超临界水堆堆芯典型瞬态三维核热耦合分析

采用超临界水堆堆芯三维核热耦合瞬态性能分析方法,研究中国百万千瓦级超临界水堆(CSR1000)在控制棒弹出堆芯、控制棒失控抽出等典型瞬态过程中堆芯的瞬态性能。堆芯三维瞬态分析表明:控制棒弹出堆芯事故过程中燃料最大包壳壁面温度峰值低于事故安全限值(1260℃),控制棒失控抽出瞬态过程中燃料最大包壳壁面温度峰值低于瞬态安全限值(850℃)。燃料温度和水密度的显著反应性反馈以及必要的保护停堆措施,能够保证CSR1000堆芯在典型瞬态过程中的安全性能。     

超临界水堆CSR1000堆芯设计优化

利用开发的超临界水堆(SCWR)堆芯稳态性能分析程序SNTA,研究分析中国百万千瓦级SCWR(CSR1000)优化堆芯燃耗性能、反应性控制能力、功率分布、最大燃料包壳温度和最大线功率密度等稳态性能,并给出与组件功率相匹配的第II流程冷却剂流量分配方案。研究表明,采用本文所述燃料组件及堆芯设计优化方法,可以有效延长堆芯燃耗寿期。     

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。     

落棒法测量所有控制棒全插时的棒价值

商用压水堆控制棒总价值没有完全通过试验来验证。利用反应堆功率运行时落棒停堆瞬态数据,用自己开发的离线反应性仪计算程序(ODRM)来计算所有控制棒全插(ARI)时的棒价值,验证了核设计。     

超临界水冷堆CSR1000反应性控制方法研究

超临界水冷堆完全依靠可燃毒物及控制棒进行反应性控制,因而可燃毒物布置方案及控制棒管理方案是其堆芯设计的关键。通过燃料组件反应性计算分析,本文选取Er2O3作为与UO2燃料混合的可燃毒物,以及与沸水堆类似的十字形控制棒,然后利用三维堆芯物理热工耦合计算方法,进行控制棒管理方案设计,建立满足总体及安全性设计要求的超临界水冷堆CSR1000平衡循环堆芯,并对堆芯关键设计参数进行评价。     

CSR150反应性控制研究

超临界水冷技术示范堆(CSR150)借鉴了中国超临界水冷堆(CSR1000)的反应性控制方式，依靠可燃毒物及控制棒进行反应性控制。本文在优选Er₂O₃作为可燃毒物的基础上，提出采用富集167Er的设计方式，以降低寿期末Er₂O₃带来的反应性惩罚。对CSR150的控制棒设计进行研究，提出控制棒分区设计方案，采用富集硼作为安全棒吸收体材料，提升堆芯反应性控制能力。通过对堆芯关键设计参数进行分析评价可知，本文提出的反应性控制方案满足CSR150设计要求。     

紧急停堆棒落棒时间对熔盐堆反应性引入瞬态的影响

紧急停堆棒落棒时间是影响反应堆安全特性的重要参数,以2 MW钍基熔盐堆为研究对象,采用RELAP5-TMSR(Reactor Excursion and Leak Analysis Program-Thorium Molten Salt Reactor)程序,建立熔盐堆系统的瞬态行为分析模型,对控制棒提棒速度的敏感性进行分析,并重点分析探索紧急停堆棒落棒时间对熔盐堆反应性引入瞬态后果的影响规律。结果表明:即使紧急停堆棒落棒时间达到10 min,哈氏合金的最高温度也仅为708.2°C,燃料盐最高温度为709.2°C,均低于安全允许限值,表明该熔盐堆具有良好的应对反应性引入事件的能力。     

SCWR堆芯三维瞬态物理-热工水力耦合程序开发

耦合三维中子时空动力学程序和超临界水堆(SCWR)热工水力计算程序,开发了适用于SCWR堆芯瞬态和事故分析的三维瞬态分析程序STTA。采用第二类边界条件节块格林函数方法 NGFMN_K求解瞬态中子扩散方程,采用串行耦合方法将SCWR子通道程序ATHAS嵌入NGFMN_K程序。通过压水堆基准题NEACRP-L-335和SCWR弹棒问题检验STTA程序,结果表明:STTA针对压水堆问题的计算结果与参考解符合良好,针对SCWR问题的计算结果合理可信,可用于SCWR堆芯的三维瞬态性能分析。     

超临界压水堆改进型控制棒组件落棒过程计算分析研究

为验证超临界压水堆改进型控制棒组件能否实现预期水力缓冲功能,采用计算流体力学分析软件Fluent、基于6自由度（6DOF）模型的铺层法动网格技术,对其落棒过程进行研究,分析了控制棒组件落棒时间和落棒末速度。结果表明：相比改进前的设计,改进型控制棒组件落棒时间虽有所增大,但仍然能满足安全要求;落棒末速度大幅下降,落棒冲击力降低,从而能够保证控制棒组件及燃料组件的结构完整性。改进型控制棒组件的设计能够实现预期的水力缓冲功能,可用于超临界压水堆堆芯设计。     

SCWR堆芯三维稳态性能分析程序系统开发及验证

基于节块法中子扩散计算程序,二次开发了具备调棒临界-燃耗计算及燃料管理能力的超临界水堆(SCWR)堆芯稳态中子学计算程序NGFMN_S。通过模块化方式耦合NGFMN_S和超临界水堆子通道热工-水力计算程序ATHAS,开发了超临界水堆堆芯三维物理-热工水力耦合稳态性能分析程序SNTA。针对超临界水堆堆芯CSR1000,通过与耦合程序CASIR及SRAC/SPROD对比检验,结果表明:SNTA程序针对CSR1000问题的计算结果与参考程序符合良好;相比于堆芯计算采用细网有限差分方法的CASIR或SRAC/SPROD程序,SNTA程序的计算效率显著提高;适用于具备强烈核热耦合特性的超临界水堆堆芯的稳态性能分析。     

AP1000自动降压系统误开启事故仿真分析

针对AP1000的具体结构和运行特点,采用FORTRAN程序设计语言,开发了AP1000瞬态热工水力计算程序RETAC。利用RETAC对AP1000自动降压系统(ADS)误开启事故进行仿真分析,得到稳压器压力、堆芯归一化热功率、堆芯归一化流量、堆芯平均温度、燃料中心最高温度和最小偏离核态沸腾比（MDNBR）等主要系统参数的响应特性。分析结果表明,在稳压器低压停堆保护的作用下,燃料中心最高温度和MDNBR未超出规定限值,满足安全准则要求。并将计算结果与美国西屋公司AP1000分析软件LOFTRAN的计算结果进行对比,对比趋势符合良好,证明了RETAC建模和自动降压系统临界流模型计算的合理性。     

Assessment of the MDNBR enhancement methodologies for the SMART control rods banks withdrawal event

For an electricity generation and seawater desalination, a 330 MW System-integrated Modular Advanced ReacTor (SMART) was developed by KAERI. The safety level of the SMART is enhanced when compared to that of the typical commercial reactors, with the aid of an elimination of a large break loss of coolant accident by placing the major components of the primary system in a reactor vessel and the adoption of a new technology and a passive design concept into the safety system. However, the events related to reactivity and power distribution anomalies have been evaluated as vulnerable points when compared to the other initiating events in the SMART, since the reactivity worth of the control rods (CR) banks is quite large due to the boron free core concept. Especially, safety margins, i.e., minimum departure from nucleate boiling ratio (MDNBR), are significantly threatened during the CR banks withdrawal event. Therefore, MDNBR enhancement methodology for the CR banks withdrawal event should be considered to further enhance the safety level of the SMART design. Two methodologies have been suggested to enhance the MDNBR during the CR banks withdrawal event: the application of a DNBR trip function into a core protection system and a turbine trip delay methodology. Sensitivity studies are performed to evaluate the two MDNBR enhancement methodologies and show that the suggested methodologies could enhance the MDNBR during the CR banks withdrawal event of the SMART.     

Developing and modeling of the “Laguna Verde” BWR CRDA benchmark

Reactivity initiated accidents (RIA) and design basis transients are one of the most important aspects related to nuclear power reactor safety. These events are re-evaluated whenever core alterations (modifications) are made as part of the nuclear safety analysis performed to a new design. These modifications usually include, but are not limited to, power upgrades, longer cycles, new fuel assembly and control rod designs, etc. The results obtained are compared with pre-established bounding analysis values to see if the new core design fulfills the requirements of safety constraints imposed on the design. The control rod drop accident (CRDA) is the design basis transient for the reactivity events of BWR technology. The CRDA is a very localized event depending on the control rod insertion position and the fuel assemblies surrounding the control rod falling from the core. A numerical benchmark was developed based on the CRDA RIA design basis accident to further asses the performance of coupled 3D neutron kinetics/thermal-hydraulics codes. The CRDA in a BWR is a mostly neutronic driven event. This benchmark is based on a real operating nuclear power plant — unit 1 of the Laguna Verde (LV1) nuclear power plant (NPP). The definition of the benchmark is presented briefly together with the benchmark specifications. Some of the cross-sections were modified in order to make the maximum control rod worth greater than one dollar. The transient is initiated at steady-state by dropping the control rod with maximum worth at full speed. The “Laguna Verde” (LV1) BWR CRDA transient benchmark is calculated using two coupled codes: TRAC-BF1/NEM and TRAC-BF1/ENTRÉE. Neutron kinetics and thermal hydraulics models were developed for both codes. Comparison of the obtained results is presented along with some discussion of the sensitivity of results to some modeling assumptions.     

Monte Carlo simulation of a research reactor with nominal power of 7 MW to design new control safety rods

The Monte Carlo simulation has been established for a research reactor with nominal power of 7 MW. A detailed model of the reactor core was employed including standard and control fuel elements, reflectors, irradiation channels, control rods, reactor pool and thermal column. The following physical parameters of reactor core were calculated for the present LEU core: core reactivity (ρ), control rod (CR) worth, thermal and epithermal neutron flux distributions, shutdown margin and delayed neutron fraction. Reduction of unfavorable effects of blockage probability of control safety rod (CSR)s in their interiors because of not enough space in their sites, and lack of suitable capabilities to fabricate very thin plates for CSR cladding, is the main aim of the present study. Making the absorber rod thinner and CSR cladding thicker by introducing a better blackness absorbing material and a new stainless steel alloy, respectively, are two studied ways to reduce the effects of mentioned problems.     

基于MCNP程序的海洋核动力平台堆芯核设计校核计算研究

海洋核动力平台是小型核反应堆与船舶工程技术的有机结合,具有机动性好、一次性装料运行周期长、功率密度大、运行成本低、节能环保等特点。本文采用蒙特卡罗粒子输运程序(MCNP),建立海洋核动力平台反应堆堆芯几何模型,计算该反应堆首循环初始装料冷态、常压下的堆芯反应性和控制棒价值,并与核设计计算结果进行对比。结果表明:MCNP程序适用于海洋核动力平台反应堆堆芯核设计校核计算,并可与核设计值互相验证。     

CSR1000堆芯流量分配的研究

以中国百万千瓦级超临界水冷堆(CSR1000)堆芯为研究对象,建立热工水力计算模型,计算出冷却剂和慢化剂温度分布、堆芯功率分布、燃料组件出口压力及流量分配等参数。计算结果表明,适当增加堆芯内部燃料组件流量比例,可以有利于径向功率展平,内外燃料组件通道出口压降,呈现"N"型变化,增大内部燃料组件的堆芯入口功率,内部组件内的流量分配也将减少,而外部燃料组件通道中的流量将增加,适当调整堆芯入口流量初始分配比例,可以使各通道功率分布展平。     

An assessment of a VVER-1000 core during Turbo-Generator load reduction test using RELAP5/MOD3.2 and WIMSD-5B/PARCSv2.7

Reactor dynamic tests, which are categorized as one of the power start-up test groups, are the most complex tests during commissioning of the new nuclear power plants. This paper presents the results of Turbo-Generator load reduction test as one of the reactor dynamic tests for VVER-1000/V446 unit at Bushehr Nuclear Power Plant (BNPP). In this test modeling because of the need for control rod bank worth and core reactivity coefficients, the core geometry has been modeled first by using WIMSD-5B/PARCSv2.7 codes for neutronic calculations. For performing the thermal-hydraulic analysis, the RELAP5/MOD3.2 computer code has been used. The control rod bank worth and core reactivity coefficients obtained from WIMSD-5B/PARCSv2.7 are compared with BNPP FSAR that confirm the ability and reliability of the method. Also comparison of the thermal-hydraulic core parameters obtained from RELAP5/MOD3.2 against actual plant data, indicate that this code can properly predict behavior of VVER-1000 reactor for this dynamic start-up test.     

动态棒价值测量软件包LIGHT的开发与数值验证

动态棒价值测量是一种快速测量控制棒组价值的方法。基于测量过程和相关的反应堆物理数值计算方法，开发了动态棒价值测量软件包LIGHT。LIGHT可产生进行动态棒价值测量所需的参数，包括静态空间因子、动态空间因子和缓发中子参数。针对基准问题和AP1000核电厂进行了数值计算并进行了比较。分析表明，计算结果具有较高的精度，说明建立的计算模型及开发的程序是正确的。     

钠冷快堆停堆保护方案研究

在借鉴中国实验快堆(CEFR)热工模型建模经验的基础上,利用Relap5程序建立霞浦示范快堆(CFR)的主要系统模型,并参考快堆安全分析中的预期瞬态无停堆保护(ATWS)的分析方法,对发生反应性意外引入事故时的安全裕度和停堆保护进行仿真研究。仿真结果表明,额定功率下发生反应性引入时,不会触发短周期的报警和停堆;当发生补偿棒失控提升5 s和10 s时的反应性意外引入事故,目前一回路保护参数整定值、信号测量延迟及安全棒落棒时间可以取其他值;当补偿棒失控提升15 s时,在目前的设计下,核功率和功率流量比信号能确保事故下的反应堆状态符合事故验收准则。当其他保护信号失效,堆芯出口钠温所触发的停堆保护若要实现同样的功能,则需保证反应堆在14.85 s之前进入深度次临界。     

钠冷快堆反应性意外引入事故时停堆保护研究

在反应堆系统中,当反应堆处于异常工况时,如果运行参数超出保护限值,则由保护系统触发相关保护动作,以保证反应堆的状态符合事故验收准则的要求。本文将通过Simulink建立钠冷快堆主要系统模型,在发生反应性意外引入事故时,借鉴快堆事故分析中预期瞬态无停堆保护（ATWS）的分析方法,基于相应保护参数的测量误差和数据处理过程对反应堆一回路的保护参数及其整定值进行研究,并确保钠冷快堆的状态在整个反应性引入事故过程中符合钠冷快堆的事故验收准则。仿真结果表明,当发生补偿棒失控提升5 s和10 s时,目前的堆芯出口钠温、功率、功率流量比等保护参数的整定值、信号测量延迟及落棒时间可取其他值。当补偿棒失控提升15 s时,只要保证保护参数整定值、相应参数的信号测量延迟及落棒时间能使反应堆在36.45 s前进入深度次临界都是可以的。