Preliminary design for inland nuclear power plants completed

The State Nuclear Power Technology Corporation (SNPTC), which is responsible for the development of third-generation nuclear power technology in China, has completed the preliminary designs

Challenges in spent nuclear fuel final disposal:conceptual design models

<正>The disposal of spent nuclear fuel is a long-standing issue in nuclear technology.Mainly,UO_2 and metallic U arc used as a fuel in nuclear reactors.Spent nuclear fuel contains fission products and transuranium elements,which would remain radioactive for 10~4 to 10~8 years.In this brief communication,essential concepts and engineering elements related to high-level nuclear waste disposal are described.Conceptual design models are described and discussed considering the long-time scale activity of spent nuclear fuel or high level waste.Notions of physical and chemical barriers to contain nuclear waste are highiightened.Concerns regarding integrity,self-irradiation induced decomposition and thermal effects of decay heat on the spent nuclear fuel are also discussed.The question of retrievability of spent nuclear fuel after disposal is considered.

Study on Site Specific Design Earthquake Ground Motion of Nuclear Power Plants in China

The main technical backgrounds and requirements are introduced with regard to earthquake ground motion design parameters in several domestic and American standards,codes and guides involved in the seismic analysis and design activities of nuclear power plants in China.Based on the research results from site seismic safety evaluation of domestic nuclear power plant projects in the last years,characteristics and differences of site specific design spectra are analyzed in comparison with standard response spectra,and the suitability of standard response spectra for domestic nuclear power plant projects is discussed.

Design and Comparative Analysis of Small Modular Reactors for Nuclear Marine Propulsion of a Ship

The fast growth in the size and difficulty of nuclear power plant in the 1970s produced an interest in smaller, modest designs that are intrinsically safe over the usage of design features. With the development of nuclear technology, there is the need for revolution in the Maritime sector, especially the advance marine propulsion. In current years, numerous reactor manufacturers are dynamically improving small modular reactor designs with even superior use of safety features. Several designs integrate the ultimate in greater safety. They totally remove specific accident initiators from the design. Other design features benefit to reduce different types of accident or help to mitigate the accident’s consequences. Although some safety features are mutual to maximum SMR designs, irrespective of the coolant technology, other features are specific to liquid-metal cooled, water, gas, or SMR designs. Results: There have been more reactor concepts investigated in the marine propulsion area by different assemblies and research laboratories than in the power generation field, and much can be learned from their experience for land applications. The extensive use of safety features in SMRs potential to make these power plants extremely vigorous, protecting both the public and the investor. Conclusion: For these two considerations, it is recognized that a nuclear reactor is the ideal engine for naval advanced propulsion. The paper will present the work to analyze the concept design of SMRs and design a modular vessel consisting of a propulsion module.

Response spectra for nuclear structures on rock sites considering the near-fault directivity effect

Near-fault ground motions, potentially with large amplitude and typical velocity pulses, may significantly impact the performance of a wide range of structures. The current study is aimed at evaluating the safety implications of the near-fault effect on nuclear power plant facilities designed according to the Chinese code. To this end, a set of near- fault ground motions at rock sites with typical forward-directivity effect is examined with special emphasis on several key parameters and response spectra. Spectral comparison of the selected records with the Chinese and other code design spectra was conducted. The bi-normalized response spectra in terms of different comer periods are utilized to derive nuclear design spectra. It is concluded that nuclear design spectra on rock sites derived from typical rupture directivity records are significantly influenced both by the earthquake magnitude and the rupture distance. The nuclear design spectra specified in the code needs to be adjusted to reflect the near-fault directivity effect of large earthquakes.

响应面法优化冷冻鸡血豆腐的制备工艺

为使肉鸡副产物之一鸡血高效利用。以新鲜鸡血为主要原料,采用低场核磁不易流动水峰面积为评价标准,通过添加不同品种与浓度的抗冻改良剂对冷冻鸡血豆腐中不易流动水峰面积影响的单因素实验,再结合响应面设计优化冷冻鸡血豆腐的制备工艺。进一步以失水率、质构、色泽指标对优化的冷冻鸡血豆腐的品质进行分析,为开发冷冻鸡血豆腐产品提供科学依据。结果表明,冷冻鸡血豆腐优化制备工艺为:添加脂酰乳酸钠0.1%,海藻酸钾0.1%,木薯变性淀粉2%。在此工艺条件下,冷冻鸡血豆腐的低场核磁不易流动水峰面积为2143,与空白组和其他优化组相比不易流动水峰面积最大,故保水效果最好。优化制备的冷冻鸡血豆腐的硬度、咀嚼性、凝胶性以及亮度都显著(P<0.05)高于空白组(其中硬度是空白组的17倍,亮度的是空白组的1.13倍)。因此,冷冻鸡血豆腐的制备工艺可显著提高解冻后鸡血豆腐的品质,可促进肉鸡副产物高值化产品的开发。

模拟事故工况下非能动核电厂安全相关涂层的可靠性测试及评估方法研究

安全相关涂层在非能动核电厂中起着重要的作用,涂层的失效会影响核电厂安全系统的功能执行,影响核安全。国内外核监管机构对于在设计基准事故(DBA)工况下涂层系统的可靠性及评估方法非常重视。文章结合非能动核电厂涂层系统的工程应用,针对其在DBA下的可靠性及评估方法进行了研究。研究表明:在DBA下非能动核电厂安全相关涂层的可靠性要综合考虑涂层的模拟DBA性能、干膜密度、导热性能等。而非能动核电厂安全相关涂层工程应用,则需从涂层的模拟DBA性能、干膜密度、导热性能、涂层碎片(数量、大小、位置和性能等)以及包络涂层碎片后的碎片裕量等角度进行综合评估,以确定在事故工况下涂层的可靠性,不对系统安全产生影响,保证核电厂更安全、高效和经济性运行。

空间堆S-N_(2)O布雷顿—有机朗肯联合系统性能研究与优化

针对空间核动力热电转换系统发电效率无法满足深空探测任务的需求,结合布雷顿循环和有机朗肯循环的热电转换模式,制定了以N_(2)O作为顶部循环工质,利用底部有机朗肯循环对系统余热进一步回收。通过建立联合系统数理模型,研究了系统分流比、顶部循环涡轮进口压力和温度对系统热力学性能、比质量和经济性能的影响。基于非支配排序遗传算法对系统进行多目标优化来获得系统关键性能指标的帕累托解集。结果表明:与单独顶部循环相比,联合系统热效率提高了约7.77%。通过多目标优化找到多组帕累托解,为高效空间核动力系统设计和优化提供理论依据。

核电全厂通信系统设备采购管理浅析

核电站全厂通信系统具有种类繁多、数量多、覆盖全厂等特点。目前,通信设备采购在设计、采购、安装单位管理3个方面均存在较多问题,此类问题制约工程进展。针对上述问题,论文从设计、采购、安装管理3个方面提出了工作对策,旨在提高采购工作效率,保证设备按时到货,保障现场安装进度。

基于多目标进化算法的反应堆辐射屏蔽优化方法研究

新型核能与核动力装置的发展对辐射屏蔽设计方法提出了更高要求。面对空间堆、船用堆等装置的小型化、轻量化设计需求,传统辐射屏蔽多目标优化方法存在优化目标少、优化参数单一、全局性差等缺陷,难以满足辐射屏蔽智能设计的需求。本文基于第三代非支配排序遗传算法和改进多目标人工蜂群算法开展面向反应堆屏蔽层重量、体积和特定区域辐射剂量等多目标约束条件下的辐射屏蔽优化方法研究,并对各算法的优化性能、优化方案进行对比分析。结果表明,本文方法相较于传统屏蔽智能设计方法展现了更好的优化性能,并在实际工程问题中体现了可靠性,可为辐射屏蔽设计优化提供新思路。

核级石墨PFC^(2D)模型的宏细观参数分析

颗粒流软件PFC能够很好地处理非连续介质力学的问题,模拟核级石墨的损伤和断裂过程具有独特的优势。在采用PFC^(2D)模拟核级石墨时,为了增加模拟的便捷性,实现了PFC^(2D)细观参数的快速标定。本文首先设计了四因素四水平的正交试验,分析了核级石墨的宏观参数与PFC^(2D)的细观参数关系;然后基于建立的核级石墨宏细观参数的关系,采用PFC^(2D)模拟石墨IG11的三点弯曲试验,并从位移场变形、荷载-位移曲线和断裂参数三个方面验证其宏细观参数关系的正确性。结果表明,弹性模量E随平节理模量E_(c)的增加线性增加;抗拉强度σ_(t)随平节理抗拉强度σ_(b)的增加线性增加;泊松比ν随平节理刚度比k_(n)/k_(s)的增加先增大后减小;数值模拟的位移场变化和荷载-裂缝口张开位移P-CMOD曲线与试验结果较吻合,此外,断裂参数的最大误差不超过3.57%。

核结构材料用多主元合金辐照损伤的研究进展

开发具有优异综合性能的核反应堆结构材料是核能发展的基础,并且是长期以来制约核能推广的难点之一。多主元合金(multiprincipal element alloys,MEAs)因具有良好的抗辐照性能、力学性能而被认为是先进反应堆结构材料的候选材料,为新型抗辐照材料的设计开辟了广阔空间。近年来,有关多主元合金在辐照损伤方面的研究多试图揭示多主元合金一些因素和特性对辐照过程中缺陷形成与演变的影响。例如:主元种类和数目、主元浓度、晶格畸变、化学短程序等。尽管现有的一些研究结果表明以上因素可以提高多主元合金抗辐照损伤能力,但是在不同辐照条件下,以上因素对多主元合金中缺陷形成和演变的影响机制存在较大差异,难以得出普适性的结论。本文围绕FCC和BCC系两类多主元合金的辐照肿胀、氦泡形成、辐照诱导元素偏析和相变、辐照硬化四方面内容,综述了近年来多主元合金在辐照损伤方面的研究进展,总结了多主元合金提高抗辐照性能的作用机制,并在此基础上对核电结构用多主元合金的未来研究方向做出了展望,包括短程序调控、高熵陶瓷、增材制造、高通量结合机器学习加速材料开发等。最后指出必须从合金成分设计的角度出发,基于材料服役的实际环境来设计新型抗辐照多主元合金。

核主泵水力优化技术与水力稳定性研究进展

随着人们对能源安全、环境保护和可持续发展的关注不断增加,核能作为一种清洁、高效的能源形式备受瞩目。核主泵作为核电站的重要组成部分,其水力性能的好坏关系到整个核电站是否能够长期安全稳定高效的运行。本文针对核主泵水力优化技术与水力稳定性的研究进展开展论述,介绍了核主泵水力性能的影响因素及其研究方法。以水力优化设计与压力脉动特性等方面为切入点深入探讨了核主泵水力优化技术与水力稳定性的研究现状,并简要介绍了核主泵压力脉动的形成原因;总结了已有的研究成果,并根据现有的研究基础展望了核主泵未来的技术发展趋势。

HL-3聚变用大容量Crowbar

[目的]面向HL-3聚变大科学装置极端故障能量快速转移,配置短路保护用大容量旁路开关需求。[方法]基于HL-3磁体电源参数,结合电力电子变流技术及可控硅触发应用基础,针对快速冲击应用中多并联可控硅一致性导通及稳定均流的严苛工况,融合电力电子变流工程实践经验,经并联分支导电线路等长型式改良;并联元件参数一致性筛选,并联臂内元件位置与参数差异化选配;通过在可控硅侧附加串联电阻物理强迫均流,并对附加电阻热稳定性核算,并联桥臂电动力冲击验算;可控硅元件高频脉冲列强触发等手段和方法。[结果]推导设计了一种稳定的可控硅220 kA级大容量聚变用crowbar旁路保护电子开关系统,经试验模拟实测验证及实载160 kA冲击考验,实现了HL-3保护用大容量旁路开关的匹配要求。[结论]在做好热动、稳定性验算的基础上,采用线路改良,附加物理强迫均流,元件参数筛选及装配,触发模式调整等综合手段,实现大容量crowbar在聚变中的配套应用是完全可行的。

框架式翻转架结构设计与应用

核工程中部分核设备需要在厂房内翻转,现有翻转工艺无法完全满足工程需要,特别是大型薄壁容器及异形核设备的翻转。文中总结了框架式翻转工艺,基于框架式翻转架的功能分析,提出了一套框架式翻转架结构设计方法,并在大型薄壁容器及异形核设备的翻转上成功应用,为工程中核设备翻转框架设计提供技术借鉴。

核电厂一回路计算机压力控制系统的研发与应用

核电机组一回路是核电厂的第二道安全屏障,其可靠性和严密性对于核电安全至关重要。我国几乎所有核电机组进行一回路承压能力和密封测试的压力控制均为手动操作,控制精度低、劳动强度大、压力读数误差大,存在很大的人因失效风险。针对该问题,研发了一套计算机控制系统来实现水压试验的自动控制,阐述了该计算机控制系统软、硬件系统设计和实现方案,并介绍了人机交互界面的功能要素和实现形式。结果表明,该系统能很好地完成一回路水压试验过程的自动控制任务。

Neutronic design investigation of a liquid injection-based second shutdown system for a typical research reactor using MCNPX

Safety systems, built on state-of-the-art technology, are essential for achieving acceptable levels of plant safety to minimize hazards to the reactor and the general public. The second shutdown system(SSS) as an engineered safety feature and a part of the reactor protection system(RPS) is a means for rapidly shutting down a nuclear reactor, keeping it in a subcritical state and serving as a backup to the first shutdown system(FSS). In this research, one SSS with two types of optimum chamber designs is proposed that take into account the main current characteristic features of the Tehran research reactor with improvements over earlier designs. They are based on a liquid neutron absorber injection that is preferably different, diverse, and independent from the FSS based on the rod drop mechanism. The major design characteristics of this SSS with two different chambers were investigated using MCNPX 2.6.0 code. The performed calculations showed that the designed SSS is a reliable shutdown system, assuring an appropriate shutdown margin and injection time, with no significant effects on the effective delayed neutron fraction while causing minimal variations to the core structure. Further, the reasonable financial cost and the prolongation of the operation cycle are additional advantages of this design.

Research on CAD/CAE/MDO integrated system for nuclear valve products

In order to realize rapid response to performance-driven design for nuclear valve products, a rapid design platform is constructed by combining the structural features with the multi-characteristic optimization requirement, which integrates parametric design, multi-characteristic simulation, and multidisciplinary design optimization of nuclear power valves together. Through the integration of heterogeneous CAX/MDO software, such as UG, ANSYS and iSIGHT, the analytical efficiency is enhanced by coupling respective and unique advantage of various software systems, and the mutual coordination and information sharing are realized between the heterogeneous systems. The application example shows that the system is convenient operation that improves the automation in design and analysis on nuclear valves. The results can provide a practical and effective development method and technical reference of the integrated design system for valve products.

Research on the attribution evaluating methods of dynamic effects of various parameter uncertainties on the in-structure floor response spectra of nuclear power plant

Consideration of the dynamic effects of the site and structural parameter uncertainty is required by the standards for nuclear power plants （NPPs） in most countries. The anti-seismic standards provide two basic methods to analyze parameter uncertainty. Directly manually dealing with the calculated floor response spectra （FRS） values of deterministic approaches is the first method. The second method is to perform probability statistical analysis of the FRS results on the basis of the Monte Carlo method. The two methods can only reflect the overall effects of the uncertain parameters, and the results cannot be screened for a certain parameter＇s influence and contribution. In this study, based on the dynamic analyses of the floor response spectra of NPPs, a comprehensive index of the assessed impact for various uncertain parameters is presented and recommended, including the correlation coefficient, the regression slope coefficient and Tornado swing. To compensate for the lack of guidance in the NPP seismic standards, the proposed method can effectively be used to evaluate the contributions of various parameters from the aspects ＆sensitivity, acuity and statistical swing correlations. Finally, examples are provided to verify the set of indicators from systematic and intuitive perspectives, such as the uncertainty of the impact of the structure parameters and the contribution to the FRS of NPPs. The index is sensitive to different types of parameters, which provides a new technique for evaluating the anti-seismic parameters required for NPPs.

堆外核测量系统输出高压纹波测试准确性研究

为解决堆外核测量系统输出高压纹波测试过程极易受到外部环境干扰,导致测试数据波动大、不稳定的问题,基于当前高压纹波测试现状,对高压纹波测试准确性进行研究。从数据仿真、结构设计、信号线缆优化、安全防护与验证、测试流程标准化等五个方面入手,通过总结其他行业高压纹波测试方法和结合堆外核测系统产生高压的频率特点,提出一种基于堆外核测量系统输出高压纹波测试准确性设计方法。利用数据仿真技术计算出符合堆外核测量系统高压纹波测试的隔直滤波电容容值,优化整个高压纹波测试外部环境以减少外部电磁干扰。通过验证测试环境的安全性、可靠性以及制定相关测试标准,确保堆外核测量系统输出高压纹波测试的准确性。该研究能够有效提升堆外核测量系统中输出高压纹波测试的准确性,解决了现存问题。