高温气冷堆范围是超临界参数循环。联合循环可实现高温和超高温气冷堆热量的高效率转化,且反应堆入口温度在反应堆压力壳材料允许的范围内,具有足够的安全性。     

Small PWRs using coated particle fuel for district heating, PFPWR50

It has been said that nuclear energy is an important option for especially developing countries to satisfy their increasing energy demand. However, it will be difficult to deploy first of a kind nuclear power plant in developing countries because extensive safety demonstration has to be conducted in industrialized countries. On the other hand, it will be essential to present rigid proof of reliable operational experience to develop proper understanding of the safety features of new reactor systems among the people around the demonstration plant sites. One of the ways to solve the issue is to integrate existing technologies supported by a great deal of data and experience into a new reactor design. Based on the consideration, a small-sized district heating reactor system based on the pressurized water reactor (PWR) technologies combined with the fuel concept of high temperature gas cooled reactors (HTGRs) has been studied. The purpose of the combination of these two existing concepts is to take the best advantages of both excellent operational experience of PWRs and the integrity of HTGR fuel, coated particle fuel, against fission products release even at high temperature. We expect that this approach will help create a breakthrough to the current stagnation of nuclear power deployment.     

The high temperature gas-cooled reactor as a source of high temperature process heat

The nuclear reactor has established itself as a future major supplier of electrical energy. The industrial market for other forms of energy, however, is almost as large and represents a new potential for the use of nuclear reactors. The high temperature gas-cooled reactor (HTGR) has been developed for commercial application in the electric power generation field. Since the HTGR is capable of delivering process heat in the temperature range of 1000–1500°F, it has many applications that would not be possible at the lower operating temperatures of water-cooled reactors. This paper briefly summarizes the development of the HTGR and outlines its salient technical features. Modifications to the reactor that enable it to be used as a process heat source are discussed. Specific applications are developed for coal gasification, steelmaking, and hydrogen production.