中国铅基研究实验堆燃料元件活性区温度场计算分析

中国铅基研究实验堆(CLEAR-Ⅰ)被确定为中国科学院加速器驱动次临界系统(ADS)专项的主选堆型。燃料元件是铅基反应堆的核心部件之一,因此需确保燃料元件的芯块中心温度和包壳最高温度符合设计准则的要求。本文利用有限元程序ANSYS对燃料元件活性区在正常运行工况和失流事故下的温度场进行了数值模拟与分析。正常运行工况下的模拟结果表明,芯块中心温度远低于UO2的熔化温度限值,包壳最高温度低于材料的使用温度限值,满足设计准则中关于上限使用温度的要求。失流事故下的模拟结果表明,失流事故发生后,芯块中心温度和包壳最高温度都会明显上升。当冷却剂流速降低到0.1m/s时,包壳最高温度将超过正常使用温度;紧急停堆滞后时间超过17.5s时,包壳的最高温度将超过事故温度限值。以上分析结果可作为燃料元件安全评审工作的基础。     

UO_2混合芯块及锆合金涂层耐事故燃料热特性研究

在UO_2芯块中添加不同份额的SiC成分,并在M5锆合金包壳外增加不同厚度的SiC涂层结构组合成耐事故燃料元件,并建立混合芯块-锆合金包壳-涂层间热传导模型。计算并调整UO_2混合芯块、SiC涂层热物性参数,以秦山第二核电厂1号和2号机组长循环燃料管理方案为背景,对比分析UO_2混合芯块不同添加成分比例,以及M5锆合金外涂层不同厚度对于燃料棒热性及裂变气体释放结果的敏感性影响。计算结果显示耐事故燃料在瞬态工况下能更有效地降低燃料芯块中心温度。     

燃料污垢对典型事故工况的影响研究

新版HAD 102/07—2020核动力厂反应堆堆芯设计中明确要求：设计分析应考虑反应堆冷却剂系统正常运行产生的腐蚀产物在包壳表面的沉积导致的燃料棒传热恶化。因此,有必要分析燃料污垢对事故工况下燃料棒传热性能的影响,特别是以燃料芯块温度和包壳温度为验收准则的典型事故工况。本文开发污垢计算模型,采用等效热导率关系式计算含污垢和氧化层的包壳热导率,即认为污垢、氧化层均匀分散在包壳层中,使得包壳热导率变化,该等效包壳层所引起的温度梯度与实际情况相同。随后,基于对“华龙一号”核动力厂事故分析结果,选取了典型非LOCA事故（弹棒事故、功率运行下单个控制棒失控抽出事故）和LOCA事故进行污垢影响研究。结果表明,考虑污垢后,事故过程中的燃料芯块中心峰值温度和包壳峰值温度均有显著上升,但依然满足事故验收准则要求。     

压水堆燃料棒三维热-力学性能的精细模拟

压水堆燃料棒工作在复杂的辐照、热和力学环境中,对其性能进行定量评估涉及多种复杂的物理现象。目前常用的燃料性能分析程序一般对结构采用简化的轴对称假设,对辐照肿胀、辐照蠕变和高温蠕变等物理现象以及辐照-热-力等物理场之间的耦合考虑并不充分。基于ABAQUS有限元求解框架,开发了压水堆燃料棒三维热-力学性能的模拟程序,利用程序对压水堆燃料棒进行了稳态分析,以及升功率和反应性引入事故两种瞬态分析。结果表明:辐照引起燃料致密化和肿胀对燃料温度变化有重要影响;芯块应变增加主要是由裂变产物肿胀引起的;芯块几何结构导致包壳应力集中发生在芯块间的交界面处;燃料棒功率的急剧变化会加快芯块表面破裂的进程;反应性引入事故会导致芯块从内部开始破裂,并会引发芯块-包壳的接触。     

板状燃料元件堆芯热工水力特性分析程序开发及验证

采用Visual Fortran 6.5程序语言,基于质量、动量和能量守恒方程,以及合理的流动传热和物性关系式,开发了板状燃料元件堆芯热工水力特性分析程序.利用该程序计算了IAEA 10MW MTR 基准题中定义的堆芯反应性引入和堆芯失流事故.结果表明:本文计算所获得的停堆时刻功率、燃料芯块最高温度、包壳外壁面最高温度以及冷却剂出口温度与文献的计算结果吻合良好,验证了本程序模型的正确性.     

高燃耗下UN-FeCrAl燃料元件稳态及瞬态热力学性能分析

UN-FeCrAl燃料元件作为耐事故燃料高燃耗应用的主要方案之一,需要评价其在高燃耗下的热力学性能。本研究基于FUPAC软件对UN-FeCrAl燃料元件在燃耗68000 MW·d·t^-1(U)下的稳态和瞬态热力学性能进行了预测。分析结果表明,稳态工况下UN-FeCrAl燃料元件热力学性能表现良好;瞬态下UN燃料的芯块中心温度最高仅为862℃,可满足芯块温度设计要求,但FeCrAl包壳的瞬态应力最大将达到459 MPa,且瞬态应变量相比于稳态应变量最大增加了0.23%,这可能会使FeCrAl包壳面临瞬态应力和瞬态应变准则超限的风险。因此后续研究应重点关注FeCrAl包壳的瞬态应力和瞬态应变性能。     

Zr-2包壳管爆破力学性能

核燃料元件是反应堆的核心部件，由燃料芯块、包壳及其构件组成。由于燃料元件的运行环境比较恶劣，中子辐照、冷却剂的腐蚀及在开堆、停堆、和运行后期燃料芯块与包壳的机械相互作用和裂变气体产物的释放，使包壳管承受双向应力，均会造成燃料元件的力学性能下降，形成安全隐患，它的安全性能直接影响反应堆的安全可靠性。为更好地模拟包壳在堆内的受力状态，一般采用内压爆破试验来获得包壳材料的断裂强度与延性数据。     

压水堆燃料包壳破损条件下裂变气体非稳态释放数值模拟研究

压水堆燃料包壳破损后,芯块-包壳间隙内积累的裂变气将释放到冷却剂中,其内部的微观机理还尚不清楚。为了揭示裂变气体释放过程中冷却剂与气体的相互作用规律,基于三维计算流体力学（CFD）方法对该物理过程展开数值模拟,所利用的模型为VOF模型以及k-ε模型。模拟结果表明,包壳破损后冷却剂首先进入芯块-包壳间隙,在芯块-包壳间隙内蒸发,引起芯块-包壳间隙内压强上升,而后裂变气体释放到子通道;裂变气体从芯块-包壳间隙释放到子通道可分为2个阶段。第一阶段:芯块-包壳间隙与子通道间压差较大,气体射流进入子通道,该阶段持续时间较短,裂变气体释放率较大,且变化也较大。第二阶段:芯块-包壳间隙与子通道间压差较小且相对平稳,裂变气体通过破口内涡的对流传质进入子通道,该阶段持续时间较短,裂变气体释放率较小,且相对稳定。      

环形燃料元件几何尺寸对其热工性能的影响研究

针对环形燃料元件，基于欧洲铅冷系统反应堆ELSY选取环形燃料元件参数，建立环形燃料元件导热模型，设定环形燃料元件的初始参数并利用MATLAB编制环形燃料元件导热计算程序，通过制定的三个评估标准研究环形燃料流量分配比、内外包壳厚度、内外气隙厚度和芯块厚度对环形燃料元件热工性能的影响并进行几何尺寸修正。研究结果表明：适当增大流量分配比、减小内包壳厚度、增大外包壳厚度、减小内外气隙间距和减小芯块厚度可改善元件的热工性能；设定流量分配比为1、内包壳厚度0.06 cm修正为0.04 cm、外包壳厚度0.06 cm修正为0.07 cm、内外气隙间距0.035 cm修正为0.015 cm、芯块厚度修正为0.05 cm，进行这些几何尺寸修正后，芯块的最高温度下降了90 K(8.6%)，绝热面位置偏离芯块几何中心不足2μm，内外通道冷却剂出口温差不足2 K，环形燃料元件热工性能得到了明显提高。     

芯块侧偏状态下的燃料棒温度场数值计算分析

燃料芯块侧偏状态下的燃料棒温度分布关系到反应堆燃料设计和安全运行。本文基于燃料棒的稳态扩散方程的一般形式,通过数值计算分析了芯块侧偏对燃料棒传热和温度分布的影响。结果表明:当燃料芯块侧偏时,芯块最高温度的位置向芯块侧偏的反方向偏移且最高温度下降,偏心率越大,最高温度的位置偏移程度越大,温降也越大。当偏心率e为0.5和0.8时,芯块最高温度分别下降1.3%和4.1%。而燃料棒包壳外壁面温度分布不均匀且最高温度随着偏心率的增大而升高,当偏心率e=0.8时,燃料棒包壳外壁面的最高温度为350℃,达到燃料棒的临界工作温度。     

FARST — A computer code for the evaluation of FBR fuel rod behavior under steady-state/transient conditions

FARST, a computer code for the evaluation of fuel rod thermal and mechanical behavior under steady-state/transient conditions has been developed. The code characteristics are summarized as follows:

1. (i) FARST evaluates the fuel rod behavior under the transient conditions. The code analyzes thermal and mechanical phenomena within a fuel rod, taking into account the temperature change in coolant surrounding the fuel rod.

2. (ii) Permanent strains such as plastic, creep and swelling strains as well as thermoelastic deformations can be analyzed by using the strain increment method.

3. (iii) Axial force and contact pressure which act on the fuel stack and cladding are analyzed based on the stick/slip conditions.

4. (iv) FARST used a pellet swelling model which depends on the contact pressure between pellet and cladding, and an empirical pellet relocation model, designated as “jump relocation model”.

The code was successfully applied to analyses of the fuel rod irradiation data from pulse reactor for nuclear safety research in Cadarache (CABRI) and pulse reactor for nuclear safety research in Japan Atomic Energy Research Institute (NSRR).The code was further applied to stress analysis of a 1000 MW class large FBR plant fuel rod during transient conditions. The steady-state model which was used so far gave the conservative results for cladding stress during overpower transient, but underestimated the results for cladding stress during a rapid temperature decrease of coolant.     

Analysis of radiant heat transfer in a BWR fuel assembly

A computer code ‘CIDER’ was developed which analyzes radiant heat transfer in a BWR fuel rod bundle under loss of coolant conditions. In the code, (1) a channel box and fuel rods are considered to be gray bodies, (2) reflection and absorption of radiation beams in the atmosphere is neglected, (3) a fuel rod is approximated by a regular polygonal rod, and (4) radiant heat flux is calculated considering circumferential temperature distribution on each fuel rod surface, which is determined from radial and circumferential heat conduction calculations in a fuel rod. It was found that the conventional model with uniform cladding temperature overestimated heat flux about 30% in a typical situation, or correspondingly underestimated the temperature rises.     

Analysis of MOX fuel behavior in reduced-moderation water reactor by fuel performance code FEMAXI-RM

To assess the feasibility of the 31% Pu-MOX fuel rod design of reduced-moderation water reactor (RMWR) in terms of thermal and mechanical behaviors, a single rod assumed to be irradiated in the core of RMWR up to 106 GWd/tHM has been analyzed by a fuel performance code FEMAXI-RM which is an extended version of FEMAXI-6 code. In the analysis, design specifications of fuel rod and irradiation conditions have been input, and available models of both MOX fuel and UO₂ fuel have been used as appropriate. The results are: fission gas release is several tens of percent, rod internal pressure does not exceed the coolant pressure, and the highest fuel center temperature is 2400 K, while cladding diameter increase caused by pellet swelling is within 1% strain. These predictions suggest that the MOX fuel rod integrity will be held during irradiation in RMWR, though actual behavior of MOX pellet swelling and cladding oxidation require to be investigated in detail.     

基于MOOSE框架的五方程两相流分析程序开发

基于多物理场耦合框架MOOSE,采用五方程两相流模型开发了模块化程序ZEBRA,实现了高阶时间、空间离散格式两相流动传热问题的求解。采用Bartolomei开展的垂直圆管过冷沸腾实验对ZEBRA进行验证,在不同热流密度、质量流密度、压力工况下,将程序计算值与实验值进行了数值验证和计算分析。结果表明：ZEBRA中五方程模型预测值与实验值符合良好,沸腾起始点和空泡份额的预测合理,表明ZEBRA初步具备了处理两相流问题的能力。     

压水堆燃料辐照性能多物理耦合并行分析程序开发

核燃料元件是反应堆的核心部件,其性能影响反应堆的安全性与经济性,利用燃料元件性能分析程序开展燃料堆内稳态辐照性能分析对于燃料设计及安全评价具有重要意义。通过开发燃料温度分布、变形计算、裂变气体释放及内压等模型,结合燃料元件热工-力学多物理耦合计算分析耦合方案,基于先进并行计算方法构建了高性能并行化燃料性能分析程序Athena。利用典型商用压水堆核电站数据及同类程序计算结果进行了程序初步验证,结果表明Athena程序计算结果合理可靠。通过定义堆芯功率及热工水力边界条件,程序能够并行开展压水堆全堆芯燃料辐照性能分析,提高燃料辐照性能分析效率,是数值反应堆原型系统（CVR1.0）的重要组成。     

Fission Gas Induced Cladding Deformation of LWR Fuel Rods under Reactivity Initiated Accident Conditions

Behavior of irradiated fuel rods under power burst conditions by accidental reactivity insertion in light water reactors (LWRs) has been studied in the Nuclear Safety Research Reactor (NSRR). In the experiments, cladding hoop deformation, which reached up to about 10%, was much larger than that of the fresh rods. The current LWR fuel behavior analysis codes, which only take account of the thermal expansion of the fuel pellets for the deformation calculation, under-predicted the plastic deformation of the cladding to be less than about 1%. Fission gas release during the pulse irradiation tests reached as high as 22% in the NSRR irradiated fuel tests. In order to describe these test results, a model of grain boundary fission gases to cause the cladding deformation has been developed and installed in a fuel behavior simulation code, FRAP-T6. In the model, the over-pressurized gases by the pulse irradiation cause grain boundary separation and stress the cladding during the tests. The model assumes that the gases remain in the fuel during the early part of pulse irradiation and are released to the open volume in the rod after the cladding deformation. The model, in combination with a fuel thermal expansion model, GAPCON, which was validated through fresh fuel tests, reproduces the NSRR test results reasonably well.     

Evaluation of LWR fuel rod behavior under operational transient conditions

To evaluate the effects of fission gas flow and diffusion in the fuel-cladding gap on fuel rod thermal and mechanical behaviors in light water reactor (LWR) fuel rods under operational transient conditions, computer sub-programs which can calculate the gas flow and diffusion have been developed and integrated into the LWR fuel rod performance code BEAF. This integrated code also calculates transient temperature distribution in the fuel-pellet and cladding.The integrated code was applied to an analysis of Inter Ramp Project data, which showed that by taking into account the gas flow and diffusion effects, the calculated cladding damage indices predicted for the failed rods in the ramp test were consistent with iodine-SCC (Stress Corrosion Cracking) failure conditions which were obtained from out-of-reactor pressurized tube experiments with irradiated Zircaloy claddings. This consistency was not seen if the gas flow and diffusion effects were neglected. Evaluation were also made for the BWR 8 × 8 RJ fuel rod temperatures under power ramp conditions.     

Modeling of Jet Pump by RELAP4 Code in Loss-of-Coolant Accident Analysis

Waterlogged fuel rod experiments performed at the NSRR are analyzed using the computer code WTRLGD, which was devised for the analyses of thermo-dynamical behavior of a waterlogged fuel rod. The numerical results are compared with the data from the experiments in order to assess the validity of the computer code. Parameters in the analyses are volumetric fraction of water, reactor period, gap width, a pin hole and the end peaks. Thus the analyses cover almost all the waterlogged fuel rod experiments at the NSRR.

The comparison shows good agreement between the experimental results and numerical ones on the transient thermo-dynamical behaviors of fuel, such as, rod internal pressure, cladding surface temperature and cladding strain. The numerical results also quantitatively agree with the experimental data concerning the effects of the above parameters on failure threshold energy. From the above findings, the computer code is assessed to be valid enough for the analyses of the failure behavior of the waterlogged fuel rod under a reactivity initiated accident condition.