U-10Mo/Zr单片式燃料元件堆内稳态热-力学性能研究

本文建立了U-10Mo/Zr单片式燃料元件的辐照性能模型以及热-力学本构关系,采用有限元方法进行非均匀辐照场中燃料元件稳态热-力学性能的数值模拟,获得并分析了U-10Mo/Zr单片式燃料元件温度、形变和应力的分布特点及变化规律。研究结果表明,燃料芯体厚度增量在芯体和包壳结合面附近达到最大,主要受到燃料辐照蠕变的影响;在较低燃耗条件下,燃料芯体高温辐照肿胀模拟结果与低温辐照肿胀试验结果相当;燃料芯体边角区域和包壳端面外侧区域存在应力集中。      

UMo/Zr单片式燃料板堆内热力耦合行为研究

UMo/Zr单片式燃料板在堆内辐照环境下会经历复杂的多场耦合及多尺度关联的行为。针对均匀辐照的堆内工况条件,建立了对UMo/Zr单片式燃料板的堆内行为进行多尺度模拟的方法,并计算分析了元件的温度场、变形和主要应力场随燃耗演化的规律,获得了芯体与包壳界面层间应力的分布与演化规律。研究结果表明,芯体的最高温度会随着辐照时间持续增长;芯体厚度随着辐照时间而增大,在靠近芯体的边界附近厚度增长较多,与辐照后相关检测结果相符;芯体的Mises应力要远小于包壳中的Mises应力;芯体和包壳界面正应力最大值位于靠近芯体角部的位置,界面角部区域较大的界面拉应力易导致此处首先产生界面破坏。     

弥散型燃料芯体结构的蒙特卡罗模拟(英文)

在研究堆中的辐照条件下，U3Si2-Al 弥散型燃料的燃料颗粒和基体界面发生相互扩散。由于相互扩散反应,在每个 U3Si2颗粒的周围形成 U3Al7Si2反应层。反应层厚度随辐照时间和裂变密度而增加。反应层的形成造成了 U3Si2燃料和铝基体的消耗。该过程导致燃料芯体几何结构的演化。根据弥散体中燃料的随机分布特点,作者采用蒙特卡罗方法发展了燃料芯体结构演化的模拟方法。每个颗粒的特性都可以用直径和位置来表示。芯体结构参数包括颗粒尺寸分布、制造状态下的燃料体积分数、反应层厚度、反应层体积、U3Si2燃料体积分数、铝体积分数、接触几率和颗粒相互连接分数。特别是对于制造状态下的燃料体积分数为 43%时,颗粒尺寸较好地服从正态分布。模拟了在 6 mm×6 mm×0.5 mm 的芯体体积中 13 000 个抽样颗粒的情况下,各芯体结构参数随反应层厚度从 0～16 μm 变化时的函数变化情况。     

弥散核燃料等效辐照肿胀计算模拟

本文将弥散核燃料芯体看作一种特殊的颗粒复合材料,利用细观计算力学的方法,假设燃料颗粒在芯体中周期性分布,建立了对芯体等效辐照肿胀进行计算模拟的有限元模型。考虑颗粒的辐照肿胀和基体材料的辐照硬化效应,分别建立了燃料颗粒和基体材料的应力更新算法,编制了用户材料子程序,在Abaqus软件中实现了芯体等效辐照肿胀的有限元模拟。计算分析了颗粒大小和体积含量对芯体等效辐照肿胀的影响,并得到了等效辐照肿胀的拟合公式。研究结果表明,影响芯体等效辐照肿胀的主要因素是颗粒的辐照肿胀和体积含量。     

界面反应引起弥散型燃料的芯体结构演化

在反应堆运行工况下，Ｕ３Ｓｉ２－Ａｌ弥散型燃料的燃料颗粒与基体的界面相互扩散形成反应层，从而导致芯体结构的演化。本文根据Ｍｏｎｔｅ－Ｃａｒｌｏ原理建立了弥散型燃料芯体的模拟方法，并用该方法模拟了燃料相体积分散为４３％和具有一定尺寸分布的球形燃料颗粒在芯片中的空间随机排列。     

回收铀燃料元件UO2芯块辐照后显微组织研究

为评价回收铀燃料元件中UO₂芯块的辐照稳定性，采用热室金相显微镜对辐照后高放射性UO₂芯块沿轴向及径向的辐照肿胀、裂纹分布、晶粒尺寸及分布和晶粒长大行为进行了观察和分析。结果表明:燃料元件芯块中均存在大量的裂纹，回收铀燃料元件UO₂芯块裂纹呈明显的环形分布特征，天然铀燃料元件UO₂芯块呈放射性发散分布特征。两者的燃料芯体晶粒呈等轴状，均出现从边缘区域向芯块中心区域晶粒逐渐长大现象，辐照后晶界变粗化。两者晶粒尺寸、形貌及分布特征并无明显差别。此外，在相同的堆内运行工况条件下，回收铀燃料元件UO₂芯块辐照肿胀不明显，芯块破碎程度及晶粒长大过程与天然铀并无明显差别。      

锆基弥散微封装燃料在稳态运行条件下的失效机理研究

为实现锆基弥散微封装燃料（M3燃料）的优化设计，进一步提升其在轻水堆（LWR）运行环境下的可靠性，需对其在稳态运行条件下的失效机理进行研究。本研究借助于ABAQUS有限元软件，通过二次开发建立了M3燃料的辐照-热-力耦合性能三维数值模拟分析方法，并基于此分析方法对M3燃料在稳态运行条件下的失效机理进行了研究。研究结果表明，稳态运行期间M3燃料的失效主要以辐照初期内致密热解碳层（IPyC层）的失效、辐照中后期疏松热解碳层（Buffer层）与IPyC层分开再接触后导致的碳化硅层失效为主。该研究结果可为后续M3燃料的优化设计提供指导。      

Monte Carlo有限元射线追迹法计算圆柱形闪烁体与光纤的耦合功率

本文建立了基于计算机模拟的蒙特卡罗有限元射线追迹法,应用该方法对不同结构尺寸的圆柱型LSO晶体和大芯径光纤耦合输出进行了模拟研究.模拟结果表明:闪烁体半径一定情况下,耦合功率增加的幅度随闪烁体厚度增加呈递减趋势,在闪烁体的厚度达到10 cm时增加闪烁体厚度,耦合功率增加趋于平缓;闪烁体厚度一定的情况下,在闪烁体的半径达到光纤立体角对应的半径后,继续增加半径,耦合功率增加幅度逐渐减少;在闪烁体的入射端面包95%的反射材料后,其耦合功率增加了90%.     

UN核芯TRISO燃料颗粒破损概率模型研究

TRISO燃料颗粒由核芯和4层包覆层组成,具有良好的裂变产物包容能力。TRISO燃料颗粒破损概率是表征TRISO燃料事故安全特性的关键参数。本文基于修正的PANAMA破损概率计算方法,在考虑UN核芯裂变气体释放导致的气体内压以及内外致密热解炭层辐照蠕变和收缩作用的基础上,开发了UN核芯TRISO燃料颗粒压力壳式破损概率计算方法,并采用IAEA基准题6和基准题9对模型进行了验证;基于开发的UN核芯TRISO颗粒破损概率计算方法,采用随机抽样统计方法分析了事故工况下UN核芯和包覆层设计参数（包括包覆层尺寸及密度）对UN核芯TRISO燃料颗粒破损概率的影响。研究结果显示,疏松热解炭（Buffer）层设计参数是影响TRISO颗粒破损概率的关键因素,可通过降低Buffer层尺寸及密度分布设计标准偏差的方法降低UN核芯TRISO燃料颗粒的破损概率。     

包覆燃料颗粒尺寸及其标准偏差对失效概率的影响

三结构同向性型(Tristructural isotropic,TRISO)包覆燃料颗粒是目前高温气冷堆和固态燃料熔盐堆采用的燃料元件。TRISO包覆燃料颗粒破损会导致裂变产物不可接受的释放,由此影响反应堆的安全运行。基于TRISO包覆燃料颗粒压力壳式破损模型,分析了TRISO包覆燃料颗粒核芯和各包覆层的尺寸对失效概率的影响,研究了TRISO包覆燃料颗粒核芯半径、疏松热解碳(Buffer)层厚度和碳化硅(Si C)层厚度的合理设计范围。同时,利用随机抽样统计的方法分析了TRISO包覆燃料颗粒核芯半径分布和各包覆层厚度分布对颗粒失效概率的影响。研究发现,降低Buffer层厚度分布的标准差至16μm可以使TRISO包覆燃料颗粒的失效概率降低一个数量级。     

Pellet compliance model based on out-of-pile simulation

The thermal and mechanical behavior of fuel rods is significantly influenced by the extent of their relocation and by compliance of the cracked pellets. Movement of the cracked pellet pieces towards the cladding results in softer pellets with crack voids which accommodate some fraction of the thermoelastic pellet deformation and make the pellet more compliant under the restraint of the cladding. It is difficult to model such a pellet compliance independently of experimental observations because the cracked pellet behavior is uncertain by nature.Electrically heated simulation of pellet-cladding mechanical interaction (PCMI) facilitates much quicker and more flexible experimentation than actual in-pile tests. Testing apparatus consists of the simulated fuel rod with hollow UO₂ pellets and a tungsten rod in the center, and a diameter measuring device including three pairs of diameter sensors. Test parameters include the pellet-cladding gap and the cladding thickness. Results show that rods with a smaller gap have a larger increasing rate of cladding diameter. This suggests that a group of cracked pellet pieces induced by thermal stress has an apparent compliance which increases with pellet-cladding gap. Results also show more sensitivity to cladding thickness than those calculated assuming pellets having intrinsic stiffness. This also suggests the compliant nature of cracked pellets.Such a compliant nature can almost be described by reducing the elasticity of the pellet. A simple pellet compliance model was obtained by fitting calculations with measurements to describe a cracked pellet as a uniform axisymmetric body with apparent elasticity.     

Effects gap heat transfer on LWR fuel behavior during an RIA transient: In-pile experimental results with helium and xenon fillet rods

Gap heat transfer characteristics and their effects on LWR fuel behavior during an RIA have been studied through the in-pile experiment with UO₂ pellet fuel rods. The report describes the experimental results obtained in the NSRR tests in which PWR type test fuel rods of helium and xenon filled as the gap gas have been irradiated in the pulse reactor, NSRR, to simulate the prompt heat up of RIAs. The relation between the cladding temperature history and the gap heat transfer conditions, and the effects of gap gas composition on fuel behavior and on the fuel failure threshold are discussed based on the in-pile experimental data.     

Numerical treatment of the cladding mechanics in the fuel rod computer code IAMBUS

IAMBUS (INTERATOM Model for Burn-Up Studies on Fuel Rods) is a digital computer code for the thermal and mechanical design, in-pile performance prediction and post-irradiation analysis of fuel rods. The mechanical analysis of the cladding is approximated by a state of generalized plane strain. The analysis includes routines for plasticity, creep and swelling due to void nucleation and growth. The numerical integration of these equations is described in detail and the accuracy of the results is discussed.     

Fuel densification and swelling: Relationship between burn-up induced axial and radial fuel dimensional changes

By comparison of axial length change measurements, both in-pile and PIE, with radial fuel dimensional changes, the following are obtained: (1) measurements of in-pile axial fuel stack length generally show a densification which often occurs at beginning-of-life; (2) recent PIE measurements of radial dimensional changes have shown, however, that axial fuel stack length changes do not reflect properly the radial swelling which is taking place. The reason is because the swelling occurs primarily in the hot center portion of the fuel pellets, thus the pellet shoulders, which determine the axial dimensions, do not experience the swelling.     

Thermal Stress Analysis of High-Burnup LWR Fuel Pellet Pulse-Irradiated in Reactivity-Initiated Accident Conditions

For RIA-simulated experiments in the NSRR with high-burnup PWR fuel and BWR fuel, numerical analyses were performed to evaluate the temporal changes of profiles of temperature and thermal stress in pellet induced by pulse power, using the RANNS code. The pre-pulse states of rods were calculated using the fuel performance code FEMAXI-6 along the irradiation histories in commercial reactors and the results were fed to the RANNS analysis as initial conditions of the rod. One-dimensional FEM was applied to the mechanical analysis of the fuel rod, and the calculated cladding permanent strain was compared with the measured value to confirm the validity of the PCMI calculation. The calculated changes in the profiles of temperature and stress in the pellet during an early transient phase were compared with the measured data such as the internal gas pressure rise, cracks and grain structure in the post-test pellet, anddiscussed in terms of PCMI and grain separation. The analyses indicate that the pellet cracking appearances coincided with the calculated tensile stress state and that the compressive thermal stress suppresses the fission gas bubble expansion leading to grain separation.     

Application of a two-dimensional ballooning model to out-pile and in-pile simulation experiments

The FRETA-B code which has thermal and mechanical models for the fuel behavior during a postulated Loss of Coolant Accident was applied to out-pile tube ballooning experiments and an in-pile integral simulation experiment MT-1 in the NRU reactor. From the analysis of the out-pile experiments, the ballooning model based on the thin-shell model together with an empirical rupture criterion was found effective in predicting the cladding deformation kinetics and the average strain left after rod rupture.The mechanical analysis of the MT-1 experiment was based on the calculation of the temperature history and its transverse distribution under reflooding which were consistent with the limited experimental data. The analysis indicated that the restriction of flow channel blockage to 70% in MT-1 was a result of an azimuthal cladding temperature difference of up to 60 K due mainly to eccentric relocation of pellet fragments.     

Molten fuel radial motion and cladding melting during a PCM event in LWRs

The thermal behavior of the fuel and cladding during off-normal operating conditions, generally termed power-cooling-mismatch (PCM), are of great interest to light water reactor (LWR) safety analysis. During a power-cooling-mismatch event, fuel melting may begin at the center of the rods propagating radially outward. The induced pressure at the center of the rod due to fuel melting, fission gas release, and UO₂ fuel vapor would tend to force such molten fuel to flow through radially open cracks in the outer unmelted portion of the pellet and relocate in the fuel-cladding gap. The zircaloy cladding, which is at high temperature during film boiling, may undergo melting at its inside surface upon being contacted by the extruded molten fuel, eventually resulting in a thermal failure of the cladding.Three topics of interest are analyzed in this paper. First, fuel conditions during a hypothesized PCM accident are assessed with regard to pellet cracking and central fuel melting. Secondly, the transient freezing of a superheated liquid penetrating an initially empty crack, maintained at constant subfreezing temperatures, is studied analytically. The analysis is presented in a dimensionless form, illustrating the effect of the governing parameters, namely the driving pressure, crack shape (that is, a divergent, a parallel wall, or a convergent crack), density ratio, Stefan number for freezing, and steady state crust thickness. The calculational results are used to assess the radial extrusion of molten UO₂ fuel observed in some in-pile tests, in which PCM conditions in a pressurized water reactor were simulated. Thirdly, conditions for potential melting of zircaloy cladding upon being contacted by the extruded molten fuel are investigated analytically. The analytical predictions were consistent with the experimental results from PCM in-pile tests.     

Influence of pellet length on interaction between fuel and cladding in simulated reactor fuel pins

An electrically heated fuel pin test apparatus has been developed for out-of-pile investigations of fuel pin parameters with a view to supplementing in-pile experiments. Sixty per cent of reactor heat ratings has been achieved with a hollow pin having an axially located electrical heater, the limitation being the melting of the UO₂ pellets. The theoretical unconstrained shapes of a heated pellet and a fuel pellet under elastic conditions were calculated. Both showed an ‘hour glass’ form suggesting that permanent circumferential ridges would occur in the cladding of a heated pin as they do in the cladding of fuel pins. These ridges were subsequently produced in heated pins, the pins being heated while immersed in cooling water at typical reactor temperatures and pressures. From a series of such tests using different pellet lengths it was found that a significant reduction in ridge height occured when the pellet ratio was one-third of the value in a typical reactor. The temperatures reached in the UO₂ pellets were estimated from a metallographic examination of a pin cross section after test. Using published data of ∫kdT for UO₂ over various temperature ranges the pin heat output at that cross section was determined.     

Fabrication of Sintered Annular Fuel Pellet for HANARO Irradiation Test

The fabrication method of an annular pellet with highly precise diametric tolerances, same dimensions, and various sintered densities has been investigated. To examine the in-pile densification and swelling of the annular pellet, 5 different types of annular pellet were prepared for a HANARO irradiation test. In order to obtain annular fuel pellets with the same dimensions and various sintered densities, we control the green density of an annular compact, the sintering temperatures, and the sintering time. For a diametric tolerance control, we have introduced a new compaction process that combines the usual double-acting pressing and cold isostatic pressing. Annular fuel pellets with the same dimensions and various sintered densities were fabricated successfully, and all the pellets satisfied the pellet specification of the HANARO irradiation test. Sintered annular pellets show an excellent inner diametric tolerance of less than ±12 μm without an inner surface grinding.