百万千瓦级压水堆严重事故下裂变产物释放分析

严重事故工况下,可能会导致安全壳失效,使大量裂变产物释入环境.本文以百万千瓦级核电厂为对象,利用一体化程序研究不同破口事故叠加全厂断电事故下裂变产物CsI在一回路和安全壳内的质量以及裂变产物向环境释放的源项,并分析安全壳喷淋措施对控制裂变产物向外释放的影响.分析结果表明,小破口事故、中破口事故和大破口事故下释放到环境的...     

百万千瓦级压水堆严重事故下裂变产物释放分析

严重事故工况下,可能会导致安全壳失效,使大量裂变产物释入环境.本文以百万千瓦级核电厂为对象,利用一体化程序研究不同破口事故叠加全厂断电事故下裂变产物CsI在一回路和安全壳内的质量以及裂变产物向环境释放的源项,并分析安全壳喷淋措施对控制裂变产物向外释放的影响.分析结果表明,小破口事故、中破口事故和大破口事故下释放到环境的...     

事故情况下裂变产物从核燃料中的释放

过去研究中源项估计过高,可安全地降到更现实的值。影响裂变产物从核燃料中释放的因素很多,要精确预测裂变产物的释放率是困难的,还需做许多工作。     

压水堆核电厂裂变产物源项计算方法研究

裂变产物作为一回路冷却剂中放射性核素的重要组成部分,在核电厂设计中具有非常重要的意义。文中对堆芯积存量计算模型、燃料包壳内裂变产物向一回路冷却剂释放模型、裂变产物在一回路中的平衡模型进行了分析与研究,并以典型压水堆核电厂为例进行了计算与验证,证实了本文中给出计算模型的合理性以及适用性,可供压水堆核电站裂变产物源项计算分析参考。     

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。     

压水堆嬗变长寿命裂变产物的弹棒事故安全性研究

为分析压水堆(PWR)嬗变长寿命裂变产物(LLFP)的堆芯瞬态安全性,基于CASMO-4、RSIM以及改进的NLSANMT/COBRA-4程序搭建了程序系统,并利用该系统研究了嬗变堆芯在弹棒事故下的安全特性,分析了寿期初和寿期末事故发生后的功率变化及燃料中心温度变化。数值结果表明:与参考PWR相比,装载99 Tc将会使温度系数变得更负,因此弹棒事故下峰值功率降低,而装载129I则相反;装载这两种裂变产物时,燃料中心温度最高可升高127~157℃,仍距UO2芯块熔化限值温度有较大裕量。     

压水堆主回路裂变产物源项计算程序开发

为了开发拥有自主知识产权的核电厂主回路源项计算程序,对裂变产物的生成、释放和迁移等各个过程进行深入研究,建立和完善各个过程的计算模型,形成了一套完整的压水堆主回路裂变产物源项计算方法,在此基础上开发了主回路裂变产物源项计算程序,并进行了初步验证。     

破损燃料组件中裂变产物向冷却剂释放量计算

分析包壳破损情况下裂变产物从燃料芯块向冷却剂的释放机理,建立裂变产物从燃料芯块向冷却剂的释放量的计算模型;采用CPR1000机型的设计参数对燃料包壳破损率、包壳破损尺寸和燃耗开展敏感性分析,计算等效逃脱率系数并与AP1000设计控制文件中给出的逃脱率系数进行比较。结果表明,包壳破损尺寸对裂变产物释放的影响较大,燃耗和包壳破损率对裂变产物释放影响较小。在包壳破口尺寸为34μm时,采用建立的计算模型计算所得部分核素的等效逃脱率系数与AP1000设计控制文件中给出的逃脱率系数极为接近。     

放射性裂变产物由燃料芯块释放到一回路的影响因素研究

采用PROFIP5程序分析燃料棒线功率密度、衰变常数和破口尺寸等对放射性裂变产物释放的影响。结果表明:当燃料棒中心温度低于1000℃时,裂变产物的释放份额与燃料棒温度无关;当燃料棒中心温度高于1000℃时,燃料棒温度越高,裂变产物的释放份额越大;燃料棒线功率密度越高,衰变常数对释放份额的影响越明显。     

反应堆中气体裂变产物释放问题的研究

在对反应堆气体裂变产物的研究中,需要了解作为固体裂变产物子体的氪、氙气体的释放率问题.本文对有较长半衰期前驱物的135Xe的测量数据进行了具体分析,根据在不同取样时间下分析所得的裂变反应事件数量应一致的原则,得到哪种释放模型更符合真实情况的结论.以期能对释放问题给出一个有价值的参考意见.     

Fission Product Model for BWR Analysis with Improved Accuracy in High Burnup

A new fission product (FP) chain model has been studied to be used in a BWR lattice calculation. In attempting to establish the model, two requirements, i.e. the accuracy in predicting burnup reactivity and the easiness in practical application, are simultaneously considered. The resultant FP model consists of 81 explicit FP nuclides and two lumped pseudo nuclides having the absorption cross sections independent of burnup history and fuel composition. For the verification, extensive numerical tests covering over a wide range of operational conditions and fuel compositions have been carried out. The results indicate that the estimated errors in burnup reactivity are within 0.1Δk for exposures up to 100GWd/t. It is concluded that the present model can offer a high degree of accuracy for FP representation in BWR lattice calculation.     

Fission Product Model for BWR Lattice Calculation Code

A benchmark calculation of full fission product was performed for thermal reactor application using an isotope transmutation code DCHAIN based on 185 nuclides with revised nuclear data library. The fission product model for BWR lattice calculation was studied and tested with the benchmark results, and a model containing 45 explicit nuclides and one pseudo nuclide was selected as a reasonably best model to predict the burn up reactivity with high precision for practically all types of fuel and reactor operating conditions. The evaluated thermal cross section and resonance integral for the pseudo nuclide are σ_2,200 = 2.6b and.RI = 10.6b, combined with the pseudo fission yield values of 1.3898, 1.3233, 1.3675 and 1.2773 for fissions from ²³⁵U, ²³⁸U, ²³⁹Pu and ²⁴¹Pu, respectively. The present results are believed as equally applicable to PWR lattice calculation.     

Effects of Fuel Oxidation and Dissolution on Volatile Fission Product Release under Severe Accident Conditions

The release of volatile fission products from high-burnup UO₂ fuel was examined in a steam atmosphere under severe accident conditions as a part of the VEGA program. The effects of fuel oxidation and dissolution were totally evaluated, by comparing the results with those from previous inert, hydrogen and steam atmosphere tests. It was shown that the oxidation of UO₂ to UO_2+x by steam generally enhances Cs and Kr release. However, the enhancement becomes smaller above the melting temperature of Zircaloy, about 2030 K, likely due to reduction of UO_2+x by molten Zircaloy. The burst release of Cs occurs above about 2300K in the hydrogen atmosphere, while the release rate does not increase so significantly for the examined temperature range (<2800 K) in the steam atmosphere. Analysis of the hydrogen atmosphere test showed that fuel dissolution is apparently connected with the burst release and that a large fraction of Cs is quickly released from the dissolved fuel above 2300 K. It is considered that the fuel dissolution rate in the steam atmosphere is about 1/1000 of that in the hydrogen atmosphere.     

Verification of Fission Product Release Model from High Temperature Engineering Test Reactor Fuel

A fuel assembly of the High Temperature Engineering Test Reactor (HTTR) is composed of fuel rods and a hexagonal graphite block. A fuel rod is composed of the fuel compacts and a graphite sleeve. The coated fuel particles are incorporated into a graphite matrix to form a fuel compact. The fuel consists of microspheres of low-enriched U02 with a TRISO coating. The TRISO coatings consist of a porous pyrolytic carbon (PyC) buffer layer followed by an isotropic PyC layer, a SiC layer and a final (outer) PyC layer.

In order to evaluate amounts of fission products released from the HTTR fuel rods during normal operation, analytical models have been developed. Fractional releases of noble gases and iodine are calculated based on release data of ⁸⁸Kr which are obtained by irradiation tests with failed coated fuel particles. The transport of the metallic fission products through the kernel, coatings, fuel compact and graphite sleeve is modeled as a diffusion process. These analytical models have been verified by comparison with measured fractional releases in in-reactor tests and have been concluded to be applicable to the HTTR fuel condition.     

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。     

Calculation of Prompt Fission Product Average Cross Sections for Neutron-Induced Fission of 235U and 239Pu

Yield-weighted average cross sections of neutron radiative capture, (n,2n), and (n,3n) reactions over prompt fission products (FPs) from ²³⁵U and ²³⁹Pu are calculated. The prompt fission production yields are taken from the ENDF/B-VII.0 library. The FPs for each fissile material exist over a range of approximately 1000 neutron-rich nuclides. Several nuclear reaction codes are utilized for calculating the cross sections on each individual fission product—EMPIRE-2.19, TALYS-1.0, GNASH, and CoH. The influence of the FP isomers on the average cross sections is examined with TALYS. We investigate the dependence of the average cross sections on the number of FPs taken for averaging. It is shown that the average capture cross section is much more sensitive to the number of FPs included, compared with the (n,2n) and (n,3n) reactions. An intercomparison of the calculated cross sections with the different reaction codes is carried out. In the capture reaction, EMPIRE predicted lower cross section than TALYS and CoH owing to different default assumptions used in the γ-ray strength function modeling. Moreover, the preequilibrium models implemented in each code give different predictions for the neutron-emission reactions, although the differences are relatively small. We also discuss a difference between the macroscopic and microscopic calculation options in TALYS for the pre-equilibrium model, optical potential model, and γ-ray strength function. The predictive capability of the reaction codes for the capture reaction is examined by comparing their calculations with the ENDF data, which are based on measurements. Compared with the historic Foster and Arthur's evaluation, our new (n,2n) predictions are similar, although our capture predictions are almost an order of magnitude higher. Recommended cross sections for use in applications have been tabulated in ENDF-formatted files.