临界热流密度实验研究进展

临界热流密度是重要的限制性热工参数，它的大小直接影响反应堆的安全性和经济性。本文介绍了我国核电站临界热流密度实验研究进展，比较了国外核电站临界热流密度研究发展状况，文章认为，我国临界热流密度实验研究的发展方向是进行全长棒束非均匀加热临界热流密度的实验研究。     

WWER-1000燃料组件特点及棒弯曲分析

本文根据WWER-1000反应堆的设计特点及其运行实践,阐述了WWER-1000燃料组件的设计特点,并与西方压水堆燃料组件进行了相应的比较.重点分析论述了WWER-1000反应堆燃料棒弯曲的特点,以及在热工水力和燃料组件设计中是如何考虑棒弯曲效应的,进行了燃料棒弯曲对临界热流密度影响实验的研究.结果表明:WWER-1000燃料组件在整个运行寿期内的性能是可以保证的.     

大型先进压水堆核电站堆芯装载方案设想

核电站堆芯装载方案是反应堆堆芯设计的重要基础,它首先必须满足核安全的要求,同时还要尽可能地提高经济性。通过分析国内、外百万千瓦级核电站的堆芯装载,对反应堆输出功率、燃料组件数、堆芯平均线功率密度进行比较,给出我国大型先进压水堆核电站示范工程反应堆堆芯装载方案的设想,为技术决策提供参考。     

燃料组件导向管冷壁效应对CHF影响试验研究

燃料组件临界热流密度（CHF）性能是压水堆堆芯热工水力设计和安全分析的基础,对反应堆的安全运行至关重要。本文采用非均匀加热典型栅元和导向管栅元两组CHF试验数据开发了具有针对性的CHF关系式并对比研究了导向管冷壁对CHF的影响,获得了导向管冷壁效应因子。研究结果表明,针对轴向功率非均匀加热,在边界条件一致的情况下导向管的存在不会降低棒束的平均功率,但会导致烧毁点的位置发生变化并使得CHF降低,导向管冷壁效应因子约为8%。     

Progress in the studies of the advanced plutonium fuel assembly

A new assembly concept, designated APA (for dvanced lutonium fuel ssembly), should make it possible to multi-recycle plutonium in pressurized water reactors. The basic idea is founded on the manufacture of a large plutonium thin annular fuel rod with an inert support, cooled on both faces. The absence of plutonium generation, combined with moderate fuel temperature should make it possible to achieve substantial burn-up fractions in these rods. The assembly is compatible with the internals of a Pressurized Water Reactor (PWR), and provides for permanent reversibility. Neutronic studies showed a compliance with actual safety/control criteria. A multi-recycling scenario was simulated for 84 years' operation with a 65 GW electrical power installed capacity, comprising forty-five 1450 MW electrical power PWRs, 32 of which are loaded with UO₂ and 13 with APAs. It showed that the plutonium inventory is controlled. Thermal-hydraulic studies showed one can find an annular rod geometry allowing one to respect both margins to Critical Heat Flux (CHF) during normal and accidental operations and void fraction limitations.     

压水堆燃料组件抗震试验研究

燃料组件属I类抗震物项,其抗震问题直接关系核电厂运行安全,通常需通过抗震试验验证反应堆燃料组件抗震分析方法的合理性。本文模拟反应堆实际堆芯燃料组件安装方式,设计压水堆燃料组件抗震试验件与试验装置,针对不同组件数量布置方案,在高性能地震模拟振动台上开展试验研究。结果表明,水介质中燃料组件的第一阶频率为2.96 Hz,最大冲击力出现在燃料组件偏中间位置处,试验获取了地震作用下燃料组件的格架冲击力、格架相对位移、模拟堆芯板与围板的加速度等响应。试验结果可用于设计基准事故工况中燃料组件抗震分析模型的建立与分析软件的验证。     

混合堆增殖乏燃料组件中子学特性初步研究

本文主要对聚变-裂变混合堆增殖乏燃料在压水堆组件中使用的可能性进行了初步研究。根据聚变 裂变混合堆增殖乏燃料的特点,给出了的聚变-裂变混合堆增殖乏燃料压水堆组件设计方案,分析组件的燃料温度系数、慢化剂温度系数等参数。结果表明：聚变 裂变混合堆乏燃料组件的特性与全铀组件的特性相似。在相同的易裂变同位素质量百分比情况下,本文给出的组件设计方案的功率不均匀系数更小。研究结果可为未来实现聚变 裂变混合堆和压水堆联合循环系统提供技术支持。     

Post DNB Heat Transfer Experiments for PWR Fuel Assemblies

Nuclear Power Engineering Corporation (NUPEC) and Mitsubishi performed heat transfer experiments on post DNB (departure from nucleate boiling) for the pressurized water reactor (PWR) fuel assemblies under the sponsorship of the Japanese Ministry of Economy, Trade and Industry (METI) as one of a series of fuel assembly verification tests. Based on the obtained experimental data, a new evaluation model for the fuel rod heat transfer behavior after DNB was developed. A large safety margin, which had remained in the present thermal-hydraulic design that did not allow DNB, was confirmed by applying the developed model to the PWR plant safety analysis.     

数值反应堆堆芯通道级三维热工水力程序CorTAF开发及初步验证

堆芯是核动力系统的核心部件,其完整性是反应堆安全运行的重要前提。传统核反应堆堆芯热工水力分析方法无法满足未来先进核动力系统的高精度模拟需求。本文依托开源CFD平台OpenFOAM,针对压水堆堆芯棒束结构特点建立了冷却剂流动换热模型、燃料棒导热模型和耦合换热模型,开发了一套基于有限体积法的压水堆全堆芯通道级热工水力特性分析程序CorTAF。选取GE3×3、Weiss和PNL2×6燃料组件流动换热实验开展模型验证,计算结果与实验数据基本符合,表明该程序适用于棒束燃料组件内冷却剂流动换热特性预测。本工作对压水堆堆芯安全分析工具开发具有参考和借鉴意义。     

氦-3回路的研究与设计

为使中国先进研究堆（CARR）具备开展压水堆燃料瞬态试验的能力,本工作对氦-3回路进行研究与初步设计。文章描述了氦-3回路的工作原理、设计参数和工艺流程。研究结果表明,氦-3回路能够快速、均匀、灵活地调节试验燃料棒的功率,是CARR实现压水堆燃料功率瞬态变化的优选方案。     

商用压水堆核燃料研发进展与应用展望

压水堆(PWR)是目前核电厂反应堆的主力堆型,而核燃料是反应堆的能量源泉和放射性裂变物质的主要来源,关乎核电厂的经济性和安全性。本文对当前国际上面向商用PWR应用研发的掺杂UO₂燃料、高铀密度燃料、微封装燃料和金属燃料的性能特点、技术状态及前景进行了归纳和评价。在掺杂UO₂燃料中,大晶粒燃料具有较高的技术成熟度,将在PWR实现大规模商用;高铀密度燃料和金属燃料在高温水腐蚀氧化问题以及事故下的行为仍待研究解决;具有极致安全的微封装燃料更适合特殊用途的小型反应堆。应协同开展先进燃料组件设计、建立设计准则以及研发高保真的性能分析技术等,以充分发挥新型燃料的可靠性及高燃耗优势。     

Hydride fuel for LWRs—Project overview

This special issue of Nuclear Engineering and Design consists of a dozen papers that summarize the research accomplished in the DOE NERI Program sponsored project NERI 02-189 entitled “Use of Solid Hydride Fuel for Improved Long-Life LWR Core Designs”. The primary objective of this project was to assess the feasibility of improving the performance of pressurised water reactor (PWR) and boiling water reactor (BWR) cores by using solid hydride fuels instead of the commonly used oxide fuel. The primary measure of performance considered is the cost of electricity (COE). Additional performance measures considered are attainable power density, fuel bundle design simplicity, in particular for BWRs, safety, attainable discharge burnup, and plutonium (Pu) transmutation capability.Collaborating on this project were the University of California at Berkeley Nuclear Engineering Department (UCB), Massachusetts Institute of Technology Nuclear Science and Engineering Department (MIT), and Westinghouse Electric Company Science and Technology Department. Disciplines considered include neutronics, thermal hydraulics, fuel rod vibration and mechanical integrity, and economics.It was found that hydride fuel can safely operate in PWRs and BWRs having comparable or higher power density relative to typical oxide-fueled LWRs. A number of promising applications of hydride fuel in PWRs and BWRs were identified: (1) Recycling Pu in PWRs more effectively than is possible with oxide fuel by virtue of a number of unique features of hydride fuel-reduced inventory of ²³⁸U and increased inventory of hydrogen. As a result, the hydride-fueled core achieves nearly double the average discharge burnup and the fraction of the loaded Pu it fissions in one pass is double that of the MOX fuel. (2) Eliminating dedicated water moderator volumes in BWR cores, thus enabling significant increase of the cooled fuel rod surface area as well as the coolant flow cross-section area in a given fuel bundle volume while reducing the heterogeneity of BWR fuel bundles, thus achieving flatter pin-by-pin power distribution. The net result is an increase in the core power density and a reduction of the COE.A number of promising oxide-fueled PWR core designs were also found in this study: (1) The optimal oxide-fueled PWR core design features a smaller fuel rod diameter (D) of 6.5 mm and a larger pitch to rod diameter (P/D) ratio of 1.39 than that presently practiced by industry of 9.5 mm and 1.326. This optimal design can provide a 27% increase in the power density and a 19% reduction in the COE provided the PWR can be designed to have the coolant pressure drop across the core increased from the reference 0.20 MPa (29 psi) to 0.414 MPa (60 psi). Under the set of constraints assumed in this work, hydride fuel was found to offer comparable power density and economics as oxide fuel in PWR cores when using fuel assembly designs featuring square lattice and grid spacers. This is because pressure drop constraints prevented achieving sufficiently high power using hydride fuel with a relatively small P/D ratio of around 1.2 or less, where it offers the highest reactivity and a higher heavy metal (HM) loading. (2) Using wire-wrapped oxide fuel rods in hexagonal fuel assemblies, it is possible to design PWR cores to operate at ∼50% higher power density than the reference PWR design that uses grid spacers and a square lattice, provided 0.414 MPa coolant pressure drop across the core could be accommodated. Uprating existing PWRs to use such cores could result in up to 40% reduction in the COE. The optimal lattice geometry is D = 9.34 mm and P/D = 1.37. The most notable advantages of wire-wraps over grid spacers are their significantly lower pressure drop, higher critical heat flux, and improved vibration characteristics.The achievement of the highest power gains claimed in this study is possible as long as mechanical components like assembly hold-down devices (both in PWRs and in BWRs) and steam dryers (only in BWRs) are appropriately upgraded to accommodate the higher coolant pressure drop and flow velocities required for the high-performance LWR designs. The compatibility of hydride fuel with Zircaloy clad and with PWR and BWR coolants need yet be experimentally demonstrated. Additional recommendations are given for future studies that need to be undertaken before the commercial benefits from use of hydride fuel could be reliably quantified.     

关于PWR及CANDU堆先进燃料管理策略的研究

阐述开展先进燃料管理策略研究的必要性与紧迫性。对我国秦山核电厂的燃料管理策略的改进进行了初步探讨,包括提高富集度延长循环长度、增大平均卸料燃耗、应用先进可靠毒物和低泄漏优化换料、改进燃料组件设计和适当提高功率等,并对可能取得的重大经济效益进行了讨论。提出研究ＰＷＲ的乏燃料在ＣＮＡＤＵ堆中应用及形成ＰＷＲ／ＣＡＮＤＵ联合燃料循环的可行性,以提高燃耗深度,增加能量输出,降低发电成本。     

压水堆燃料辐照性能多物理耦合并行分析程序开发

核燃料元件是反应堆的核心部件,其性能影响反应堆的安全性与经济性,利用燃料元件性能分析程序开展燃料堆内稳态辐照性能分析对于燃料设计及安全评价具有重要意义。通过开发燃料温度分布、变形计算、裂变气体释放及内压等模型,结合燃料元件热工-力学多物理耦合计算分析耦合方案,基于先进并行计算方法构建了高性能并行化燃料性能分析程序Athena。利用典型商用压水堆核电站数据及同类程序计算结果进行了程序初步验证,结果表明Athena程序计算结果合理可靠。通过定义堆芯功率及热工水力边界条件,程序能够并行开展压水堆全堆芯燃料辐照性能分析,提高燃料辐照性能分析效率,是数值反应堆原型系统（CVR1.0）的重要组成。     

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.     

压水堆核电站采用环形燃料元件可行性研究

环形燃料是一种由两层包壳和环形芯块构成的内、外两面冷却的新型、高效和安全的燃料元件,能够在保持或增进现有反应堆安全性能的前提下,大幅提高核电厂功率密度20%～50%,是高性能轻水堆核燃料的主要发展趋势之一。开展了环形燃料概念设计、堆芯物理、热工水力、反应堆安全、辐照性能、经济性和制造可行性等方面的研究,结果显示出压水堆核电厂采用环形燃料的优势和可行性。     

压水堆核电站大破口失水事故分析

压水堆核电站安全分析报告是核安全监管部门对其进行安全审查的重要文件,大破口失水事故是核电站运行的设计基准事故,是安全分析报告中的重要内容。本文使用RELAP5/MOD3.2进行压水堆冷管段大破口失水事故的计算,对比发现一回路冷管段发生双端断裂大破口时燃料元件包壳温度峰值（PCT）最高,且长时间维持在较高温度,此条件下反应堆最危险。计算结果表明,事故发生后,一回路压力迅速下降,堆芯冷却剂的流动性变差,导致堆芯裸露,燃料包壳温度又重新回升。通过安注系统和辅助给水系统等一系列动作,能保证燃料元件包壳温度不超过1204 ℃的限值。     

适用于新型PWR燃料组件的CHF关系式的开发及应用

以中国核动力研究设计院（NPIC）的棒束临界热流密度（CHF）实验数据为依据，基于具有自主知识产权的子通道分析程序CORTH，采用最小偏离泡核沸腾比（DNBR）点法开发了适用于新型压水堆（PWR）燃料组件的CHF关系式（CF-DRW关系式）并对其进行了应用分析。典型事故分析结果表明，采用CF-DRW关系式的计算结果相比FC-2000关系式具有相当或者更大的热工裕量。      

Critical heat flux and heat transfer above mixture level under high-pressure boil-off conditions in PWR type and tight-lattice type fuel bundles

Heat transfer tests were conducted in PWR 17 × 17 type and tight-lattice type fuel bundles under high-pressure boil-off (very-low flow, mass fluxes lower than 100 kg/m²s) conditions. There is almost no significant difference in both critical heat flux (CHF) (or dryout point) data and convective heat transfer data above the mixture level between the PWR type and tight-lattice type bundles. The “complete vaporization equation” predicts well the CHF data, i.e. the dryout occurs nearly at the elevation where the thermal-equilibrium quality reaches 1.0. The Groeneveld CHF table used in the RELAP5/MOD3 code should be improved in the region of mass flux between 10 and 100 kg/m²s. The radiative heat transfer has an important contribution to total heat transfer above the mixture level. The Dittus-Boelter correlation, with use of the film temperature in evaluating steam properties, predicts well the convective heat transfer above the mixture level.     

Design of highly moderated pressurized water reactor based on critical heat flux considerations

The difficulty of applying the existing critical heat flux correlations to the design of a modified 100% mixed plutonium uranium oxide fuelled assembly are outlined. A core with increased moderating ratio (RMA) was designed. The purpose of this design modification is to permit the consumption of large amounts of plutonium which cannot be burned in a conventional pressurized water reactor (PWR) owing to safety considerations. The design criteria required that the minimum departure from nucleate boiling ratio (DNBR) in the modified fuel does not exceed the value of the standard N4 PWR, which served the basis for the design. The desired conditions were achieved by reducing the fuel pin diameter to increase the moderating ratio and increasing the number of grids with mixing vanes to improve the minimum DNBR in the modified assembly. The design methodology and some of the proposed design options are presented.