CAP1400控制棒提升极限分析方法研究

控制棒提升极限用于限定控制棒组棒位和可溶硼浓度的范围,以防止慢化剂温度系数突破限值。CAP1400核电厂采用机械补偿运行策略,使控制棒及硼浓度运行范围大为扩展,同时功能独立的M棒组和AO棒组同时插入堆芯使得插棒情况更为复杂,因此与传统核电厂相比,CAP1400核电厂的控制棒提升极限更难界定。本文建立了适用于CAP1400核电厂的控制棒提升极限分析方法,并给出计算结果。本文提出的方法合理地解决了复杂的控制棒运行情况给提升极限造成的影响,并充分地利用了电厂实测数据对提升极限进行修正。基于本文方法得到的提升极限精确且具备一定的保守性,所以便于电厂实际运行时使用。     

混合能谱超临界水冷堆反应性引入事故安全分析

本文研究了混合能谱超临界水冷堆(SCWR-M)在发生控制棒失控提升事故和弹棒事故这两类反应性引入事故后的反应堆系统响应。首先利用修改的可用于超临界条件下的系统程序RELAP5对混合能谱超临界水冷堆进行系统建模,并计算分析在功率运行工况下事故过程中功率、流量及包壳温度等重要参数的变化趋势,最后对反应性参数如控制棒价值、控制棒抽出速率和负反馈系数进行了参数效应分析。结果表明,在设计工况下混合能谱超临界水冷堆系统可有效地将衰变热导出堆芯,保证了燃料棒的完整性。另外,反应性参数对控制棒失控提升事故的安全性影响不大,但对弹棒事故的包壳峰值温度影响很大,过于保守的反应性参数估计会使安全裕量大为减小。     

CANDU6重水堆37R燃料和37M燃料的反应性比较

37R燃料的每根元件尺寸相同,中心元件的冷却剂流道面积较小,事故工况下热工裕量相对较小。37M燃料减小中心元件尺寸,从而增大中心元件和整个燃料棒束的热工裕量。本文从反应堆物理角度定量分析两种燃料的反应性差异,采用WIMS程序和RFSP程序,计算了温度系数、空泡系数、重水纯度和慢化剂毒物浓度变化导致的反应性变化。计算结果表明37R燃料和37M燃料的反应性系数差别很小。     

秦山核电站物理启动试验

介绍了秦山核电公司300MW压水堆核电站首次物理启动试验,包括首次临界和低功率物理试验。测试项目包括临界硼浓度,控制棒价值,硼价值,功率分布,慢化剂温度系数,最小停堆硼浓度,弹棒束价值等。试验结果表明,各项参数满足了堆芯设计和安全上的要求。     

用伪随机振荡法测量反应堆动态特性

本文介绍利用伪随机双值序列(PRBS)作为激励信号,驱动反应堆控制棒,引入小反应性扰动,测量反应堆有功率下的频率响应。在小型数字计算机上作相关分析,再利用简化的理论模型拟合实测的频响曲线,从而获得燃料和慢化剂的反应性温度系数及其时间常数等有意义的动态参数。     

CSR1000控制棒落入堆芯三维瞬态分析

针对超临界水堆(SCWR)控制棒落入堆芯事件特点,采用堆芯三维瞬态性能分析方法,利用开发的SCWR堆芯三维瞬态物理-热工水力耦合程序STTA,建立SCWR堆芯落棒瞬态三维计算模型和分析流程,研究分析超临界水堆CSR1000在控制棒落入堆芯瞬态过程中的堆芯性能,分析评价落棒瞬态下CSR1000堆芯的安全性能。堆芯三维落棒瞬态分析表明,当落入堆芯棒束价值较高时,落棒初期堆芯功率下降较快,之后由于水密度的反应性反馈,堆芯功率缓慢回升至新的平衡,堆芯功率下降速率超过了停堆信号整定值,将触发保护停堆;当落入堆芯棒束价值较低时,由于水密度的反应性反馈,堆芯功率下降缓慢,堆芯功率下降速率未能达到停堆信号整定值,不能触发保护停堆。控制棒落入堆芯对堆芯轴向功率分布影响很小,高价值落棒导致的落棒区域燃料组件功率坍塌相对低价值落棒更明显。无论是高价值落棒还是低价值落棒,瞬态过程中最大包壳壁面温度均低于瞬态安全限值850℃。水密度的显著反应性反馈及必要的保护停堆措施能保证CSR1000堆芯在控制棒落入堆芯过程中的安全性能。     

固定棒位法测量控制棒总价值

控制棒价值测量的准确度与效率对核电厂的安全性与经济性具有重要影响。在动态刻棒等反应性测量工作中,本底与中子源对探测器有显著影响,致使根据实测电流计算得到的反应性显著偏离真实值。基于点堆逆动态方程,通过对本底与中子源影响的分析,利用固定棒位状态下的测量数据计算反应性并得到控制棒总价值,给出了一种不受本底与中子源影响的简便的控制棒总价值测量计算方法,并在零功率实验装置上进行验证。结果表明,该方法可有效避免本底和中子源组件对反应性探测的影响,并简化了离线理论计算,其与周期法计算结果的相对偏差在1%以内。     

粗网节块法中控制棒尖端效应研究

如何处理控制棒部分插入节块情况、避免出现控制棒尖端效应是采用粗网方法计算控制棒微分价值时的一个重要问题。本文在相关研究工作的基础上,研究了该问题的一种处理方法。该方法通过对插棒组件进行附加的一维轴向细网计算来产生部分插棒节块的等效均匀化截面,并对部分插棒节块应用轴向不连续因子,数值验证结果表明,该方法即使在不进行三维-一维耦合迭代计算的情况下,仍然可以有效地消减控制棒尖端效应。     

基于PAnySimu的PWR核电站堆芯建模与仿真研究

基于PAnySimu仿真支撑系统对PWR核电站一回路堆芯部分进行建模与仿真分析.通过研究分析岭澳二期3/4号机组堆芯实际结构,将其分为功率计算、堆芯传递计算、控制棒引起的反应性、反应性反馈、毒物计算五个模型.在此基础上,分析堆芯中子通量,考虑控制棒位置、燃料和慢化剂温度、氙和钐中毒、硼浓度等因素对中子通量的影响.利用P...     

医用同位素生产堆零功率物理试验研究

中国原子能科学研究院于2007年初受中国核动力研究设计院的委托,为新型医用同位素生产堆（MIPR）技术研究进行了一系列零功率物理实验。实验包括：最小临界质量测量、控制棒微分、积分价值测量、临界棒位和后备反应性测量、停堆深度测量、温度系数测量、气泡效应测量。     

Benchmark Analysis of TRIGA Mark II Reactivity Experiment Using a Continuous Energy Monte Carlo Code MCNP

The benchmark analysis of reactivity experiments in the TRIGA-II core at the Musashi Institute of Technology Research Reactor (Musashi reactor, 100 kW) was performed by a three-dimensional continuous-energy Monte Carlo code MCNP4A. The reactivity worth and integral reactivity curves of the control rods as well as the reactivity worth distributions of fuel and graphite elements were used in the validation process of the physical model and neutron cross section data from the ENDF/B-V evaluation. The calculated values of integral reactivity curves of the control rods were in agreement with the experimental data obtained by the period method. The integral worth measured by the rod drop method was also consistent with the calculation. The calculated values of the fuel and the graphite element worth distributions were consistent with the measured ones within the statistical error estimates. These results showed that the exact core configuration including the control rod positions to reproduce the fission source distribution in the experiment must be introduced into the calculation core for obtaining the precise solution. It can be concluded that our simulation model of the TRIGA-II core is precise enough to reproduce the control rod worth, fuel and graphite elements reactivity worth distributions.     

Gd burnable poison system for reactivity control of the first cycle of a PWR

The reactivity control of a PWR core may be performed by a system of burnable poison (BP) rods. In such a case, the soluble B system may be eliminated and the BP rods will be responsible for the excess reactivity provided for fuel depletion and fission products accumulation. A strong negative moderator temperature coefficient is a desirable safety feature, inherent to a poison-free moderator. The design objective of a PWR core controlled completely by a system of BP rods is achieved by utilization of Gd as the poison material and annular geometry of a BP rod. The proposed concept is tested as a retrofittable option for the current generation, as well as new PWR plants. A plausible incore fuel-management scheme is demonstrated, with planar power distribution, close to an acceptable range. The fuel-cycle penalty due to the residual poison content at EOC is relatively small.     

RIA基准题验证堆芯三维物理-热工耦合程序

以COBRA-Ⅳ和NLSANMT程序为基础,开发了堆芯三维物理-热工耦合程序C4/NK。针对两个典型的反应性引入事故（RIA）,即NEACRP弹棒基准题和提棒基准题,分别进行了验证计算。与参考值和其他程序的计算结果相比,C4/NK耦合程序具有较好的精度,能正确模拟瞬态过程中的物理-热工反馈现象。     

CFR600核设计中控制棒价值非均匀效应的群常数修正方法研究

CFR600堆芯反应性控制和停堆仅使用控制棒,其价值计算的准确性对核设计至关重要。CFR600核设计计算中,组件使用直接体积均匀化,不考虑非均匀效应。但控制棒非均匀效应较强,需进行修正。本文研究控制棒非均匀效应的群常数修正方法,推导通量权重和反应性等效方法的理论计算公式;结合细网差分程序,开发完成群常数修正计算程序CREC;对CFR600安全棒和补偿棒的12群群常数进行修正计算研究,并验证了控制棒价值非均匀修正的计算结果。通量权重和反应性等效方法的计算结果与参考值吻合较好,此两种方法均可对控制棒价值非均匀效应进行有效修正。     

Moderator Temperature Coefficient in BWR Core

Temperature dependences of infinite multiplication factor k∞ and neutron leakage from the core must be examined for estimation of moderator temperature coefficient. Temperature dependence on k∞ has been investigated by many researchers, however, the dependence on neutron leakage of a BWR with cruciformed control rods has hardly been done. Because there are difficulties and necessity on calculations of three space dimensional and multi-energy groups neutron distribution in a BWR core.

In this study, moderator temperature coefficients of JPDR-II (BWR) core were obtained by calculation with DIFFUSION-ACE, which is newly developed three-dimensional multi- group computer code. The results were compared with experimental data measured from 20 to 275°C of the moderator temperature and the good agreement was obtained between calculation and measurement.

In order to evaluate neutron leakage from the core, the other two calculations were carried out, adjusting criticality by uniform absorption rate and by material buckling. The former underestimated neutron leakage and the latter overestimated it. Discussion on the results shows that in order to estimate the temperature coefficient of BWR, neutron leakage must be evaluated precisely, therefore the calculation at actual pattern of control rods is necessary.     

基于Dragon与Donjon程序的液态熔盐实验堆临界计算与分析

基于组件输运程序Dragon与堆芯节块法程序Donjon,对包含有上下熔盐腔室、控制棒、实验孔道与中子源孔道的液态熔盐实验堆堆芯进行了计算与分析,给出了液态熔盐实验堆不同组件的等效均匀化模型。根据液态熔盐实验堆特性将中子能群划分为5种少群能群结构,基于所划分的每一种少群能群结构,对单根控制棒与不同控制棒组插入堆芯后的有效增殖因数和控制棒价值进行了计算分析。结果表明,7群能群结构具有更好的计算结果。基于7群能群结构开展了堆芯径向与纵向功率分布,以及控制棒拔出后堆芯的温度反应性系数计算分析,其计算结果与MCNP5计算结果相近,证明了模型等效的合理性以及Dragon和Donjon程序对液态熔盐实验堆的适用性。     

3He辐照考验装置反应性引入事故分析

在高通量工程试验堆（HFETR）中,3He回路内气体压力变化会向反应堆引入反应性,进而影响到HFETR的运行安全。本文利用蒙特卡罗（MCNP）程序计算了3He辐照考验装置反应性变化速率,并利用RELAP5程序对3He屏失压与HFETR 1根控制棒失控提出叠加事故进行了分析。结果表明,正常工况下,3He回路辐照试验不影响HFETR 正常运行;3He屏失压事故与HFETR事故工况叠加不会影响HFETR安全。      

Efficient code system of core characteristics for small cylindrical thermal reactor

A new code system for the overall neutronic calculation of a thermal reactor by a simple and effective way is presented. The code covers microscopic library compilation, macroscopic constant generation, cell calculations by multi-group treatment for neutron transport equation and core calculations over three zones for fuel and one zone for moderator. The Dancoff correction factor required in the interpolation of the self-shielding factors of resonance nuclides is automatically calculated by the installed collision probability routines. The burn-up calculation and Garrison and Ross model of fission product have been included. Also the effect of control rod on the reactivity of the reactor with special treatment for the control rod based on the homogenization technique has been included. Making a comparison with SRAC95 code system has checked the adopted code.     

Interpretation of Spatial Kinetics for Rod Drop Experiment in Loosely Coupled-Core Systems

Abstract

The coupled two-core reactor systems with various degrees of spatial coupling were constructed in the Kyoto University Critical Assembly (KUCA) to study the spatial kinetics observed in the control rod drop experiment. By applying the two-mode and the two-point kinetic models to the space-dependent rod worths measured on the basis of the one-point model, the first-harmonic λ-mode eigenvalue separation and the reactivity coupling coefficient were inferred. The present values of these parameters agreed with the results obtained by the reactor noise measurements and the diffusion calculations.

The experimental results show that the magnitudes of the spatial kinetic phenomena including the dependence of the rod reactivity worth on the detector position, the reactivity interaction effect between control rods and the transient flux tilts induced by the rod drop, which have been significantly observed in large thermal and fast power reactors, are inversely proportional to the eigenvalue separation. Applying the two-mode model, the inherent reactivity worths of control rods were also inferred from the space-dependent ones.