157组燃料组件组成的堆芯燃料管理研究

本文应用SCIENCE程序包对157组燃料组件组成的压水堆堆芯进行换料优化燃料管理研究,给出了3个年换料和2个18个月换料共5个设计方案,每个设计方案给出了从首循环到第8循环共8个循环的主要计算结果,并进行了分析比较。综合来看,OUT-IN装载的设计方案功率峰值偏低,IN-OUT装载的设计方案功率峰值偏高,但均在设计限值以内;1/4堆芯换料设计方案的平均卸料燃耗最深,表明其组件燃耗得最充分,经济性较好。     

换料组件数为64组的堆芯燃料管理方案研究

本文应用SCIENCE程序包对157组燃料组件组成的压水堆反应堆堆芯进行18个月换料优化燃料管理研究,给出了采用64组换料组件时的共4个设计方案,每个设计方案给出了从首循环到第8循环共8个循环的主要计算结果,并进行分析比较。如果采用现有富集度4.45%的燃料组件即方案1,平衡循环长度较短,为460EFPD。为了增加循环长度,在仍然采用64组换料组件的情况下,需提高换料组件的富集度到4.95%,即方案2。而为了降低组件最大卸料燃耗,采用了两种富集度的燃料组件,且采用不同的组合,即方案3和方案4。     

大亚湾核电站第十循环堆芯换料设计

大亚湾核电站第十循环堆芯是自提高燃料组件富集度后实现18个月燃料循环的第二个循环堆芯。堆芯换料设计采用SCIENCE核程序包进行计算,辅以HADESⅡ处理程序,自动生成一维模型和综合法程序输入数据,并且自动生成换料设计报告。报告介绍了堆芯换料设计须提交给核电站的设计文件和所用的计算机软件,并对启动物理试验实测值与设计预计值进行了比较分析。     

大功率压水堆堆芯燃料管理设计

设计了一种大功率压水堆堆芯,对其中可燃毒物装载方案、平衡循环布置、首循环装料及过渡循环方案进行了研究。采用特征统计算法CSA燃料管理优化程序,快速高效地搜索堆芯装载和可燃毒物配置优化方案。采用堆芯核设计程序CPACT进行全堆计算,结果真实可靠。分别设计了18个月和24个月换料两种方案,计算结果表明,在满足堆芯燃料管理所有限值要求的前提下,两种方案均从第4循环开始进入平衡循环。     

秦山核电二期工程长燃耗堆芯可行性方案论证

简要介绍了对秦山核电二期工程反应堆换料采用的高性能燃料组件的长燃耗堆芯进行核设计可行性方案的论证，经论证表明，推荐方案中：如反应堆采用三分之一，混难事型换料方案和载钆可燃毒物，平衡堆芯批卸料燃耗达４２ＧＷ·ｄ／ｔ（Ｕ）左右时，平衡循环长度可达４１０等效功率天（ＥＦＰＤ）满足各方面限值要求。     

中广核CPR1000核岛堆芯概念设计和安全裕度评估初探

CPR1000压水堆核电站是中广核集团20多年来经过渐进式改进和自主创新形成的中国改进压水堆核电站。CPR1000的参考设计是岭澳II期核电站加改进设计。在未来的10～15年内,CPR1000将是中广核集团主要建设的核电站类型之一。CPR1000的初始堆芯设计采用什么样的装料方式和燃料循环方式是必须首先解决和确定的重要设计前提,这是整个核岛设计、安全分析核执照申请的核心和基础。基于大亚湾核电站和岭澳核电站多年的燃料管理经验和运行经验以及国外类似核电站运行和设计经验,并且综合考虑了初始堆芯的特点和难点,以及不同堆芯设计和燃料管理策略的特点,对CPR1000的初始堆芯进行了设计。通过初步研究,本文提出了CPR1000初始堆芯采用的燃料组件类型,分析CPR1000采用从首循环开始进行18个月换料过渡的堆芯设计技术方案,并对CPR1000首循环实施18换料进行了堆芯设计安全裕度初步分析与评估。     

压水堆多堆联合堆芯装载设计技术应用研究

在压水堆之间开展的多堆联合堆芯装载设计技术能提高堆芯装载设计的灵活性,研究其对燃料经济性和机组在燃料组件损坏情况下应对能力的影响。分析了多堆联合堆芯装载设计技术对来自其他机组燃耗过燃料组件的相容性要求。计算了燃耗过燃料组件的余热与冷却时间的关系,列出了可用于燃耗过燃料组件运输的大负荷乏燃料运输容器。分析了首循环堆芯和换料堆芯面对燃料组件损坏后堆芯装载设计的应对能力,以及在采用多堆联合堆芯装载设计技术后的应对能力改善情况。开展了多堆联合首循环堆芯装载模拟设计并分析了其经济性。研究结果表明,首循环出现燃料组件损坏的风险比后续换料堆芯大,而由全新燃料组件组成的首循环应对燃料组件损坏的能力最低;通过多堆联合堆芯装载设计技术可以把首循环转化为换料堆芯,不仅能凭借换料堆芯的设计灵活性提高应对燃料组件损坏的能力,还能节省约3.2亿元人民币的燃料费。因此,多堆联合堆芯装载设计技术能提高首循环的燃料经济性和机组面对燃料组件损坏后的应对能力。     

大亚湾核电站18个月换料堆芯热工水力设计

大亚湾核电站从第九循环开始由年换料向18个月换料转换，使得原堆芯热工水力设计不再适用18个月换料的热工水力设计采用法马通新开发的临界热流密度（CHF）关系式－－FC关系式，并用全统计法代替原来的确定论方法确定DNBR设计限值。由于在过渡循环中AFA 2G和AFA 3G燃料组件混装，使混合堆芯的最小DNBR小于均匀堆芯的最小DNBR。本文确定了一个包络的混合堆芯DNBR亏损规律。并在此基础上得到了过渡循环和平衡循环的堆芯物理限值线。以及新的超温ΔT超功率ΔT保护定值。     

采用TVS-2M组件的VVER堆芯燃料管理研究

使用KASKAD程序包,对田湾核电站从首循环开始使用TVS-2M组件展开研究,提出相应的燃料组件设计。以此为基础展开燃料管理研究,提出3个燃料管理方案(年换料方案和两个长周期换料方案)。对每个方案中堆芯的安全参数及其他重要参数进行分析,结果表明各种安全参数均满足设计要求。长周期的换料方案是从首循环就开始使用TVS-2M组件,并且只经过2个循环的过渡,寿期长度便达到了长周期的要求。长周期换料方案可提高电厂的年均能力因子,并在整个堆芯寿期内减少大修次数,因而每年节约30.8%的大修费用,因此电厂的经济效益得以提高。     

RPN系统中间量程探测器性能监督方法

为提高福清核电RPN系统设备可靠性,基于设备工作原理及福清现场实践经验,构建了一套RPN系统中间量程探测器的性能监督方法。该方法针对RPN系统中间量程探测器的4种主要失效模式,选取典型参数进行监督,以期提前发现故障征兆,及时采取干预措施,提升设备可靠性。     

数字化核仪表系统样机堆上试验及验证

针对目前国内核电厂核仪表系统设备主要依赖进口的现状,设计研发了一套数字化核仪表系统样机,系统样机主要包括中子探测器组件、信号调理和处理样机以及信号监控设备。通过介绍样机在商用堆上的安装和试验情况,详细分析了反应堆启堆、升功率、满功率及降功率运行期间的试验数据。试验结果表明,中子探测器与信号调理和处理样机配合良好,整套系统样机运行稳定可靠。      

Analytical Method for Recalibration of Incore and Excore Nuclear Instrumentation System of PWRs

In pressurized water reactors (PWRs) of nuclear power plants, it is periodically required to recalibrate the correlation between the incore and excore nuclear instruments, where the recalibration guarantees the accurate monitoring of incore axial power distributions by excore instrumentation. In order to obtain the correlation between incore power distribution and excore detector response, however, it is necessary to perturb the axial power distribution of the reactor by control rods or to introduce a xenon spatial oscillation.

In this paper a simplified analytical method that eliminates such perturbations is proposed to evaluate the correlation between incore power distribution and excore detector currents which is utilized to recalibrate the nuclear instrumentation system. This method provides as good an accuracy as that obtained by conventional methods.     

船用核动力装置16N在线检测系统研究

在线监测主蒸汽系统的放射水平是提高船用核动力和气轮机系统的可用率和运行重要措施之一。核电厂已装备了NaI探头^16N辐射在线监测系统,但不适用于船用。讨论了以BGO为探头的紧凑型船用^16N辐射在线检测仪研发的关键问题,检测系统设计方案及其关键技术。     

基于数值微分核脉冲信号数字处理方法

核反应堆核测量系统测量探测器输出的核脉冲信号,该信号后沿拖尾很长,在计数率较高时容易产生信号堆积和基线漂移等问题,导致源区计数率测量上限仅能达到10⁵ Hz左右。文中基于数值微分方法,采用数字处理技术,在时域上分析了核脉冲信号经过前置放大、信号成形、低通滤波和脉冲甄别后的输出,并利用探测器实测信号进行了仿真。仿真结果表明,基于数值微分的数字处理方法可以实现相邻0.4μs脉冲信号的识别和测量,将源区测量计数率上限提升到2×10⁶ Hz以上。     

浅谈核电厂仪表控制系统电磁兼容性要求

针对现有核电厂设备电磁兼容性测试的不足,探讨了国内核电厂仪表控制系统在核电厂电磁环境下的电磁兼容性要求和评价方法。建议加强核电厂仪表控制系统的电磁兼容性设计,以保障核电厂安全、稳定的运行。     

核电厂堆外中子测量系统闪发高计数率问题研究

从故障现象、原因排查和解决措施三个方面,分析了我国核电厂近期发生的堆外中子测量系统闪发高计数率异常中所涉及的电缆接头问题和探头故障问题,提出了核电厂应关注堆外中子测量系统设备制造和安装的质量等建议,为解决和避免类似的堆外中子测量系统闪发高计数率问题提供借鉴。     

In vessel neutron instrumentation for sodium-cooled fast reactors: Type,lifetime and location

The in vessel instrumentation of sodium-cooled fast reactors must deliver measurements that are reliable and easy to interpret over several reactor cycles in order to fulfill the safety requirements. This paper compares, with respect to this requirement, three types of detectors that are widely used in neutron measurements: fission chambers, boron-lined proportional counters, self-powered neutron detectors. We use neutron spectra that are computed for preliminary design of sodium-cooled fast reactor in different representative locations: in diluting tubes within nuclear fuel assemblies, or in the lateral neutron protections. With an evolution code, we compute the expected signal for each type of detector, to assess whether its level is sufficient, and also its evolution over three operating cycles, to examine whether it is compatible with long term measurements. The conclusion is that fission chambers are the only type able to deliver an interpretable signal for a wide dynamic of reactor power and for three or more operating cycles. The two other types are shown to be inadequate.     

Integrated software safety analysis method for digital I&C systems

The digitalized Instrumentation and Control (I&C) system of Nuclear power plants can provide more powerful overall operation capability, and user friendly man-machine interface. The operator can obtain more information through digital I&C system. However, while I&C system being digitalized, three issues are encountered: (1) software common-cause failure, (2) the interaction failure between operator and digital instrumentation and control system interface, and (3) the non-detectability of software failure. These failures might defeat defense echelons, and make the Diversity and Defense-in-Depth (D3) analysis be more difficult. This work developed an integrated methodology to evaluate nuclear power plant safety effect by interactions between operator and digital I&C system, and then propose improvement recommendations. This integrated methodology includes component-level software fault tree, system-level sequence-tree method and nuclear power plant computer simulation analysis. Software fault tree can clarify the software failure structure in digital I&C systems. Sequence-tree method can identify the interaction process and relationship among operator and I&C systems in each D3 echelon in a design basis event. Nuclear power plant computer simulation analysis method can further analyze the available backup facilities and allowable manual action duration for the operator when the digital I&C fail to function. Applying this methodology to evaluate the performance of digital nuclear power plant D3 design, could promote the nuclear power plant operation safety. The operator can then trust the nuclear power plant than before, when operating the highly automatic digital I&C facilities.     

Development of the HL-2M digital pulse analysis system

A digital pulse analysis system is an important diagnostic system in nuclear physics experimental research. In response to the demand for reflecting the particle state in a nuclear physics experiment, we have designed and developed a real-time digital pulse analysis system and applied it to the digital nuclear pulse waveform discrimination of different detectors in the HL-2 M tokamak. The system is based on the peripheral component interconnect extensions for instrumentation(PXI) platform, while its software was written in LABVIEW. The key technologies involved in the system implementation include digital pulse analysis technology,digital discrimination technology, pulse height analysis technology, etc. The system has been applied to the plastic scintillator detector at the Neutron Source Lab of the University of Science and Technology of China. And the experimental results indicate that the system can discriminate between neutron(n) particles and gamma(γ) particles well when used to measure the plastic scintillator detector.     

核仪表系统系数设置影响因素分析

核电厂在投入商业运行前,需进行一系列的调试试验确保系统和运行的可靠性。2010年8月,某百万千瓦级核电厂在50％FP甩负荷到厂用电的调试中意外停堆,导致停运两天,拖延了试验的进度。经现场分析发现意外停堆与RPN系数设置不当有关。中国核动力研究设计院快速响应并对事件原因进行了分析,提出了新的RPN系数,顺利完成了50％FP和100％FP甩负荷到厂用电试验。文章首先分析了RPN系数对中子注量率变化率计算的影响,然后分析了影响RPN系数设置的主要因素,给出了RPN系数设置的基本原则,为后续电站的调试试验与运行提供了参考,以避免同类事件再次发生。