基于严重事故剂量后果的堆芯重要核素选择

用于事故放射性后果评价的事故源项与堆芯源项密切相关,而通用的堆芯源项程序,如ORIGEN的计算结果包含几百个甚至更多的核素,如何在大量核素中筛选出重要核素,对于简便、准确地进行放射性后果评估具有重要意义。通过对事故后核素迁移至环境的跟踪,评价每个核素对最终放射性后果的贡献。结果显示30～50个核素的剂量贡献即可包络95%以上的剂量后果,这些核素可作为计算的基准,也是堆芯源项核素选择的依据。计算方法对剂量后果评价具有重要的参考价值。     

含MOX燃料堆芯衰变热及裂变产物积存量的特性研究

我国尚无MOX燃料的工程经验,需开展大量的论证工作。以国内M310型堆芯为对象,对使用30%MOX燃料的堆芯燃料管理方案进行分析,比较含MOX燃料堆芯和全UO2堆芯的衰变热、乏燃料水池热负荷和堆芯裂变产物积存量的特性差异。结果表明,堆芯衰变热、乏燃料水池热负荷变化较小,都不会超过原来的15%,堆芯裂变产物积存量中少量核素偏差较大,135Xe、136Cs等超过40%,其余核素变化较小。研究表明,含MOX燃料堆芯的辐射特性存在一定差异,但变化非常有限,不会对电厂的运行和安全产生影响。     

MOX乏燃料衰变热计算方法研究

目前,铀钚混合氧化物(MOX)燃料已成为一种可用于商业核电厂成熟再循环核燃料。经过燃耗过的燃料在正常停堆或事故后停堆时会产生大量的衰变余热,而乏燃料衰变热是事故分析、余热排出系统和乏燃料池冷却系统设计的重要输入参数之一。UOX乏燃料中裂变产物主要来自于U和Pu等可裂变核素的裂变,U贡献最大;MOX乏燃料裂变产物主要来自于U、Pu和Am等可裂变核素的裂变,Pu贡献最大。UOX乏燃料衰变热可使用ANS—5.1的方法进行计算,但ANS—5.1中的衰变热计算方法不完全适用于MOX燃料。MOX燃料是核燃料可持续发展的重要途径,因此必须研究采用新方法计算MOX乏燃料的衰变热。该文研究使用ANS—5.1计算MOX乏燃料裂变产物衰变热,再使用ORIGEN—S程序计算MOX乏燃料的重核衰变热贡献份额,综合得到MOX乏燃料的总衰变热。     

池式研究堆衰变热计算与实验研究

采用量热法测量反应堆额定功率运行75.0 h停堆后45 h内的衰变热功率,拟合出归一化衰变热功率的经验关系式.与反应堆衰变热几种半经验公式和标准对比结果表明,实验结果在经验公式计算值范围内,并与EJ/T 745标准预测值符合较好.     

西安脉冲堆衰变热功率精细化计算

为了给脉冲堆的余热导出数值计算提供更为精细可信的能量源项,通过耦合MCNP程序和ORIGEN2程序,提出了计算方法XAPRDH以及开发了同名程序。方法实现上,首先将每个燃料元件的燃料部分（含控制棒跟随体）以轴向10等分、径向3等分的形式分割为30个独立单元,全堆共形成3180个单元;然后通过灵活调用MCNP程序和ORIGEN2程序获取每个单元的材料组分、核反应截面、中子通量密度和裂变功率,最终实现对各单元衰变热的独立计算和跟踪。分析表明,本文的燃耗评价数据与文献值符合较好,与实验值在测量误差范围内吻合,全堆衰变热计算结果也与行业标准符合,说明本文建立的衰变热精细化计算方法可行,计算结果可信。      

钠冷快堆系统分析程序FR-Sdaso衰变热计算模型开发与验证

反应堆停堆后的余热导出是反应堆的重要安全功能之一,停堆初期余热由裂变功率和衰变热构成,停堆后期余热主要取决于衰变热。本文开发了应用于钠冷快堆系统分析程序FR-Sdaso的衰变热计算模型,该模型可考虑裂变功率和功率历史的影响。通过与ANSI/ANS-5.1-2005标准和SAS4A/SASYS-1程序对比进行了模型验证。FR-Sdaso程序的计算结果与ANSI/ANS-5.1-2005标准的最大相对偏差约为0.1%,与SAS4A/SASYS-1的最大相对偏差在10-8量级,初步证明了所开发模型的正确性。最后,基于中国实验快堆的设计数据,分析了紧急停堆过程中裂变功率对衰变热的影响,结果表明,忽略裂变功率的影响导致衰变热的最大相对偏差约-7%,出现在停堆初期。因此,计算停堆初期衰变热时应考虑裂变功率的影响。     

乏核燃料衰变热的γ射线能谱测量

推荐了一种用γ射线能谱确定乏核燃料残余热功率的方法。该方法是建立在残余热与^137Cs活性之间的相关性的基础上的，对于冷却时间在10到50年的燃料，这种相关性几乎成线性。采用量热法测得的69个沸水堆和压水堆燃料组件的衰变热测量值的可用数据，得到的结果的标准偏差为3％。     

熔岩玻璃体核素衰变热功率的计算

温度是影响熔岩玻璃体溶解速度的关键因素,为此,本文计算了核试验后10~300 000d内熔岩玻璃体中核素衰变热功率,评估了核素衰变热功率对熔岩玻璃体的温度和溶解速度的影响程度。采用了国际原子能机构给出的100kt TNT当量地下核试验产生的、半衰期大于1a的放射性核素含量,利用其中裂变产物核素137 Cs的含量推算累积裂变产额大于0.1%、半衰期为1d~1a的短寿命裂变产物核素的含量。分析了各核素的放射性衰变特点,采用ENDF/BⅦ库中核素衰变辐射的平均α能量、平均电子能量和平均电磁辐射能量计算各核素在熔岩玻璃体内因衰变而沉积的能量。计算结果表明:核素衰变热功率呈分段幂函数衰减;在10~2 000d、2 000~60 000d和60 000d之后的时段内,衰变热功率分别主要源于短寿命裂变产物核素、长寿命裂变产物核素和锕系元素。核素衰变热功率对熔岩玻璃体的温度和溶解速度的影响不大,1 000d后影响非常小。     

从“第五不稳定核素系”展望^238U增强利用的前景

对目前水反应堆中＾２３８Ｕ的利用现状作了剖析，根据作者在以前文章中介绍的“第五不稳定核素系”的性质，提出了使＾２３８Ｕ得到更好利用的可能性，并对可能遇到的问题作了估计。     

Evaluation of Proliferation Resistance of Plutonium Based on Decay Heat

The present paper is prepared for the peaceful use of nuclear energy. Proliferation resistance of plutonium can be enhanced by increasing the decay heat of plutonium. For example, it can be enhanced by increasing the isotopic fraction of ²³⁸Pu, which has the largest decay heat among plutonium isotopes. In the present paper, the proliferation resistance of plutonium was evaluated based on decay heat with a physical assessment model. New criteria were proposed to evaluate the proliferation resistance based on isotopic compositions of plutonium from the viewpoint of decay heat. The present criteria were applied to evaluate the proliferation resistance of plutonium produced in typical Light Water Reactor and Fast Breeder Reactor based on an evaluation function “Attractiveness” as case studies.     

基于衰变链计算的裂变产物缓发γ能谱分析

在军控核查技术中,缓发γ能谱是核材料的“指纹”。为计算和分析铀裂变产物的缓发γ能谱,本文将各种类型的衰变链简化为基态线性链和激发态线性链,推导了零时前后各级核素数目的变化公式,建构了计算缓发γ射线能谱的C语言程序代码,并通过实验对理论推导进行了验证。通过分析几种核素的缓发γ射线计数发现,计算结果与实验数据吻合较好。     

Reconsideration of the Theoretical Supplementation of Decay Data in Fission Product Decay Heat Summation Calculations

In the early stage of the development of the fission product (FP) decay data library for decay-heat summation calculations a serious disagreement between calculations and sample-irradiation measurements was experienced world wide. This problem was essentially circumvented by introduction of the nuclear model calculations. By applying the recent results using the total absorption gamma-ray spectrometer to the decay heat calculations we found that TAGS results seem to be free from the pandemonium problem and, in this respect, it provides a solid basis of the summation calculations without the supplementation of the nuclear model calculations. Among the typical FP nuclides which deserve the future TAGS experiment there are ⁹⁸Nb, ¹⁰⁰Nb, ¹⁰⁵Mo and ¹⁰²Tc and others from the present survey.     

Decay Heat Calculations Based on Theoretical Estimation of Average Beta- and Gamma-Energies Released from Short-Lived Fission Products

Fission product decay heat of ²³⁵U was calculated for short cooling-time on the basis of a preliminary version of a new decay data library recently completed by the Japanese Nuclear Data Committee. It was shown that a full adoption of recent publications of decay schemes to derive average energies of β- and γ-rays, ē_β, and ē_γ, leads to a large underestimation of the γ-ray component of the decay heat and to an overestimation of the β-ray component. In order to remedy this, theoretical values of ē_β and ē_γ were introduced for high Q-value decays, which were obtained with a gross theory of β-decay. It improved remarkably the agreement between calculation and experiment not only for the ²³⁵U decay heat but for ²³⁹Pu and ²⁴¹Pu as well. It was concluded that a large part of decay schemes recently published for high Q-value nuclides are inappropriate to use in calculations of ē_β and ē_γ, because they fail to reproduce the effect of β- strengths at high excitations, which makes ē_β small and ē_γ large. The use of the gross theory introduces this effect correctly into the values of ē_β and ē_γ and, hence, leads to a quite good prediction of both β- and γ-ray components of the decay heat.     

球床式高温气冷堆的余热不确定性分析

反应堆在停堆后相当长时间内仍具有较高的剩余发热是核电站的重要特性,也是核电站安全分析的关键。因此,对反应堆余热及其不确定性进行分析,对于合理设计余热排出系统、研究论证燃料元件在事故后的安全特性等均具有重要意义。本工作结合德国针对球床式高温气冷堆制定的余热计算标准,介绍了球床式高温气冷堆剩余发热及其不确定性的计算方法,并结合200 MWe球床模块式高温气冷堆示范工程（HTR-PM）的初步物理设计,对长期运行在满功率平衡堆芯状态下的反应堆停堆后的余热及其不确定性进行了计算分析,为进一步的事故分析提供依据。     

Effects of interaction between molten zircaloy and irradiated MOX fuel on the fission product release behavior

As a first step for obtaining experimental data on the effects of high-temperature chemical interaction on fission product release behavior, we focused on the dissolution of irradiated uranium plutonium mixed oxide (MOX) fuel by molten zircaloy (Zry) and carried out a heating test under the reducing atmosphere. Pieces of an irradiated MOX fuel pellet and cladding were subjected to the heating test at 2373 K for five minutes. The fractional release rate of cesium (specifically ¹³⁷Cs) was monitored during the test and its release behavior was evaluated. The observation of microstructures and measurements of elemental distribution in the heated specimen were also performed. We demonstrated experimentally that the fuel dissolution by molten Zry accelerated the release of Cs from the fuel pellets.     

非能动先进压水堆核电厂严重事故下裂变产物行为研究

非能动先进压水堆核电厂在严重事故下,安全壳可能发生失效,导致大量放射性物质向环境释放。本文针对非能动先进压水堆核电厂可能发生的早期失效、中期失效、晚期失效三种释放类别,建立百万千瓦级非能动先进压水堆的事故分析模型,分别针对自动卸压系统第二级卸压阀误开启,DVI管线上发生当量直径为4英寸的破口,以及热管段发生当量直径为2英寸的破口的典型严重事故序列,在研究事故进程的基础上,分析事故下裂变产物释放和迁移的特性,重点关注惰性气体、挥发性裂变产物和非挥发性裂变产物在核电厂的分布,最终计算释入环境的裂变产物源项。本文分析结果可为严重事故管理以及厂外放射性后果评价提供支持。