大型先进压水堆弹棒事故模拟及敏感性分析

相对于传统堆型,大型非能动先进压水堆堆芯设计具有重大改变,这些改变对弹棒事故分析具有重要影响,进而影响反应堆的安全性。通过选取典型的四类工况（寿期初满功率、寿期初零功率、寿期末满功率和寿期末零功率）,利用中子动力学软件和燃料性能分析程序开展大型先进压水堆CAP1400的弹棒事故模拟计算,验证大型先进压水堆弹棒事故工况下的安全性,并针对弹棒事故分析关键输入参数开展敏感性分析。计算分析结果表明：大型先进压水堆发生弹棒事故时,其结果能够满足验收准则的要求,反应堆处于安全可控状态;弹棒事故分析中功率峰值对弹棒价值最敏感,事故分析结果对停堆反应性敏感性较小。     

岭澳核电站硼稀释事故分析

在岭澳核电站混合堆芯及提高富集度论证项目中的反应性事故中,硼稀释事故分析占据着重要的地位。因为硼稀释事故分析的结论中要对冷停堆和热停堆下的控制棒棒位进行定义,其中热停堆下的控制棒棒位又是次临界下提棒事故分析的前提,而且在次临界提棒事故裕量不够时,可以通过硼稀释事故分析来对热停堆标准工况下的棒位进行一定的调整以提高次临界提棒事故裕量。对混合堆芯论证项目中的硼稀释事故分析的方法、结论进行了说明。     

超温ΔT紧急停堆整定值优化研究

对超温ΔT停堆信号中的关键参数进行优化,通过功率运行控制棒组(RCCA)失控抽出事故分析对优化后的超温ΔT停堆信号进行验证研究,采用热工水力子通道分析程序和瞬态分析程序对超温ΔT整定值设定进行分析,新的整定值将对停堆时间、最小偏离泡核沸腾比(DNBR)和反应性引入速率限值方面产生影响.分析结果表明,优化后的整定值在保证反应堆安全裕量的前提下增加了运行裕量,提高了反应堆经济性并能满足反应堆安全运行的要求.     

CNP1500核电站的DNBR与DNBR热工裕量分析

通过使用FLICA—IV程序对CNP1500核电站进行具有包络性的DNBR与DNBR裕量分析。在稳态DNBR分析中，使用确定论的DNBR验收准则值，在失流和落棒事故分析中，使用全统计法的DNBR验收准则值。尽管在CNP1500核电站的稳态、失流和落棒事故分析中具有包络性的焓升因子值1．7被使用，但DNBR仍然满足验收准则的要求，说明CNP1500核电站的堆芯是安全的。除落棒事故的DNBR裕量不满足15％的热工裕量外，稳态和失流事故的DNBR热工裕量满足15％的URD要求。     

基于MANTA/SMART三维物理热工耦合的落棒事故分析

传统CPR落棒事故分析采用保守的确定论分析方法,设计保守性较大,已制约了机组的持续改进及后续堆型的研发。为进一步挖掘落棒事故的热工设计裕量,本文以某大型压水堆为研究对象,建立了基于MANTA/SMART程序三维物理热工耦合的落棒事故分析方法,并与CPR落棒分析方法的过程及结果进行了对比分析。研究结果表明,基于MANTA/SMART的三维落棒事故分析方法合理可行,能更真实全面地反映反应堆瞬态过程中的参数特征,并能有效挖掘出落棒事故分析的安全裕量。     

三维耦合分析软件的落棒事故分析

作为数值反应堆中必不可少的物理和热工部分,中广核研究院有限公司开发了三维物理热工耦合分析软件,通过动态链接库技术实现了自主研发的核反应堆系统瞬态分析软件和三维核设计软件的耦合,并已与国际基准题结果对比验证。本文为耦合软件的应用,围绕华龙一号的落棒分析问题,开展不同落棒组合的耦合计算分析,并研究停堆棒组落棒和温度调节棒（R）棒组两组落棒对堆芯功率的影响。分析结果表明,非中心对称的棒组落棒事故会导致堆芯径向功率出现不对称,并使得堆芯出口回路温度不同。落棒反应性价值越大,R棒调节后的稳态功率回升相比初始稳态差异越大,DNBR公式计算值的变化趋势与功率呈现相反规律。     

大型先进压水堆硼稀释事故模拟分析

基于美国先进三维节块法堆芯计算程序,建立大型先进压水堆堆芯首循环,选取四个最不利的保守事故工况点,包括满功率工况、启动工况、热备用工况、冷停堆工况,分别进行硼稀释事故分析,计算得到初始条件下的硼浓度以及硼稀释事故的临界硼浓度,最终计算总的硼稀释时间、报警发生时间以及从报警到临界的时间,分析大型先进压水堆发生硼稀释事故工况下的安全性。计算结果表明:在发生硼稀释事故工况下,反应堆有足够的时间在丧失全部停堆裕量前终止硼稀释。     

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。      

CNP1000核电厂的安全性与经济性

核电厂的安全性是最重要的，但是没有经济性的核电厂是不受欢迎的。URD要求的15％的热工裕量不是法规文件。核安全部门关心的是反庆堆的安全而不是热工裕量。增大反应堆的热工裕量，就意味着在同等经济规模条件下的核电厂要降低其反应堆的热功率（经济性）。过去设计的反应堆都是严格按照核安全法规设计，而且采用非常保守的计算方法、公式和计算机程序进行设计，所得到的热工裕量非常小或者没有，但是这些反应堆仍然在安全运行着，如果现在采用新的计算方法、公式和计算机程序计算这些运行核电厂的热工裕量，应该是有所提高的。同时，用不同类型的计算方法、公式和计算机程序得到的热工裕量也是不相同的，所以热工裕量不是评价反应堆是否安全的标准。在经济不发达的中国，反应堆的安全性和经济性同样是非常重要。增大反应堆的热工裕量主要是为了防止核电厂在正常运行时偏离设计安全限值、增加反应堆应付事故和严重事故的能力。核电厂设计应该俦考虑如何保证在任何事故条件下反应堆能够及时停堆、不失电、提高ECCS的非能动能力和可靠性，同时使用那些被实验和实践证明的新设计方法、公式和计算机程序进行反应堆设计，切实提高反应堆的安全性和可靠性，在保证核安全的前提下充分提高核电厂的经济性。通过使用最新的子通道分析程序和最佳估算（方法）大破口失水事故分析程序对CNP1000核电厂（2775MW热功率，3.66m堆芯和3150MW热功率，4.27m堆芯）进行了DNBR裕量和大LOCA线功率裕量分析，计算的DNBR值和峰值包壳温度都满足验收准则的要求，其DNBR裕量和线功率裕量都满足15％的要求，反应堆是安全的。从安全和经济的角度，CNP1000核电厂应该选择3150MW热功率，4.27m堆芯为宜。     

非参数统计方法在DNBR限值计算中的应用

非参数统计方法是一种不依赖某种特定分布的统计推断方法,它通过抽样计算得到总体在一定概率水平和置信度下的容许限。DNBR裕量是核电站设计中的重要参数,能够反映核电站的安全水平。本文应用基于Wilks公式容许限的非参数统计方法及VIPRE-01程序计算300MW核电厂在全失流事故(LOFA)下的DNBR设计限值,并与ITDP方法得到的DNBR限值相比较,以期获得一定的DNBR裕量。结果表明,相对于ITDP方法,非参数统计方法获得了2.96%的DNBR裕量,该方法由于减少了分析过程中的保守性,能够提供更大的DNBR裕量,而DNBR裕量增加有利于堆芯换料方案的改进。     

跨间交混格架对EPR堆芯DNBR裕量的影响

采用FLICA Ⅲ-F子通道程序,分析了AFA 3GLE燃料组件加装跨间交混格架（MSMG）后对台山EPR堆芯热工参数和最小DNBR的影响。分析结果表明,在名义工况下加装MSMG后,轴向功率呈余弦分布和轴向功率偏差AO=+9%将分别提高EPR堆芯的DNBR裕量约为24%和28%,同时增加EPR堆芯压降约10.1%。     

RTDP和MSG方法比较研究

偏离泡核沸腾比(DNBR)设计限值是反应堆安全分析的基础,合理的计算方法有助于发掘更多的安全裕量。本文对比分析了核电站设计中常采用的RTDP和MSG方法,介绍了方法原理和计算流程,并以广东岭澳一期核电站为例,分别进行了DNBR设计限值计算。结果表明:虽然两种方法的原理不同,但在相同的工况和统计学输入参数条件下,DNBR设计限值相近,设计中可根据具体应用需求进行选择。     

Design features and thermal margin assessment of core protection and monitoring systems of an integral reactor,SMART

The Korea Atomic Energy Research Institute has developed the SMART integral reactor, and SCOPS and SCOMS were also newly developed as advanced real-time core protection and monitoring systems for SMART. SCOPS calculates the minimum DNBR and maximum LPD based on several on-line measured core state parameters, and SCOMS calculates the limiting conditions for operation variables and assists the operator in implementing the technical specification requirements for monitoring. The design features and characteristics of SCOPS and SCOMS were described. The performance of the SCOMS power distribution synthesis method was evaluated and shows negligible power distribution synthesis errors. A technically reliable uncertainty analysis method was developed, and a preliminary uncertainty analysis was evaluated. The overall analysis results are similar or more improved compared to those of cycle 1 for Younggwang units 3&4 of Korea. In particular, uncertainty factors of SCOMS are much improved because of an improvement in the power distribution synthesis and DNBR calculation algorithm. Finally, thermal margins were estimated, and the DNB overpower margin of SCOMS is large enough to accommodate a 40% required overpower margin and 15% top-tier requirement thermal margin.     

压水堆核电厂堆芯功率能力验证分析

介绍了压水堆核电厂换料堆芯功率能力验证分析的原理和方法。利用中子学计算程序对换料堆芯正常运行工况（一类工况或工况I）和中等频率事故工况（二类工况或工况II）中可能的堆芯功率分别进行模拟。从反应堆物理和热工水力学的角度论证反映一、二类工况堆芯安全性的线功率密度裕量和偏离泡核沸腾比（DNBR）裕量。从而验证一类工况反应堆运行区域和二类工况超漏、超功率保护限值。本文还给出了大亚湾核电站18个月换料堆芯功率能力验证分析的结果。     

部分参数统计方法在30万kW核电厂应用的初步研究

偏离泡核沸腾(DNB)设计基准是反应堆热工水力设计中的重要基准之一,为评价该设计基准是否满足热工水力设计要求,需确定堆芯偏离泡核沸腾比(DNBR)设计限值。本文研究了使用统计学方法确定不确定性的部分参数统计方法原理,并应用该方法和堆芯子通道分析程序对30万kW核电厂DNBR设计限值进行计算。计算结果表明,反应堆冷却剂流量全部丧失事故最小DNBR分析采用部分参数统计较STDP额外获得了约5%的裕量。本文结果为DNBR设计基准的验证提供了关键判据。     

压水堆硼稀释事故的计算

一、概述核电厂压水堆的初始过剩反应性是通过控制捧、固体可燃毒物和溶解在主回路冷却剂中的硼酸等三种方式联合控制的。随着反应堆的燃耗和裂变毒物的积累,堆的过剩反应性不断减少,需要通过化容系统,控制硼酸的浓度来进行补偿。如果化容系统发生故障或操作员误操作,就会给回路注入无硼或低于规定浓度的补给水,给反应堆引入正反应性,造成硼稀释事故。按核安全法规的要求,应对换料、冷停堆、热停堆和启动等工况进行计算,并要求在这些     

Neutronics analysis of megawatt-class gas-cooled space nuclear reactor design

ABSTRACT

Neutronics analysis was conducted for a proposed megawatt-class gas cooled space nuclear reactor design. The reactor design has a high operating temperature of up to 1500 K. Annular UO₂ fuel rods were used to reduce the central temperature of the fuel. The thermal power is 2.3 MWt and is converted into electric power by a direct Brayton cycle. The control rods were arranged in different configurations and were analyzed in order to evaluate the influence on the reactor design in different scenarios. The calculation results reveal that the control rods arrangements have influences on the begin-of-life (BOL) excess reactivity and the shutdown reactivity. The distribution of control rods affects the neutron economy and leakage in the fuel region, consequently affecting the reactivity. It is also known that the reactivity in flooding scenarios are not sensitive to different control rod arrangements. Meanwhile, according to calculation results, the proposed reactor design has enough shutdown reactivity margin which will allow for flexible control strategy. Further analysis is still needed for more detailed and accurate parameters of the reactor design.     

A methodology for benefit assessment of using in-core neutron detector signals in core protection calculator system (CPCS) for Korea standard nuclear power plants (KSNPP)

Core Protection Calculator System (CPCS) is a digital computer based safety system generating trip signals based on the calculation of departure from nucleate boiling ratio (DNBR) and local power density (LPD). Currently, CPCS uses ex-core detector signals for core power calculation and it has some uncertainties. In this work, a quantitative economic benefit assessment of using in-core neutron detector signals is carried out. In-core detector signals which directly measure the inside neutron flux of core are applied to CPCS to obtain more accurate power distribution profile, DNBR and LPD to reduce the calculation uncertainties. In order to improve axial power distribution calculation, piecewise cubic spline method is applied. Simulation is also carried out to verify its applicability to power distribution calculation in this work. Simulation result shows that the improved method reduces the calculational uncertainties significantly and it allows larger operational margin. It is also assured that no power reduction is required while Core Operating Limit Supervisory System (COLSS) is out-of-service when the improved method is applied.     

应用抽样统计方法计算DNBR限值

应用根据Wilks公式发展的抽样统计方法及VIPRE-W程序计算DNBR和出口含汽率的参数不确定性,再结合模型与DNB关系式不确定性,得到DNBR的设计限值为1.220。与RTDP方法得到的DNBR限值进行比较,得出此方法能得到更多的DNBR裕度。     

Measurement of Shutdown Margin

A method was developed and investigated for directly measuring the shutdown margin in a reactor operating at a subcritical steady state. Measurements were made on the Iowa State University UTR-10 reactor with conditions ranging from almost critical to fully shutdown. The method is based on the fact that the frequency dependent transfer function of the reactor is a measure of the response of the reactor to an input disturbance. The magnitude of the transfer function of a subcritical reactor depends on the frequency of the disturbance and the steady-state multiplication factor. Since the multiplication factor is a measure of the reactivity, the amplitude of the transfer function is a direct indication of the shutdown margin. A sinusoidal disturbance was introduced by means of an oscillator operating in the center of the core. The oscillator was operated at a fixed frequency of 10 cps in order to eliminate the frequency dependence. The fluctuation was measured by a BF3 detector located against the outside of the core opposite the oscillator. The signal emitted by the detector contained bombardment noise and instrument noise in addition to the desired information. The signal was passed through a band-pass filter and was then cross-correlated with a pure sine wave to obtain the transfer function. A plot of the transfer function vs. the multiplication factor was a smooth function from which the multiplication factor could be found with a confidence of ± 0.04% near critical and with a confidence of ± 0.17% in the fully shutdown condition.