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。     

压水堆燃料棒裂变产物释放产生比计算方法

为对压水堆一回路源项进行准确分析,在研究了美国核管制委员会(Nuclear Regulatory Commission, NRC)发布的计算气体裂变产物释放产生比(Release to Birth Ratio, R/B)的ANSI/ANS-5.4标准及法国METEOR 1.5程序计算气体裂变产物释放产生比的方法后,分别利用几种方法计算了燃料棒在不同功率水平以及不同燃耗下的几种气体裂变产物的释放产生比,对几种方法进行了研究分析。结果表明,燃料棒的裂变产物释放产生比会随着其功率以及燃耗而增大,相较于美国ANSI/ANS-5.4-2011标准,法国METEOR 1.5程序方法更为保守,而NRC原有的ANSI/ANS-5.4-1982标准最为保守,计算出的释放产生比最大,ANSI/ANS-5.4-2011标准能较好地适用于压水堆核电站裂变产物释放份额的计算。     

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。     

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.     

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.     

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.     

Modification of PROMETHEUS Reactor as a Fusion Breeder and Fission Product Transmuter

This study presents the analyses of the fissile breeding and long-lived fission product (LLFP) transmutation potentials of PROMETHEUS reactor. For this purpose, a fissile breeding zone (FBZ) fueled with the ceramic uranium mono-carbide (UC) and a LLFP transmutation zone (TZ) containing the ⁹⁹TC and ¹²⁹I and ¹³⁵Cs isotopes are separately placed into the breeder zone of PROMETHEUS-H design. The neutronic calculations are performed by using two different computer codes, the XSDRNPM/SCALE4.4a neutron transport code and the MCNP4B Monte Carlo code. A range of analyses are examined to determine the effects of the FF, the fraction of ⁶Li in lithium (Li) and the theoretical density (TD) of Li₂O in the tritium breeder zone (TBZ) on the neutronic parameters. It is observed that the numerical results obtained from both codes are consistent with each other. It is carried out that the profiles of fission power density (FPD) are flattened individually for each FF (from 3 to 10%). Only, in the cases of FF ≥ 8%, the system is self sufficient from the point of view of tritium generation. The results bring out that the modified PROMETHEUS fusion reactor has capabilities of effective fissile breeding and LLFP transmutation, as well as the energy generation.     

核裂变产物对地表环境Ba同位素体系的影响

近年来,钡(Ba)同位素研究在地表生态环境领域的应用越来越广泛。然而,上世纪的核试验和核事故泄露在地表环境中释放了大量含有人工核素Ba的裂变产物,它们可能会改变地表自然样品的Ba同位素特征,从而不利于Ba同位素研究在该领域的发展和应用,因此需要评估其影响程度。本文根据Ba同位素最新研究进展和国际上已发表的数据,总结不同地表储库中Ba同位素的分布特征,利用核裂变理论计算核试验裂变产物的Ba同位素组成及裂变成因Ba的加入对环境样品Ba同位素组成的改变程度。结果显示,自然样品加入占其Ba含量0.01%的裂变成因Ba,能显著改变它的Ba同位素组成,因此需要引起重视。此外,基于计算结果发现,同时测量样品δ^137/134Ba和δ^138/134Ba值,根据其异常特征可以有效地排除受到沾染的样品。     

模块式小型堆MAAP建模及严重事故裂变产物释放特性研究

建立模块式小型堆(ACP100)严重事故分析程序(MAAP)电厂模型,并经过稳态调试,各稳态运行参数与设计参数的误差在1%左右,表明所建立模型准确度较高。采用所建MAAP模型对ACP100严重事故进行了模拟,给出了事故进程及裂变产物向环境释放变化趋势,结果表明:在安全壳保持完整的条件下,惰性气体向环境的累积释放份额与时间成线性关系,随时间增加而增大;其他元素组向环境的累积释放份额在一段时间后达到最大,之后保持不变。该分析结果为ACP100严重事故条件下放射性释放和场外剂量分析奠定了基础。