SOP堆芯次临界状态监测问题研究

对事故后一些物理现象下的堆外中间量程测量通道IRC(Intermediate Range Channel)的响应进行了分析,包括停堆后(N-1)组控制棒组件下落、反应堆水池淹没、堆芯空泡份额或堆芯裸露等。分析发现,事故工况下IRC的响应并不一定反映堆芯次临界度的变化。从状态导向事故运行程序SOP(State Oriented Procedure)堆芯次临界度监测的目的出发,对SOP中堆芯次临界度监测阈值的选取和一回路状态功能监测优先级的确定给出了建议。     

ADS次临界堆堆芯功率控制策略硏究

ADS次临界堆的功率控制与临界堆有本质的不同,临界堆大多以控制棒为控制手段来调节功率,而ADS次临界堆以控制棒和外中子源作为控制手段来调节功率.因此ADS系统的控制更为复杂且具有特殊性,有必要确定ADS次临界堆的控制策略.将控制棒反应性和外中子源强度同时设为受控对象,针对不同的堆芯相对功率偏差下,选择不同的受控对象,实...     

反应堆科学技术——医用同位素生产堆零功率物理试验研究进展

2007年初，中国核动力设计研究院委托中国原子能科学研究院进行医用同位素生产堆技术研究。经近10个月的紧张工作，对YSR铀溶液临界装置（图1）进行堆芯容器的更换及相关设施的技术改进，为今后的一系列物理实验做准备。实验内容包括：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、空泡系数测量。     

脉冲堆反应性测量系统的研制

研制了基于脉冲堆核测量系统探测器的数字反应性仪。选择适合该堆动态反应性实时监测的数值计算和数字滤波方法,研究了落棒法和插棒法测量控制棒价值的空间效应的修正方法。经过在脉冲堆上测量验证,结果令人满意。     

基于凤凰快堆寿期末控制棒提棒实验的SARAX程序系统确认

寿期末控制棒提棒实验是在法国钠冷快堆Phenix(凤凰快堆)退役之前开展的最后一次实堆测量实验,实验中测量了低功率状态下的控制棒价值和满功率状态下的径向功率分布。本实验采用西安交通大学开发的快堆中子学计算程序系统SARAX进行建模和计算,其计算过程采用基于点截面的超细群方法进行能谱计算,采用超级均匀化(SPH)因子方法进行组件均匀化计算,以及采用多群中子输运节块方法进行堆芯计算,最终计算了实验中4个临界状态的有效增殖因子、控制棒价值、堆芯反应性系数及功率分布等参数。计算结果表明:SARAX的计算结果与实验值吻合较好,计算精度优于传统的快堆物理计算程序,可以用于钠冷氧化物混合燃料(MOX燃料)快堆的核设计。     

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。     

DF-VI快中子临界装置2号堆芯临界实验

DF-VI快中子临界装置在改造完成、堆芯发生了变化以后,进行了重新启动和一系列的实验测量。测量内容有：根据29次临界实验的数据对2号堆芯平均临界元件数和临界质量进行了计算;应用周期法和棒补偿法对控制棒价值进行了刻度;用逆动态反应性计对安全棒和安全块的价值进行了测量;对单根边缘元件价值和径向元件价值分布进行子测量。通过以上实验测量,确定了DF-VI快中子临界装置2号堆芯的主要安全运行参数。     

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。     

插棒法空间效应修正因子计算中MC方法的应用

研究了MC方法在插棒法空间效应修正因子计算中的应用。用MCNP／4B计算了脉冲堆稳态堆芯第一循环插棒前后探测器位置的中子注量率,得到了探测器位置的空问因子。设计了修正空间效应的方法,并用该方法测量了脉冲堆所有控制棒的积分价值,结果令人满意。     

西安脉冲堆反应性测量空间效应修正方法研究

反应性是反应堆重要的物理参数,在西安脉冲堆上增加反应性实时监测功能,能够为操纵员提供反应性实时大小和变化趋势,有利于其安全运行。逆动态方法以其实时性好、能测量任意反应性引入而被广泛应用于反应性测量,但由于控制棒的移动会导致中子注量率空间分布前后不一致,而出现偏差。本文对逆动态方法和静态空间效应因子进行了理论分析,给出了相应的计算方法;以西安脉冲堆为研究对象,使用蒙特卡罗（MCNP）程序计算了探测器三维空间的响应函数,同时计算了归一化节块功率密度,由此得到静态空间效应因子;最后在脉冲堆上进行了不同棒速下插控制棒实验,处理实验数据得到控制棒积分价值曲线。结果表明,进行空间效应修正是必要的,经修正后计算得到的控制棒价值曲线更稳定,计算结果与真实值误差更小。      

先进反应性测量系统的研发与应用

相比于传统的反应堆控制棒价值测量方法,快速的动态棒价值测量方法要求反应性测量设备具有更高的精度和性能,以准确获取和处理堆外探测器的电流信号,并需通过额外的堆芯中子学计算对试验过程中的空间效应进行修正。为此本研究开发了一套包含先进物理试验测量仪（APTC）和动态棒价值测量软件包（LIGHT）的先进反应性测量系统（SMART）,并对SMART开展了一系列验证试验。结果表明,SMART具备完整的物理启动试验功能,其精度和性能能够满足包括动态棒价值测量在内的物理启动试验的要求;在300 MW压水堆核电厂中的成功应用也充分验证了SMART的工程应用能力。      

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。     

Parametric tests and measurements after shimming of a beryllium reflector in a miniature neutron source reactor (MNSR)

After 10 years operation of Pakistan research reactor-2 (PARR-2), a miniature neutron source reactor (MNSR), a beryllium reflector was added to compensate the loss of reactivity due to burn up of fuel. Beryllium shim plates have been placed at the top of the core in a tray provided for this purpose. The control rod was dismantled and withdrawn from the core and the reactor was made subcritical with cadmium shimming. To monitor the neutron population during this experiment, two additional neutron monitoring channels based on BF₃ were installed around the core. Measurement of important Parameters such as effective delayed neutron fraction, decay constant, excess reactivity, control rod worth, temperature coefficient of reactivity, thermal neutron flux, cadmium ratio was done after the addition of Be reflector. Increase in reactivity worth due to addition of Be shim was 1.0 mk.     

快堆控制棒组件非均匀效应修正方法研究

控制棒组件是快堆控制系统和安全保护系统的重要组成部分,快堆控制棒价值的准确求解至关重要。基于PASC?5程序的快堆少群均匀化群常数计算中使用直接体积均匀化方式,这会导致控制棒价值严重高估,必须对控制棒组件的非均匀效应进行修正。基于群常数修正的思路,本论研究了体积?通量权重、反应率之比守恒和反应性守恒3种方法在快堆控制棒组件非均匀效应修正中的应用;基于二维特征线程序开发了群常数修正因子计算程序FRHP。通过中国实验快堆算例进行测试验证,修正后的控制棒价值计算结果与MCNP计算的参考结果符合较好,表明3种方法均能对控制棒组件的非均匀效应实现有效修正,其中反应性守恒方法修正效果最好。     

基于CORCA的带固定中子源堆芯求解与共轭计算的软件实现

深度次临界状态下,传统源倍增法在核反应堆反应性测量上具有精度低的特点,为提高测量精度,本文对CORCA软件进行扩充,开发了具备固定源问题求解和带不连续因子中子价值求解功能的CORCAFIX软件,并采用对照程序和实堆数据对CORCA-FIX软件进行了计算验证。验证结果证实,CORCA-FIX在求解带固定源堆芯的深度次临界状态时有着较高的精度,输出的结果应用于实堆数据后获得了更好的次临界度测量结果,且满足工程应用中反应性测量的偏差准则。     

Dependence of control rod worth on fuel burnup

One important parameter in the design and the analysis of a nuclear reactor core is the reactivity worth of the control rods, i.e. their efficiency to absorb excess reactivity. The control rod worth is affected by parameters such as the fuel burnup in the rod vicinity, the Xe concentration in the core, the operational time of the rod and its position in the core. In the present work, two different computational approaches, a deterministic and a stochastic one, were used for the determination of the rods worth dependence on the fuel burnup level and the Xe concentration level in a conceptual, symmetric reactor core, based on the MTR fuel assemblies used in the Greek Research Reactor (GRR-1). For the deterministic approach the neutronics code system composed by the SCALE modules NITAWL and XSDRN and the diffusion code CITATION was used, while for the stochastic one the Monte Carlo code TRIPOLI was applied. The study showed that when Xe is present in the core, the rods worth is significantly reduced, while the rod worth variation with increasing burnup depends on the rods position in the core grid. The rod worth obtained with the use of the Monte Carlo code is higher than the one obtained from the deterministic code.     

Void reactivity evaluation by modified conversion ratio measurements in LWR critical experiments

We have developed a void reactivity evaluation method by using modified conversion ratio measurements in a light water reactor (LWR) critical lattice. Assembly-wise void reactivity is evaluated from the “finite neutron multiplication factor”, k*, deduced from the modified conversion ratio of each fuel rod. The distributions of modified conversion ratio and k* on a reduced-moderation LWR lattice, for which the improvement of negative void reactivity is a serious issue, were measured. Measured values were analyzed with a continuous-energy Monte Carlo method. The measurements and analyses agreed within the measurement uncertainty.

The developed method is useful for validating the nuclear design methodology concerning void reactivity.     

A Reactivity Meter and Its Application

In order to measure reactivity from instant to instant during reactor operation, real time computation based on reactor kinetic equations must be made.

The authors developed some reactivity meters capable of such computation based on the analog computer technique. After their characteristics were checked by an analog computer and a kinetic simulator, they were applied to reactor experiments and operation. It was found that the reactivity meter method for measurement of reactivity worth (especially for control rod calibration) is better than other current methods, and the reactivity meter can also serve as a tool in nuclear instrumentation of reactors for its ability to indicate reactivity even in reactor start up conditions.     

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.