秦山三期重水堆核电站技术创新实践

文章介绍了秦山第三核电有限公司消化吸收引进技术,坚持改进创新,提升重水堆核电机组综合性能;坚持自主科技攻关,实施汽轮机功率提升改造及深入开拓重水堆技术优势,大力推进重水堆钻-60生产、回收铀和钍资源利用技术开发。通过不断地科技创新和实施技术改造,核电站的安全性和可靠性也不断提高,核电站整体运营水平处于国际先进水平,取得了良好的经济效益、环境效益和社会效益。     

钍燃料重水堆堆芯特性研究

现有的CANDU重水堆(简称"重水堆")以天然铀作为燃料,但重水堆由于其独特的堆芯设计,具有较好的燃料灵活性,还可以烧低浓铀、回收铀和钍等燃料。研究现有重水堆改烧钍燃料后对堆芯特性和运行安全的潜在影响。使用DRAGON程序建立了重水堆的无限栅元模型,研究比较了钍燃料和天然铀燃料的重要堆芯特性参数。结果表明,尽管2种燃料下的堆芯特性有所差异,但钍燃料利用实际上有助于提升重水堆的运行安全。     

先进重水反应堆综述

重水反应堆是一种重要的堆型。重水堆要占领更大的市场，将面临三个挑战，即降低成本、提高安全性和可持续发展。根据铀富集度的不同和燃料管理战略．燃料运行周期从60天到180天将轻水堆(LWR)乏燃料元件用于重水反应堆，是实现铀资源最佳利用的范例，而且混合氧化物(MOX)燃料也将引入重水反应堆。本文介绍了印度的先进重水堆，该堆率先采用了钍燃料；俄罗斯联邦正在开发高度安全的气冷重水慢化堆；加拿大在基于CANDU6成熟经验的基础上，开发出下一代重水堆Ng CANDU，功率为65MWe。在经济性和固有安全性和操作性能方面均有大的改进。     

印度准备实施下一步钍计划

【澳大利亚铀信息中心《每周新闻简报》2002年4月26日报道】 印度正在准备开始建造一个500 MW的原型快中子增殖堆(PFBR),这将是钍计划实施3个阶段中的第二阶段。PFBR将是池式、钠冷、使用钚-铀氧化物燃料(MOX)并且有一个产生铀-233的钍再生区的反应堆。用于MOX的(反应堆级)钚是在(第一阶段的)商用重水堆产生的。PFBR的设计寿期是40年。项目申情和环境影响评估文件正在等待政府的批准。(该项目的第三阶段,即利用由钍产生的铀-233,还未确定,该计划能够利用印度大量的钍资源。) 在Kalpakkam有一个小的快中子增殖试验堆已经运行了许多…     

压水堆回收铀在重水堆应用示范验证试验

回收铀（RU）是一种重要的核能资源,随着核电发展和铀资源价格的上涨将更加受到重视,迄今为止国际上尚未很好地解决其有效利用问题。鉴于我国既有压水堆又有重水堆的现状,本文提出利用重水堆烧RU的设想,开发了一种与天然铀燃料中子学等效的由RU和贫铀（DU）混合而得的等效天然铀（NUE）燃料,并在秦山运行重水堆上开展随堆示范验证试验,以积累RU利用相关运行经验,为后续全堆应用提供了关键的技术支持。     

重水堆核燃料循环

重水堆(HWR)一个最重要的特点就是中子经济性好，高的中子经济性使得重水堆可以使用天然铀。重水堆除可以用天然铀之外，还可用低富集度铀、轻水堆乏燃料回收的铀、MOX燃料和钍燃料等。这使重水堆的燃料循环具有更大的灵活性。     

印度将进行钍燃料研究

【英国《国际核工程》2001年5月刊报道】 印度原子能委员会(AEC)主席Anil Kakodkar博士表示,印度较小的铀储量已使闭式核燃料循环取得飞速发展。由于预计钍的储量是迄今所知的可利用的铀资源的5～6倍,因此将这种非裂变物质用于大规模的能源生产已成为印度核动力计划的一个重要目标。 印度的钍储量居世界第三。 基于闭式核燃料循环的核动力计划分为3个阶段,涉及使用天然铀做燃料的压重水堆(PHWR)、使用含钚燃料的快增殖堆(FBR),以及使用钍的先进核动力系统。印度正在开发与上面3项中的最后一项内容相关的技术。 (伍浩松 译 哈琳 校)…     

核电站与核反应堆

印度开始了核能计划新阶段 【美国《核子周刊》2002年7月25日刊报道】 印度主管当局已经批准在Kalpakkam开始建造一个500 MW的原型快中子堆。预计该机组在2009年投入商业运行。该堆使用铀钚碳化物燃料(来自其现有的加压重水堆(PHWR)的反应堆用钚),有一个钍再生区,使易裂变的铀-233增殖。这将使印度宏大的钍计划进入第二阶段,并且准备利用印度丰富的钍资源最终实现“钍做燃料的反应堆”(实际上是用铀-233做燃料)。印度在Kalpakkam的小型快中子增殖试验堆自1985年开始运行。 捷克启动新的反应堆 【美国《核燃料》2002年5月27日报道】 捷…     

钍资源及其利用

钍是一种赋存在自然界中的天然放射性元素，在地壳中比铀更丰富，其丰度约为铀的3～4倍。广泛分布在各种不同的地质环境中。世界各国现已查明可经济回收的钍资源量达数百万吨。钍可广泛应用于光学、无线电、航空、航天、冶金、化工、材料等领域，更重要的是它可用作核燃料。随着核电发展对铀需求的不断增加，钍基燃料循环的研发工作业已引起广泛关注，通过大量的研究证实，钍在核能方面的应用具有广阔的前景，未来可有效地补充铀资源的不足。结合钍的物理、化学性质，以及近年世界各国对钍基燃料循环的研发成果，简要介绍世界钍资源的分布、钍资源量、钍资源的地质类型和产出地质背景，以及钍在核能中的应用潜力。     

快堆及其辐照燃料后处理工业的发展现状

快堆既可发电,又可生产核燃料钚-239,它能最有效的利用铀资源,可提高铀利用率50—60倍。快堆的开发已有30多年,目前已运行、计划建造、拆卸停运的快堆达32座。欧州快堆的开发已进入商业化,法国凤凰快堆已发电150亿度,超凤凰快堆(1240MWe)     

CANDU堆应用RU的PWR/CANDU联合核燃料循环的研究

对压水堆乏燃料后处理回收铀（RU)在秦山三期CANDU堆中应用的可行性和经济性进行分析。使用ORIGEN2程序．对后处理回收铀在生产后放置不同时间后核素的成份和放射性活度进行了计算。证明RU燃料元件生产的放射性水平是可以接受的。使用DRAGON／DONJON程序对应用RU的秦山三期CANDU堆的时均堆芯和瞬时堆芯校验分析表明：采用简单的2燃耗区，2、4棒束的换料方案能满足最大通道功率、最大棒束功率限制。通过放射性分析和堆芯物理分析可以看出，秦山三期CANDU堆在不改变堆芯结构及运行模式的条件下，从天然铀(NU)燃料过渡到RU燃料是可行的。通过对秦山三期CANDU堆应用RU的经济性分析，可以看出PWR／CANDU联合核燃料循环的策略既可节约铀资源(23％)，提高燃料的能量输出(4l％)．又减少了废燃料的处置量(66％)．可大大降低核电成本。     

PWR/CANDU联合核燃料循环研究

根据我国已拥有PWR和CANDU核电站的具体情况 ,提出一种PWR/CANDU联合核燃料循环的策略 ,即把压水堆的乏燃料后处理后的回收铀 (RU)用作为CANDU堆的核燃料 ,既可节约铀资源 ,提高燃料的能量输出 ,又减少了废燃料的处置量 ,可大大降低核电成本。由于CANDU堆对核燃料循环的固有灵活性 ,堆芯结构及运行方式不需作重大改变 ,即可完成从天然铀到RU的过渡。又由于RU较低的放射性活度 ,这对CANDU堆的燃料制造是可以接受的 ,因而只需对现有燃料制造生产线稍加屏蔽措施 ,对运输和运行中燃料管理操作等都勿须改变。因而这一策略是具有重大经济效益和吸引力的     

Utilization of spent PWR fuel-advanced nuclear fuel cycle of PWR／CANDU synergism

High neutron economy, on line refueling and channel design result in the unsurpassed fuel cycle flexi-bility and variety for CANDU reactors. According to the Chinese national conditions that China has both PWR and CANDU reactors and the closed cycle policy of reprocessing the spent PWR fuel is adopted, one of the advanced nu-clear fuel cycles of PWR/CANDU synergism using the reprocessed uranium of spent PWR fuel in CANDU reactor is proposed, which will save the uranium resource (-22.5%), increase the energy output (-41%), decrease the quantity of spent fuels to be disposed (-2/3) and lower the cost of nuclear poower, Because of the inherent flexibility of nuclearfuel cycle in CANDU reactor, and the low radiation level of recycled uranium(RU), which is acceptable for CANDU reactor fuel fabrication, the transition from the natural uranium to the RU can be completed without major modifica-tion of the reactor core structure and operation mode.It can be implemented in Qinshan Phase Ⅲ CANDU reactors with little or no requirement of big investment in new design. It can be expected that the reuse of recycled uranium of spent PWR fuel in CANDU reactor is a feasible and desirable strategy in China.     

Computational analysis of neutronic effects of ThO2 rods loaded in CANDU 6 fuel assemblies

Thorium as a suitable fertile with higher natural resources in comparison with uranium resources has been remarkably considered by different nuclear energy user countries in the last decades. Its prominent features such as suitable possibility for power flattening of a nuclear reactor, applicable breeder blanket to produce~(233)U fissile as well as neutron leakage prevention from a nuclear core has caused its application as power flatter, breeder material or other aimed utilizations be evaluated by the researches. In the present study, neutronics of a modeled CANDU 6loaded with Th O_2 and UO_2fuel rods have been computationally studied. The study aimed at reprocessing of burned Th O_2 seeds at CANDU 6 reactor to recover the total produced uranium, which is to be going under another compound fuel cycle. The obtained results showed all the core reactivity coefficients are sufficiently negative. The modeled core 949 GWd burn-up concluding in 99.99 %depletion of~(235)U initial loads. 18.38 kg of~(233) U was produced in the burnt Th O_2 fuel after 1-year burn-up time. In addition, 31.84 kg of~(239) Pu was produced in the UO_2 spent fuel rods after the burn-up time. After a proposed cooling time, about 50.01 kg of~(233)U will be available in the spent Th O_2 fuel.     

等效天然铀入堆辐照试验中子学性能分析

本文基于重水堆堆芯核设计程序系统,计算分析了装入等效天然铀（NUE）燃料的试验堆芯的中子学性能。对选择两个燃料通道进行入堆试验的方案进行了论证分析,通过与相应的设计限值以及全天然铀（NU）燃料堆芯中子学性能的比较,检验了实际入堆NUE燃料的中子学等效性。研究表明,实际入堆NUE燃料满足燃料的等效性要求,两个NUE燃料通道入堆试验方案从核设计角度是可行的,堆芯安全性不受影响。     

Experimental and analytical dynamic analysis of coolant channel and its internals of Indian 540 MWe PHWR reactor

The horizontal coolant channel is one of the important parts of primary heat transport system in PHWR type of reactors. There are in all 392 channels in the core of Indian 540 MWe reactor. Each channel houses 13 natural uranium fuel bundles and shielding and sealing plugs one each on either side of the channel. The heavy water coolant flows through the coolant channel and carry the nuclear heat to outside the core for steam generation and power production in the turbo-generator. India has commissioned one 540 MWe PHWR reactor in September 2005 and another similar unit will be in operation very shortly. For a complete dynamic study of the channel and its internals under the influence of high coolant flow, experimental and analytical study has been carried out. The operating life of a typical coolant channel typically range from 10 to 15 full power years. Towards the end of its operating life, its health monitoring becomes an important activity. Vibration diagnosis plays an important role as a tool for life management of coolant. A good correlation has been achieved between the results of experimental and analytical models. Through the study dynamic characteristics of the coolant channel under simulated loading condition, an attempt has been made to develop a diagnostics to monitor the health of the coolant channel over its operating life. Experimental study has been also carried out to characterize PHWR fuel vibration under different flow conditions. Such results are published for the first time.     

Optimization of depleted uranium bundle loading in fresh core of Indian PHWR by evolutionary algorithm

This paper is concerned with the Indian design of a 220 MWe pressurized heavy water reactor (PHWR) having natural uranium (NU) fuel and heavy water as moderator and coolant. At the beginning of life, it is necessary to flatten the power by loading some depleted uranium (DU) bundles to achieve a nearly full power operation. The determination of best possible locations of DU bundles, which maximize fuel economy as well as safety, is a large-sized combinatorial optimization problem with constraints. In the past, 384 DU bundles have been loaded in locations determined by manual intuition in an Indian PHWR and maximum permissible power of 93% full power (FP) was obtained. In the present paper, a modern evolutionary algorithm called estimation of distribution algorithm (EDA) is used to improve upon this distribution. Optimum distributions of DU bundles which maximize K_eff and give 100% FP without violating safety parameters such as maximum permissible bundle power, channel power, channel outlet temperature and permitted reactivity worths of shut-down systems are obtained. Another aspect studied in this paper is to find out how far one can increase the number of DU bundles loaded in the core. This will minimize the NU bundles requirement, extract more power from DU bundles and thus provide better fuel utilization. The idea is to conserve NU bundles. The optimum distribution of DU bundles has been obtained for the total number of DU bundles ranging from a few hundreds to a few thousands. It is found that, depending on various conditions, about 60–80% of the core can be loaded with DU bundles leading to a substantial saving in NU bundles. Some variation in the implementation of EDA to generate loading pattern of PHWR reactor core is also studied.     

Safety characteristics of the multipurpose fast reactor (MPFR)

The safety characteristics of a long-life multipurpose nuclear reactor (MPFR) with self-sustained liquid metallic fuel and lead coolant, which is proposed to meet the requirements for the energy production in the future, were investigated. The application of liquid plutonium–uranium metallic alloys used as a nuclear fuel demonstrated high potential to reach excellent reactor shutdown characteristics against anticipated transients without scram such as unprotected loss-of-flow and unprotected transient overpower. The calculations indicated that the thermal expansion of liquid fuels would cause the negative reactivity insertion that would be larger in magnitude than any other thermally induced reactivity changes. This created the reactivity balance for the passive shutdown and power stabilization capabilities of the MPFR core. It was found that MPFR satisfies such design characteristics to be a potential candidate providing the replacement of fossil fuels by alternative energy sources in the next century.     

快堆和我国核能的可持续发展

我国核能发展战略的第二步——快堆，因其易裂变燃料在堆中可增殖和可嬗变高放长寿命核素的特陛，实现热堆-快堆匹配闭式核燃料循环可保证核能的可持续发展。作为我国快堆工程技术发展的起步，65MW热功率中国实验陕堆已处调试阶段。在当前压水堆发展计划的基础上，加快陕堆及其相关闭式核燃料循环的发展以实现如下三个战略目标：（1）2030年前批量建成示范决堆，增加核电容量；（2）2050年核电容量发展到240GW，约占国家总电力生产的16%；（3）2050—2100年实现核能大规模替代化石能源，大大减少CO2的排放。