基于MOOSE平台的棒状燃料元件性能分析程序开发与验证

为实现对压水堆棒状燃料元件的精细化模拟,本文基于有限元平台MOOSE开发了棒状燃料元件性能分析程序BEEs。针对程序的燃耗、氧化层厚度和燃料温度模块分别进行了对比验证,采用BR3 Rod实验算例验证了长期稳态工况下BEEs程序的整体分析功能。结果表明,BEEs程序能获得合理准确的模拟结果,初步具备了稳态工况燃料性能分析能力。     

The development of a transient version of the HOTROD computer code

The full length fuel rod steady state performance code HOTROD has been developed to predict fuel performance during a transient. This paper explains the theory used to calculate the transient fuel temperatures using the Crank-Nicolson method. Transient HOTROD predictions of SGHWR and PWR axial clad temperatures during a loss-of-coolant accident are given and are compared with those predicted by other codes. The code is being further developed to model Zircaloy clad ballooning and the creep equations coupled to the heat transfer equation derived in this paper.     

BEAF - a computer program for analysis of light water reactor fuel rod behavior

BEAF - a computer program for analysis of light water reactor fuel rod behavior was developed. The BEAF code, which is appropriate for on-line prediction of fuel rod behavior, can analyze fuel rod thermal and mechanical behaviors using the axisymmetric, plane strain approximation and finite difference method to realize a fast running time.In the mechanical analysis, a new cracked pellet compliance model is introduced, in which pellet cracking and crack healing, pellet initial relocation, modified elastic moduli of a cracked fuel pellet, and stress dependent hot pressing are considered. Adding to those capabilities, fission gas flow and diffusion in the fuel-clad gap are analyzed to take into account the slow fission gas dilution effect on the gap conductance during power ramp.     

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。     

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.     

高燃耗下裂变气体释放行为研究与程序校验

采用燃料棒性能分析程序COPERNIC,针对哈尔登（Halden）测试燃料组件 （IFA）519.9 DK 辐照试验燃料棒辐照试验进行了计算分析,研究了高燃耗下裂变气体释放行为,并与试验数据进行了对比验证。结果表明,在燃耗达到约100 GW?d/t(U)的辐照过程中,该程序对裂变气体释放率的预测值与试验测量结果符合较好;程序未精确预测芯块孔隙率在高燃耗“边缘结构”内的演化过程,但不影响其对燃料棒辐照综合性能分析的准确性和合理性。      

REA analysis of the Iranian VVER-1000 core by modification of PRORIA code for annular pellets

The PRORIA code and its recent modifications are described here. The PRORIA code analyzes the transient response of the core against the reactivity increase caused by the control rod rapid withdrawal. The code solves and analyzes neutronic and thermal–hydraulic equations simultaneously. The code is designed for western PWR-type reactor performance. The equations representing thermal–hydraulic and neutronic should be modified to use the code to analyze VVER-1000 reactor core transients, because The VVER-1000 reactor fuel has a central hole in the fuel pellet. In a cylindrical solid fuel pellet, operation of an oxide fuel material at high temperature alters its morphology and the inner region is restructured to form a void at the center surrounded by a dense fuel region. Most of the restructuring occurs within the first few days of operation with slow changes afterward. Hence, the effects of a central hole in mathematical equations and in the transient are investigated. After the code modification, three accident scenarios with control rod ejection are simulated. The results are in good agreement with those reported in the plant’s FSAR. The results show that the peak fuel temperature in the hot fuel pin is lower than what the original code predicts by 150–500 °C. Furthermore, the Doppler reactivity effect, when the fuel pellet has a central hole, is higher than the solid fuel pellet.     

Mixed-oxide (MOX) fuel performance benchmarks

Within the framework of the OECD/NEA Expert Group on Reactor-based Plutonium disposition (TFRPD), fuel modeling code benchmarks for MOX fuel were initiated. This paper summarizes the calculation results provided by the contributors for the first two fuel performance benchmark problems. A limited sensitivity study of the effect of the rod power uncertainty on code predictions of fuel centerline temperature and fuel pin pressure also was performed and is included in the paper.     

Analysis of the degradation of a PWR control rod during a loss of coolant accident, using the ICARE2 code

Conditions leading to AIC control rod damage during a loss of coolant accident in a PWR geometry, even in absence of violation of the LOCA licensing criteria, are investigated using several versions of the ICARE2 code (IPSN). Before being applied to the reactor case, the code and the modelling procedure are validated against the out-of-pile severe fuel damage experiment CORA-5. Three particular initial configurations are considered for the subsequent control rod damage analysis: nominal control rod and guide tube geometry, zircaloy guide tube bowing with concurrent cladding thickness reduction and finally control rod cladding perforation. For each of these cases the thermal, mechanical and chemical behaviour is presented. Phenomena such as ballooning and cladding failure of fuel rods, guide tube failure, melt relocation and final fluid channel cross-section modification are described. Finally, the conclusions of numerous sensitivity studies are discussed and some suggestions are given for possible improvements of the ICARE2 code.     

Modeling the behavior of Xe,I, Cs,Te, Ba,and Sr in solid and liquefied fuel during severe accidents

This paper describes the primary physical/chemical models recently incorporated into a mechanistic code (FASTGRASS) for the estimation of fission product release from fuel, and compares predicted results with test data. The theory of noble gas behavior is discussed in relation to its effect on the release behavior of I, Cs, Te, Ba, and Sr. The behavior of these fission products in the presence of fuel liquefaction/dissolution and grain-growth phenomena is presented, as is the chemistry of Sr, Ba, I, and Cs.Comparison of code predictions with data indicates the following trends. Fission product release behavior from solid fuel strongly depends on fuel microstructure, irradiation history, time at temperature, and internal fuel rod chemistry. Fuel liquefaction/dissolution, fracturing, and oxidation also exert a pronounced effect on release during fuel rod degradation. For low burnup fuel (e.g., TMI-2), appreciable fission product retention in previously liquefied fuel can occur due to the low concentration of fission products, and the limited growth of bubbles in the liquefied material.Many of the calculations described in this paper were made with a version of FASTGRASS developed for use on a personal computer (IBM compatibile).     

Vibration of nuclear fuel bundles

Several mathematical models have been proposed for calculating fuel rod responses in axial flows based on a single rod consideration. The spacing between fuel rods in liquid metal fast breeder reactors (LMFBRs) is small; hence fuel rods will interact with one another due to fluid coupling. The objective of this paper is to study the coupled vibration of fuel bundles. To account for the fluid coupling, a computer code (AMASS) is developed to calculate added mass coefficients for a group of circular cylinders based on the potential flow theory. The equations of motion for rod bundles are then derived including hydrodynamic forces, drag forces, fluid pressure, gravity effect, axial tension and damping. Based on the equations, a method of analysis is presented to study the free and forced vibrations of rod bundles. Finally, the method is applied to a typical LMFBR fuel bundle consisting of seven rods.     

Calculational procedure for determining creep collapse of LWR fuel rods

Observed collapses in pressurized water reactor fuel rods have been attributed to the radiation enhanced creep of Zircaloy cladding into regions where separations in the fuel pellet stack have occurred. A computer code, COLAPX, has been written to determine the growth of ovality and the ultimate collapse of fuel rod cladding under reactor operating conditions. This paper describes the theoretical bases of this code, the finite element formulation used, the constitutive relations between the displacement fields and the element forces, and the radiation, temperature and stress dependent material model for creep of Zircaloy tubing. Comparisons of the creep rate predictions and of the ovality predictions with data from irradiated tubes and fuel cladding are presented.     

The D-COM blind problem on fission gas release: The predictions of the TRANSURANUS and future codes

The concentration of retained xenon, the percentage of porosity and the UO₂ grain size have been measured as a function of radial position in the base irradiated rod AG11-8 and the transient tested rod AG11-10. In the base irradiation, densification of the fuel took place and slight grain growth occurred at the pellet centre. Gas release was not detected. During the transient test, 15–20% of the xenon inventory was released from the fuel grains. Gas release was accompanied in the central region of the fuel by an increase in the porosity from 4.7 to 6–8%. These findings are compared with the predictions made by the fuel performance code TRANSURANUS. The code predictions are in good agreement with the experimental observations. FUTURE was used to investigate the development of gas bubbles and the mechanisms controlling gas release in the rods during the base irradiation and the transient test. According to FUTURE fission gas will have accumulated on the grain boundaries during the base irradiation. The code indicates that variations in the fuel microstructure resulting from the base irradiation will have caused the level of gas release to vary along the fuel stack in rods AG11-9 and AG11-10 during the transient test. FUTURE also suggests that fission induced bubble re-solution became increasingly important for release during the latter stages of the transient test. Moreover, the code calculations imply that bubble migration could have played a significant role in the release process.     

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 mechanical load on cladding induced by fuel swelling during power ramp in high burn-up rod by fuel performance code FEMAXI-6

Mechanical load on cladding induced by fuel swelling in a high burn-up BWR type rod was analyzed by a fuel performance code FEMAXI-6. The code was developed for the analysis of LWR fuel rod behaviors in normal operation and transient conditions using finite element method (FEM).During a power ramp for the high burn-up rod, instantaneous pellet swelling can significantly exceed the level that is predicted by a “steady-rate” swelling model, causing a large circumferential strain in cladding. This phenomenon was simulated by a new swelling model to take into account the fission gas bubble growth. As a result it was found that the new model can give reasonable predictions on cladding diameter expansion in comparison with PIE data. The bubble growth model assumes that the equilibrium state equation holds for a bubble under external pressure, and simultaneous solution is obtained with both bubble size determination equation and diffusion equation of fission gas atoms. In addition, a pellet-clad bonding model which has been incorporated in the code to assume solid mechanical coupling between pellet outer surface and cladding inner surface predicted the generation of bi-axial stress state in the cladding during ramp.     

核燃料棒温度场和系统熵增分布计算与研究

基于热力学第二定律以及压水核反应堆燃料棒稳态传热偏微分方程的一般形式,通过熵增的数值计算,讨论分析燃料棒内热量的传递过程和方向,以及能量品质的得失。先对典型二维矩形域传热问题进行数值计算,并采用解析解对该数值方法进行了验证。然后对含有内热源的单根燃料棒进行传热计算,讨论其温度和熵增分布情况,通过熵增云图分析可以展现燃料棒内热量的传递过程和品质变化,为核反应堆热工设计提供有益参考。     

Improved lumped parameter formulation for simplified LWR thermohydraulic analysis

Transient heat transfer in a nuclear fuel rod is modelled by an improved lumped parameter approach. Hermite approximation for integration is used to obtain the average fuel and cladding temperatures in the radial direction. Thermohydraulic behaviour of a pressurized water reactor (PWR) during reactivity insertion and partial loss-of-flow is simulated by using a simplified mathematical model of reactor core and primary coolant. Transient temperature response of fuel, cladding and coolant is analysed.