破损燃料组件修复后的物理和热工计算分析

以采用AFA3G燃料组件的中国改进型三环路压水堆(CPR1000)核电机组为研究对象,对装入反应堆后的正常燃料组件和修复燃料组件的堆芯物理和热工性能进行分析评估。结果表明:燃料组件内更换1根燃料棒对燃料组件反应性的影响小于-0.03%,该影响可以忽略;修复的燃料组件在换棒位置周围的燃料棒相对功率略微升高约5.6%;燃料组件内更换1根不锈钢棒对燃料组件的相对功率影响约为0.1372%~0.2698%,对组件燃耗的影响大约为0.11%,对堆芯慢化剂温度系数的影响大约为0.03%,对组件出口慢化剂温度的影响大约为0.03%;对堆芯功率峰因子、堆芯临界硼浓度、堆芯停堆裕量和堆芯出口慢化剂温度基本没有影响。     

采用哑棒修复燃料组件的核特性分析方法及其工程应用

在核电厂正常运行过程中,由于一回路杂物的存在或燃料操作失误,出现了少量燃料棒损伤的情况,通过采用哑棒替换损伤燃料棒可修复损伤燃料组件并回堆使用,可避免降低核电厂运行经济性。本文通过模拟采用不锈钢和锆合金哑棒替换破损燃料棒对燃料组件进行修复,分析修复后燃料组件中子学特性及修复燃料组件对堆芯运行核特性参数的影响机理,评估采用哑棒修复燃料组件并回堆使用对堆芯运行安全的影响,对采用哑棒修复燃料组件建立了完整的核设计分析方法和流程。该方法对采用哑棒修复燃料组件的核设计分析具有广泛的适用性,对采用修复燃料组件的堆芯换料设计具有实际的指导意义。该分析方法和流程的建立在国内反应堆物理分析领域尚属首次,目前该技术已应用于恰希玛一期核电厂堆芯换料设计的工程实践。     

VVER燃料组件变形对象限功率倾斜的影响

本文以田湾核电厂VVER堆芯为参考对象,通过调整燃料棒栅距模拟组件弯曲变形,从而定量分析燃料组件的变形对象限功率倾斜的影响。研究表明：栅距变化引起的慢化效应和能谱变化引起的核素效应是导致组件反应性变化的关键因素,随燃耗加深而逐渐变形的燃料组件通过上述两个因素的共同作用来影响堆芯象限功率倾斜。     

CEFR组件替换反应性价值试验测量与计算分析

组件替换反应性价值定义为测量位置组件替换成相应组件时引入的反应性变化。中国实验快堆物理启动试验中组件替换反应性价值测量试验方案中,试验测量了8个典型位置,其中6个位置为燃料组件替换成不锈钢组件,另外两个为不锈钢组件替换成燃料组件。测量结果显示,燃料组件替换反应性价值由内至外依次减少,内圈燃料组件替换反应性价值约-980 pcm,外圈燃料组件替换反应性价值约-470 pcm,补偿棒棒组测量和单根补偿棒测量的结果差别微小。使用CITATION程序对试验方案进行了理论计算,结果表明,计算结果与实验值符合良好,检验了CITATION程序的工程设计实用性。     

燃料组件修复装置的设计

核电站反应堆燃料组件修复装置包括燃料棒更换装置、燃料棒定位小车和燃料篮倾翻装置3套设备。燃料篮倾翻装置能实现待检修燃料组件翻转的功能;燃料棒更换装置在燃料棒定位小车上能进行3个方向的移动,对准工位后对单根燃料棒进行检修。     

超临界水冷堆环形燃料组件核热耦合分析

在超临界水冷堆预概念设计中,组件设计是十分重要的,将影响堆芯性能。超临界水冷堆中水密度变化剧烈的特性要求必须进行核热耦合分析。从中子学及热工性能角度,使用三维核热耦合程序对环形燃料组件进行了优化设计。应用中子学计算程序FENNEL-N对环形燃料组件进行三维扩散计算,可得到组件内单棒功率分布,应用热工计算程序SUBSC对组件进行子通道分析。在计算过程中,分析了燃料棒间距及燃料棒与组件壁盒之间的间隙对组件性能的影响。计算结果显示,增大棒间距和棒壁间隙能提高组件k_inf,但会增大组件内功率峰因子;子通道受热不均匀性对组件热工性能影响较大,通过加入定位格架的方式能展平冷却剂出口温度,降低最大包壳温度。对环形燃料组件的安全分析表明,从中子学角度该组件是安全的。     

压力管式超临界水堆极限事故下“无堆芯熔化”概念评估

在自主研发的事故分析程序SCTRAN的基础上,开发并验证了二维导热模型和辐射换热模型,并将改进后的SCTRAN应用于加拿大压力管式超临界水堆在失水事故（LOCA）叠加丧失紧急堆芯冷却系统（LOECC）事故中的堆芯安全评估,并对燃料棒到慢化剂之间的传热效率以及关键的影响因素进行了评估。计算结果表明,在LOCA叠加LOECC工况下,燃料棒到燃料通道的辐射换热和燃料棒到蒸汽的自然对流换热能够有效导出反应堆的衰变余热,最高功率的燃料组件内、外圈燃料棒的最高包壳温度分别为1278℃和1192℃,均低于不锈钢包壳的熔化温度,因此整个事故过程中不会发生堆芯熔化。      

次临界或低功率启动工况下提棒事故分析

由于控制棒抽出引起堆芯内反应性失控增加,从而导致核功率剧增的事故定义为一组控制棒组件抽出事故。这种瞬态可能是反应堆控制系统或棒控系统失灵引起的。多普勒负反应性反馈效应能在保护动作延迟的时间内将功率限制在可接受的水平。该事故中,燃料棒表面可能发生偏离泡核沸腾(departure from nucleate boiling,简称DNB),导致燃料元件包壳烧毁;燃料芯块也可能发生熔化,对包壳产生不利影响。文章对岭澳混合堆芯和提高富集度论证次临界或低功率启动工况下提棒事故进行了分析。分析结果表明,事故瞬态中不会发生燃料芯块熔化或燃料元件包壳烧毁,可以保证燃料元件的完整性,燃料设计满足限制准则。     

钠冷快堆控制棒分段设计研究

在钠冷快堆中,反应堆运行时的反应性补偿和停堆安全主要由控制棒来实现。当前的钠冷快堆设计中,一般含有安全棒、补偿棒和调节棒。其中,补偿棒中10B的富集度较高,使补偿棒的燃耗较高,且发热量较大,并造成周围燃料组件功率峰因子偏大。本文提出一种分段设计方案,可用于改进上述缺点。该方案相比于传统方案,控制棒发热减小约30%,控制棒燃耗减小50%,并能有效改善周围燃料组件的功率峰因子,控制棒更换周期可提升1倍。     

燃料组件弯曲的高保真全堆芯中子学计算方法及应用

为了研究燃料组件弯曲变形对堆芯功率分布的影响，提出了一种等效模拟压水堆堆芯内燃料组件弯曲的方法，即根据弯曲前后燃料组件四周的水隙材料的原子数目守恒原则，通过保持弯曲前后的水隙宽度不变，改变弯曲后水隙内所有核素的原子核密度，近似等效燃料组件弯曲后四周水隙的变化。通过蒙特卡罗程序NECP-MCX和确定论数值反应堆程序NECP-X对其正确性进行验证，并基于NECP-X程序对欧洲先进压水堆（EPR）全堆芯的燃料组件弯曲工况进行了模拟分析，计算结果表明：由于局部慢化效应变化，燃料组件小幅弯曲对堆芯功率分布影响相对较大，全堆芯问题中最大的偏移量在2 mm左右时可使组件功率的相对变化达到5%左右。     

Self-Limiting Power Excursion Characteristics of Light Water Reactor, (I)

The Doppler limited power excursion characteristics of a light water reactor and the shutdown mechanism by scram were analyzed on the Hitachi Training Reactor (HTR). For the purpose of the pulse operation tests, modifications were applied to the HTR to provide pulsing capability; a pulse rod was added, together with a back up device for shutdown, and provision of three instrumented fuel assemblies, equipped with thermocouples; the Al-clad fuel rods were replaced by stainless steel clad rods.

About 100 runs of pulse operation tests were performed in fullest security with reactivity insertions ranging up to 1.0 % Δk/k, in which last case the peak power reached 38 MW, with a reactor period of 29 msec.     

Dynamics aspects of plutonium burning in an inert matrix

Burnup calculations have been performed on a mini fuel assembly containing 21 fuel rods and four water holes at the corners. The fuel rod positions were filled with 4% enriched UO₂ fuel and with either reactor grade or weapons grade plutonium mixed in an inert matrix. The ratio between the UO₂ and the IMF rods was varied to investigate the influence of the UO₂ fuel on the dynamics of the assembly. From a simple reactor model with one delayed neutron group and first-order fuel and temperature feedback mechanisms, the linear transfer function from reactivity to reactor power was calculated that was subsequently used in a root-locus analysis. From this, it is concluded that only 20% of the fuel rods need to be made of UO₂ to have a fuel that is linearly stable up to 1000 days of irradiation.     

Research on Segment Design of Control Rod in Sodium-cooled Fast Reactor

In sodium-cooled fast reactors, control rods are commonly used to compensate for the excess reactivity and shut down the reactor. The traditional sodium-cooled fast reactor design consists of the safety rod, shim rod and regulating rod. The 10B enrichment of the shim rods is relatively higher, which unavoidably increases the burnup, the heat generation and the power peak factor of the fuel assemblies around the shim rods. To solve this issue, the segment design of control rods was proposed. Compared with traditional design, the new design can significantly reduce the heat generation by about 30 percent and burnup of control rods by about 50 percent, as well as improve the power peak factor of the fuel assemblies around the shim rods. The replacement cycle of the control rods can be extended by time.     

International Solvent Extraction Conference ISEC ′90

The supercritical-water-cooled power reactor (SCPR) is expected to reduce power costs compared with those of current LWRs because of its high thermal efficiency and simple reactor system. The high thermal efficiency is obtained by supercritical pressure water cooling. The fuel cladding surface temperature increases locally due to a synergistic effect from the increased coolant temperature, the expanded flow deflection due to coolant density change and the decreased heat transfer coefficient, if the coolant flow distribution is non-uniform in the fuel assembly. Therefore, the SCPR fuel assembly is designed using a subchannel analysis code based on the SILFEED code for BWRs.

The SCPR fuel assembly has many square-shaped water rods. The fuel rods are arranged around these water rods. The fuel rod pitch and diameter are 11.2 mm and 10.2 mm, respectively. Since coolant flow distribution in the fuel assembly strongly depends on the gap width between the fuel rod and the water rod, the proper gap width is examined. Subchannel analysis shows that the coolant flow distribution becomes uniform when the gap width is 1.0 mm. The maximum fuel cladding surface temperature is lower than 600°C and the temperature margin of the fuel cladding is increased in the design.     

RAPID Operator-Free Fast Reactor Concept without Any Control Rods Reactor Concept and Plant Dynamics Analyses

The 1,000kWe metal fueled sodium-cooled fast reactor concept “RAPID” to achieve highly automated reactor operation has been demonstrated. RAPID (Refueling by All Pins Integrated Design) is designed for a terrestrial power system which enables quick and simplified refueling. It is one of the successors of the RAPID-L, the operator-free fast reactor concept designed for lunar base power system. The essential feature of the RAPID concept is that the reactor core consists of an integrated fuel assembly instead of conventional fuel subassemblies. In this small-size reactor core, 14,000 fuel pins are integrated and encased in a fuel cartridge. Refueling is conducted by replacing a fuel cartridge. The reactor can be operated without refueling for up to 10 years.

Unique challenges in reactivity control systems design have been addressed in the RAPID concept. The reactor has no control rod but involves the following innovative reactivity control systems: lithium expansion modules (LEM) for inherent reactivity feedback, lithium injection modules (LIM) for inherent ultimate shutdown, and lithium release modules (LRM) for automated reactor startup. All these systems adopt ⁶Li as a liquid poison instead of B₄C rods. In combination with LEMs, LIMs and LRMs, RAPID can be operated without an operator. In this paper, the RAPID reactor concept and its transient characteristics are presented.     

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.     

Determination of Heats of Formation for Solid and Liquid Lithium Metasilicates by Vaporization Method

In-reactor experiments were performed in Nuclear Safety Research Reactor of Japan Atomic Energy Research Institute to study the failure behavior of stainless steel clad fuel rods under a simulated reactivity initiated accident (RIA) condition. A single test fuel rod with stainless steel cladding was contained in a capsule filled with water at room temperature and atmospheric pressure and irradiated by pulsing power simulating an RIA. It was revealed through the experiments that the failure mechanism of the stainless steel clad fuel rod was cladding melting, which was different from oxygen-induced embrittlement observed in the Zircaloy clad fuel rod in the same test condition, and the failure threshold energy was determined to be about 240cal/g·UO₂ (–1,000 kJ/kg·UO₂), which was about 20 cal/g·UO₂ (–85 kJ/kg·UO₂) lower than that of the Zircaloy clad fuel rod. It was also found that the mechanical energy was generated by explosive vaporization of coolant due to molten fuel-coolant interaction as a consequence of the fuel rod failure accompanying fuel pellet fragmentation at an energy deposition of nearly 380 cal/g·UO₂ (–1,600 kJ/kg·UO₂) or more.