空间离散对弥散颗粒燃料燃耗计算的影响

单栅元燃耗计算是全堆芯燃耗计算的基础,栅元空间离散对燃耗计算的结果有显著影响。弥散颗粒燃料由于双重非均匀性的存在,空间离散的情况更为复杂。本文基于ALPHA组件程序,分析了颗粒在平源区上归类的宏观离散方案与颗粒内部细分燃耗区的微观离散方案对弥散颗粒燃料燃耗计算的影响。算例包括无毒物的UC颗粒单栅元,含Gd₂O₃层的QUADRISO颗粒单栅元和含UC颗粒与Gd₂O₃毒物颗粒的双颗粒单栅元。数值结果表明,无毒物栅元宏观需分3圈以上,含Gd₂O₃栅元宏观需分5圈以上;无毒物算例微观不需要分圈,含Gd₂O₃层的QUADRISO颗粒需在微观燃料区细分2圈,双颗粒问题的Gd₂O₃毒物颗粒微观需分12～15圈。     

高燃耗下核燃料颗粒内部的应力状态及关键影响因素研究

高燃耗是先进反应堆堆芯的发展方向,高燃耗下核燃料内部微结构的精细化建模是燃料强度分析和性能评估的基础。本研究开发了高燃耗燃料颗粒微结构的自动化建模和力学计算程序,综合考虑燃料颗粒内气孔尺寸、位置的非均匀分布特征,系统分析了基体材料的力学性能、燃料颗粒间距、运行环境静水压力及裂变碎片损伤层对燃料颗粒开裂行为的影响规律。结果表明：燃料颗粒间距越大,燃料颗粒越不易开裂;裂变碎片损伤层的存在使得燃料颗粒开裂风险小幅增大;燃料颗粒内气孔尺寸、位置分布的非均匀性,会导致燃料颗粒从多处开裂,且颗粒在外层开裂的概率更大;开裂危险区普遍具有气孔尺寸较大且大气孔串联的特征;基体材料对燃料颗粒表面作用的约束压应力具有较大的波动性,但应力均值随燃料元件所受静水压力的增大而近乎线性增大;增大弥散燃料基体材料的弹性模量可在一定程度上抑制燃料颗粒的开裂行为;燃料元件所受静水压应力越大,燃料颗粒越不易开裂,而燃料颗粒间距缩小能削弱环境压应力的影响程度。本工作为高燃耗条件下弥散燃料安全评估及优化设计提供了分析方法及数值参考。     

基于弥散燃料颗粒开裂的裂变气体释放模型

根据弥散燃料颗粒开裂后裂变气体的3种释放途径,分别建立了裂纹连通释放模型、气泡连通释放模型以及原子扩散释放模型,综合得到了基于弥散燃料颗粒开裂的裂变气体释放模型,并采用该模型对裂变气体释放量进行了计算。结果表明:裂变气体释放量主要由裂纹连通释放途径贡献;燃耗深度越高,裂变气体释放量的增加速率会越大;随着退火温度的增加,裂变气体释放量迅速增加,而退火时间越长,裂变气体释放量的增加速率越低。通过裂变气体释放量模型计算得到的裂纹宽度与实验观察到的裂纹宽度符合较好,对比结果验证了基于弥散燃料颗粒开裂的裂变气体释放模型的合理性。      

韩国的高燃耗燃料技术

高燃耗燃料开发是韩国的国家级研发项目,涉及包壳、UO2芯块、定位格架、性能评估程序和燃料组件测试等关键技术领域。经过合金成分设计、制管、堆外和堆内试验,新的包壳合金被开发出来。由于新合金采用了优化的成分和热处理工艺,这种新型Zr-Nb管的抗腐蚀能力和蠕变强度远优于ZrO4管。氧化铀芯块的大晶粒制造技术使用了不同氧化铀晶种和微量掺杂的铝。通过对氧化铀芯块的碎片进行氧化和热处理获得UO2晶种,然后将其加入UO2粉末中。通过制造含有W通道的UO2芯块来提高热导率。为了分析燃料性能,建立了新的高燃耗模型和相应的程序。该程序经过了国际数据库和我们自己数据库的验证。开发的定位格架有波浪形接触弹簧和搅混翼。力学和水力测试表明该格架在棒支撑、耐磨性、临界热流密度等方面都有良好表现。最后,研发了燃料组件测试技术,建造了力学和热工水力测试装置,已投入使用。不久以后,这些成果都将被用于韩国核燃料计划,提高韩国压水堆的安全性和经济性。     

国外压水堆高燃耗燃料发展现状及前景

许多国家正在大力发展压水堆高燃耗燃料,其批平均目标卸料燃耗能够达到5.0—5.5万兆瓦天/吨铀,因此显著地提高了铀利用率。高燃耗燃料即将陆续取代现行的燃料,这标志着压水堆燃料的更新换代。本文简要介绍高燃耗燃料的设计特点、燃料管理、经济效益、发展计划及发展前景等。     

高燃耗燃料元件易引发卡棒事件

一、事件的基本情况据资料报道 ,美、法、德、日等国都在积极地开展压水堆核电站燃料的高燃耗方面的研究试验工作。其目的是延长燃料元件使用寿命、提高铀燃料利用率和降低燃料循环成本 ,进而有利于降低核电成本。它也是在核燃料一次通过反应堆前提下的一种有效的节铀途径。有资料表明 ,如果压水堆燃料元件燃耗由 3.3万兆瓦日 /吨铀提高到 5万兆瓦日 /吨铀 ,并采用 1 2个月的换料周期 ,就可节约铀 1 3- 1 6 %。近几年国外研究结果认为 ,压水堆卸料燃耗达到 4 .5万兆瓦日 /吨铀时已没有多大的破损风险。据报道 ,1 994至1 9 96三年中 ,当压水…     

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。      

CF燃料棒高燃耗下的性能评价

本文首先介绍了FUPAC软件中与N36合金包壳相关的计算模型,后针对高燃耗情况下,采用FU-PAC软件对CF燃料棒的性能进行理论分析,结果表明:在CF燃料棒燃耗达到60 000 MWd/t,寿期内I、Ⅱ类运行工况下,其结构完整性能够得到保持.     

高燃耗燃料结构孔隙率的变化

研究了高燃耗燃料结构（HBS）孔隙率的变化。用电子探针（EPMA）和扫描电子显微镜（SEM）对平均棒燃耗约为105GW·d／t（HM）的UO2燃料的测量结果表明：在靠近燃料与覆盖层界面形成了一个超高燃耗结构，局部燃耗为300GW·d／t（HM）。这种结构的特点足气体气孔尺寸达到了15μm。在HBS靠内部区域发现了大量尺寸在3～5μm的气孔。对孔径分布图进行分析后的结果表明，孔径为3．5μm和7．5μm的气孔占大多数。可以用一个简单的由空位扩散动力学和聚合构成的模型来解释这些现象：初始超压值为50-70MPa就能获得从1μm长大到3．5μm的气孔。那些孔径约为7—8μm的超大气孔是由中等直径的气孔聚合而成的。这基于以下假设：①这些气孔只是轻微的超压；②聚合过程发生在试验方法所能观察到的孔隙度不变的地方。     

Experimental Observations on Fuel Pellet Performance at High Burnup

The Japanese and Spanish nuclear industries have conducted joint experimental programmes since early 1990's to address fuel performance issues such as fuel volume change and fission gas release. These efforts have produced large amount of valuable information on in-reactor performance of fuel materials representing current and potential future fuel designs. A large number of thoroughly characterised fuel rods composed of different materials have been irradiated in the Spanish PWR Vandellós II for up to five irradiation cycles achieving rod average burnup of up to 75 MWd/kgU.

This paper looks into the fuel pellet performance at high burnup only based on the extensive PIE programme both on-site and in hot-cells carried out over this fuel and other related data on similar fuel rods thus supporting and enriching the conclusions.     

Research Program to Elucidate Outside-in Failure of High Burnup Fuel Cladding

Among a series of power ramp tests on 25 Zr-lined segment rods of burnup ranging from 43 to 61 GWd/t, five segment rods failed during the power ramp tests. One segment rod irradiated for 3 cycles (43 GWd/t) failed with a pinhole due to PCI/SCC. The rest of higher burnups failed with an axial crack on the outer surface. The failure threshold power tended to decrease as burnup increases.

Post irradiation examinations revealed increased cladding hydrogen absorption and its precipitates in the cladding outer rim after 4 and 5 cycle irradiations, in contrast to a uniform hydride distribution and a small hydrogen content after 3 cycle irradiation. Metallographic observations suggested an axial crack failure mode induced by the combined effects of high stress and hydrides precipitated in a radial direction during power ramp.

The axial crack failure during the power ramp is supposed to be initiated by a cracking of radial hydride formed by hydride re-distribution and re-orientation at the cladding outer rim and to propagate through a process of hydride concentration and precipitation at the crack tip. Research programs of experimental and analytical studies to clarify the conditions of such mechanism are on-going focusing on the hydrogen behavior and mechanical performance of the irradiated cladding.     

Post Irradiation Examination of High Burnup PWR Fuel

Post irradiation examination (PIE) of a high burnt lead fuel assembly, which was irradiated to demonstrate fuel integrity at high burnup, was performed before the start of the full batch loading of high burnup fuel of 48GWd/t maximum fuel assembly burnup.

The lead fuel assembly was 17×17 B-type PWR fuel which was supplied by Nuclear Fuel Industries, Ltd. (NFI) and achieved the maximum burnup of 45 GWd/t after 4 cycles of irradiation in Ohi Unit 1 of the Kansai Electric Power Co. Inc. (Kansai).

Twelve fuel rods extracted from the lead fuel assembly at the reactor site were examined at the hot-cell facility of Japan Atomic Energy Research Institute (JAERI) in Tokai-mura.

Visually, the fuel rods appeared to be in good conditions, but some small spallings were observed at the second span from the top where oxide film was relatively thicker than other spans. Even in this span, the maximum oxide film thickness was less than 50 μm Fission gas release rate was less than 1%, which caused only a small increase in fuel rod internal pressure. Mechanical properties of the fuel cladding were evaluated by tensile tests.

These PIE results were within the range of other PIE data previously obtained from domestic and foreign PWR fuel rods. The data confirmed that the integrity of B-type fuel would be maintained at least up to 48 GWd/t.     

Burnup Determination of High Burnup and Dry Processed Fuels Based on Isotope Dilution Mass Spectrometric Measurements

Destructive methods were used for the burnup determination of a PWR nuclear fuel irradiated to a high burnup in power reactors, and of a dry processed fuel fabricated from a spent PWR fuel and irradiated in the Hanaro research reactor. The total burnup was determined from a measurement of the Nd and Cs isotope burnup monitors. The methods included U, Pu, ¹⁴⁸Nd, ¹⁴⁵Nd+¹⁴⁶Nd, total of the Nd isotopes, ¹³³Cs and ¹³⁷Cs determinations by the isotope dilution mass spectrometric method (IDMS) by using quadrupole spikes (²³³U, ²⁴²Pu, ¹⁵⁰Nd, and ¹³³Cs). The methods involved two sequential anion exchange resin (AG 1X8 and 1X4) separation procedures and a Cs purification with a cation exchange resin (AG 50WX4) separation procedure. The results obtained by the Nd and Cs isotopes from the mass spectrometric measurement were compared with those by the ORIGEN code.     

Oxygen Potentials of UO2 Fuel Simulating High Burnup

The oxygen potentials at 1,000 and 1,300°C and the lattice parameters of UO₂ fuels with soluble fission product elements (Zr, Ce, Pr, Nd, Y), simulating high burnup of up to 10^a,_o have been measured by means of thermogravimetry and X-ray diffraction. The oxygen potentials for (U, FP)O₂+x fuels are higher than pure UO_2+x; at a given composition and increase positively with increasing simulated burnup. They can be represented as a function of the mean uranium valence at compositions of 0/M>2.01. The lattice parameters of stoichiometric (U, FP)0₂.00 fuels decrease linearly with simulated burnup, and can be expressed as a (pm) = 547.02–0.1225, where B is burnup in ^a._o     

Measurement of Burnup in FBR MOX Fuel Irradiated to High Burnup

The burnup of fuel pins in the subassemblies irradiated at the range from 0.003 to 13.28%FIMA in the JOYO MK-II core were measured by the isotope dilution analysis. For the measurement, 75 and 51 specimens were taken from the fuel pins of driver fuel and irradiation test subassemblies, respectively. The data of burnup could be obtained within an experimental error of 4%, and were compared with the ones calculated by 3-dimensional neutron diffusion codes MAGI and ESPRIT-J, which are used for JOYO core management system.

Both data of burnup almost agree with each other within an error of 5%. For the fuel pins loaded at the outer region of the subassembly in the 4th row, which was adjacent to reflectors, however, some of the calculation results were 15% less at most than the measured values. It is suggested from the calculation by a Monte Carlo code MCNP-4A that this difference between the calculated and the measured data attribute from the softening of neutron flux in the region adjacent to the reflector.     

乏燃料燃耗与中子发射强度的关系

对辐照过的低富集氧化铀燃料的燃耗与中子发射强度间的关系进行分析和研究。计算了不同初始富集度、不同燃耗、不同冷却时间的乏燃料的中子发射强度,经分析,证实了燃耗与中子发射强度间存在的幂函数关系,并对影响幂函数关系的各种因素进行了研究。发现幂函数关系中的系数受初始富集度、冷却时间、燃耗范围的影响;如果冷却时间大于2a,这个关系不受辐照历史的影响;如果冷却时间小于2a,这个关系受辐照历史的影响。     

基于cosRMC的新一代PWR栅元和组件燃耗计算

新一代压水堆与现有压水堆的重要区别之一是燃料富集度不同,考虑到燃料制造、燃料燃耗等问题,目前压水堆的UO₂燃料富集度通常小于5%,MOX燃料中易裂变Pu含量通常小于6%。新一代压水堆的燃料富集度有可能超过现有标准,平均燃耗有望达到70 GW•d/tU,这对反应堆计算软件提出了新的要求。本文基于反应堆蒙特卡罗程序cosRMC对新一代压水堆栅元和组件基准进行了中子学分析,包括裂变反应率分布、中子通量密度分布及核子密度随燃耗的变化等,并对含Gd棒的组件燃耗计算进行了细致分析。计算结果表明,cosRMC的计算结果与国际上其他程序的计算结果符合较好。通过程序之间结果对比发现,随着燃耗的增加,不同程序计算的Pu含量差别变大。