乏燃料燃耗测量装置的研制

正乏燃料组件燃耗整体测量装置采用3种燃耗分析方法,分别是高分辨率γ能谱法、总γ法、总中子法。高分辨率γ能谱法使用1台高纯锗探测器对乏燃料组件的中心位置进行测量,通过测量得到的~(134)Cs、~(137)Cs等相关核素的信息,计算乏燃料组件中心位置的燃耗。总γ法采用多个电离室组成探测器阵列对单根乏燃料组件进行整体测量,测量结果反映乏燃料组件发出的γ射线强度在组     

基于计算关系曲线的无源中子燃耗测量方法的研究

采用燃耗信任制技术可显著提高乏燃料贮存及运输的经济性,也是国际上该领域的发展趋势。非破坏性燃耗测量是采用燃耗信任制技术必须解决的关键问题之一。在非破坏性燃耗测量方法中,基于计算关系曲线的无源中子燃耗测量方法可以精确地测量乏燃料组件的平均燃耗,结合总γ方法,还可以测量出乏燃料组件的末端燃耗。根据该方法的基本原理,在调研分析的基础上,确定了燃耗测量分析方法及其流程。其中,确定乏燃料燃耗与中子发射强度之间的关系、中子发射强度计算方法以及Keff的快速计算方法是测量分析方法的关键技术。     

燃料管相对燃耗测量技术研究

为了准确测量燃料管燃耗和验证燃料管运行的正确性,保证反应堆安全运行、提高燃料的利用率和经济性,针对大直径燃料管相对燃耗测量存在的问题,根据燃料管结构设计了燃料管γ扫描测量系统。首先,用已知活度的10 mm 137Cs标准点源验证了装置的可靠性和方法的可行性。然后,选取了同一燃料组件3种不同直径燃料管进行γ扫描,分别测量了燃料管径向和轴向相对燃耗分布,根据径向相对燃耗分布确定了燃料管阴阳面。最后,在燃料管阳面最大燃耗处进行燃料管轴向γ扫描,获得燃料管轴向相对燃耗分布,并对比各层燃料管相对燃耗大小。通过本方法确定的燃料管相对燃耗测量相对误差小于4%,实现了燃料管相对燃耗的精确测量,为其它类型燃料元件相对燃耗测量提供了一种新思维。     

核电站乏燃料组件燃耗测量系统的改进和大亚湾乏燃料组件燃耗测量

本工作对叉形探测器乏燃料测量系统进行了改进。改进的乏燃料组件测量系统仍包容了总γ、高分辨γ和无源中子3种测量方法。系统的改进涉及以下4个方面。1)改进了探测体对乏燃料组件的测量几何,使总γ、高分辨γ和无源中子对乏燃料组件的4个面均能进行均匀测量,可测得组件燃耗的     

用质谱法测量反应堆元件的燃耗

文章介绍了用质谱计测量反应堆元件的燃耗的基本方法,铀同位素比值法和燃耗监测核素~(148)Nd含量测定法的测量误差分别是(3.0—4.0)％和(1.6—2.0)％。用质谱法获得的燃耗值与用γ谱法和放化法获得的燃耗值是一致的。 文章给出了铀同位素比值、α_5值和燃耗值沿反应堆元件轴向的分布曲线,铀同位素比值与燃耗值的关系曲线。铀同位素比值和燃耗值的关系曲线是线性的,它与裂变生成的毒物核素~(149)Sm同燃耗的关系曲线是一致的。     

压水堆核电厂乏燃料组件γ源强研究

本文运用ORIGEN-ARP计算研究了压水堆核电厂反应堆平衡循环的乏燃料组件的γ源强,对影响γ源强的因素,包括总燃耗、各循环燃耗比例和能群结构划分方式进行了分析。分析结果表明:乏燃料组件中,裂变产物产生的γ源强始终占主要部分。在卸料后的不同冷却时刻,γ总源强与总燃耗或末端燃耗密度存在正比关系。采用不同γ能群结构划分方式对γ总源强计算结果的影响较大。     

10MW高温气冷堆的燃耗测量研究

10MW高温气冷堆的燃耗测量系统是采用非破坏性高纯锗γ谱仪在线监测来确定燃耗值,利用MCNP4A程序对测量系统的衰减因子进行计算,基于核燃料裂变核素的γ射线能谱分析,以137Cs和134Cs核素活度作为测量对象,并对燃耗测量结果进行讨论.     

基于二次辐照的CMRR乏燃料组件燃耗测量技术研究

由于研究堆堆芯装载灵活多变、运行模式复杂,传统的燃耗无损检测技术精度不高。基于乏燃料二次辐照的燃耗测量技术具有不依赖于乏燃料组件的运行历史数据、测量精度较高的优点。本文研究了该方法中裂变产物来源甄别技术,建立了燃耗测量原理装置,分析了装置相对测量效率,完成了中国绵阳研究堆（CMRR）典型乏燃料组件的燃耗测量实验。测量结果表明,对于燃耗为15%左右的乏燃料组件,235U质量的测量不确定度好于5%。      

CENTER燃料组件堆内辐照考验燃耗计算研究

为验证中国工程试验堆（CENTER）燃料组件设计,在燃料组件正式定型前需开展组件辐照考验,CENTER燃料组件在高通量工程试验堆（HFETR）内采用随堆辐照方式进行辐照考验。根据CENTER燃料组件特点,开展了HFETR辐照考验CENTER燃料组件燃耗计算方法研究,确定了CENTER燃料组件辐照考验堆芯物理计算采用镶嵌耦合方法。结果表明,燃料组件平均燃耗计算值与测量值偏差为3.25%,满足辐照考验要求。      

西安脉冲堆燃料元件燃耗无损实验测量

介绍了西安脉冲堆燃料元件燃耗测试理论、所用设备及实验方法,重点阐述了燃耗测量中用到的中子/γ自吸收校正、127 Cs活度校正、探测效率标定等关键技术.对堆芯内两根具有代表性的燃料棒D5、G14开展了燃耗测量,并工理论计算结果进行了比较.结果表明:燃耗测量结果与理论估算值在不确定度范围内一致,该燃耗测试分析系统是可靠、适用的.     

中国实验快堆某辐照实验组件燃耗分布测量

在中国实验快堆（CEFR）上建立了实验组件燃耗分布测量的实验装置。对CEFR某一辐照实验组件中的4#及6#燃料元件棒进行了相对燃耗分布的测量,并与理论计算结果进行了比较。结果表明：两根燃料元件棒虽处于实验组件的不同位置,但相对燃耗分布基本一致;燃耗分布的实验测量结果与理论计算结果符合较好;实验组件燃耗分布测量的相对误差在10.2%以内。本文工作为开展快堆乏燃料组件燃耗测量奠定了基础。     

Characteristics of a Geological Disposal System for the Increasing Burn-up of Spent Nuclear Fuel in Korea

The characteristics of a geological disposal system that can accommodate increasingly higher burn-up levels of spent fuel were assessed based on the Korea reference disposal system concept. First, a status investigation that included a projection of spent fuel quantity versus burn-up was carried out to demonstrate the trend toward higher burn-up levels. Next, the main features of the Korea reference disposal system were introduced. Finally, the disposal tunnel length, excavation volume, and raw materials (e.g., a cast insert, copper, bentonite and backfill) necessary for a disposal system were comprehensively analyzed to define the characteristics and overall effects on geological disposal at increasingly higher burn-up levels. Our study determined that it is reasonable to use a canister containing 4 spent fuel assemblies with burn-up levels up to 50GWD/MTU, while a canister containing 3 spent fuel assemblies can accommodate burn-up levels beyond 50GWD/MTU. A remarkable increase of 33% in disposal tunnel length and that of 30% in excavation volume were observed as the burn-up increased from 50 to 60GWD/MTU. However, this was offset by a reduction of 17% in raw materials used in canister fabrication. Therefore, it seems that spent fuel at increasingly higher burn-up levels is not a serious concern for deep geological disposal in Korea.     

Expansion of the capabilities of the GA-4 legal weight truck spent fuel shipping cask

Abstract

General Atomics has developed the model GA-4 legal weight truck spent fuel cask, a high-capacity cask for the transport of four pressurised water reactor (PWR) spent fuel assemblies, and obtained a certificate of compliance (CoC, No. 9226) in 1998 from the US Nuclear Regulatory Commission (NRC). The currently authorised contents for this CoC, however, are much more limiting than the actual capability of the GA-4 cask to transport spent PWR fuel assemblies. The purpose of this paper is to show how the authorised contents can be significantly expanded by additional analyses without any changes to the physical design of the package. Using burn-up credit as outlined in US NRC Interim Staff Guidance 8, Revision 2, the authorised contents can be significantly expanded by increasing the maximum enrichment as the burn-up increases. Use of burn-up credit eliminates most of the criticality imposed limits on authorised package contents, but shielding still limits the use of the cask for higher burn-up, short-cooled fuel. By reducing the number of assemblies transported (downloading) to two and using shielding inserts, even high-burn-up fuel with reasonable cooling times can be transported.     

TRIGA fuel burn-up calculations and its confirmation

The Cesium (Cs-137) isotopic concentration due to irradiation of TRIGA Fuel Elements FE(s) is calculated and measured at the Atominstitute (ATI) of Vienna University of Technology (VUT). The Cs-137 isotope, as proved burn-up indicator, was applied to determine the burn-up of the TRIGA Mark II research reactor FE. This article presents the calculations and measurements of the Cs-137 isotope and its relevant burn-up of six selected Spent Fuel Elements SPE(s). High-resolution gamma-ray spectroscopy based non-destructive method is employed to measure spent fuel parameters. By the employment of this method, the axial distribution of Cesium-137 for six SPE(s) is measured, resulting in the axial burn-up profiles. Knowing the exact irradiation history and material isotopic inventory of an irradiated FE, six SPE(s) are selected for on-site gamma scanning using a special shielded scanning device developed at the ATI. This unique fuel inspection unit allows to scan each millimeter of the FE. For this purpose, each selected FE was transferred to the fuel inspection unit using the standard fuel transfer cask. Each FE was scanned at a scale of 1 cm of its active length and the Cs-137 activity was determined as proved burn-up indicator. The measuring system consists of a high-purity germanium detector (HPGe) together with suitable fast electronics and on-line PC data acquisition module. The absolute activity of each centimeter of the FE was measured and compared with reactor physics calculations. The ORIGEN2, a one-group depletion and radioactive decay computer code, was applied to calculate the activity of the Cs-137 and the burn-up of selected SPE. The deviation between calculations and measurements was in range from 0.82% to 12.64%.     

吸收光谱法直接测定乏燃料溶解液及1AF料液中Pu(Ⅵ)的含量

钚是乏燃料后处理过程最重要的产品。乏燃料溶解液和1AF料液中Pu(Ⅵ)的含量影响钚的收率,因而需要准确测量。采用吸收光谱法研究建立了1AF料液中Pu(Ⅵ)的分析方法,方法检测下限为5.8 mg/L,两次重加回收率分别为103%和96%,采用燃耗为45 000 MWd/t(以U计)的乏燃料溶解液和1AF料液进行了总钚含量测量方法的验证,测量结果与混合K边密度计-X射线荧光法测量结果吻合,相对偏差不大于3%。     

Static analytical and experimental research of shock absorber to safeguard the nuclear fuel assemblies

The Ignalina Nuclear Power Plant (NPP) has two RBMK-1500 graphite-moderated boiling water multi-channel reactors. The Ignalina NPP Unit 1 was shutdown at the end of 2004, while Unit 2 is foreseen to be shutdown at the end of 2009. At the Ignalina NPP Unit 1 remains approximately 1000 spent fuel assemblies with low burn-up depth. A special set of equipment was developed to reuse these assemblies in the reactor of Unit 2. One of most important items of this set is a container, which is used for the transportation of spent fuel assemblies between the reactors of Unit 1 and Unit 2. A special shock absorber was designed to avoid failure of fuel assemblies in case of hypothetical spent fuel assemblies drop accident during uploading/unloading of spent fuel assemblies to/from container. This shock absorber was examined by using scaled experiments.The objective of this article is the estimation whether the proposed design of shock absorber fulfils the function of the absorber and the optimization of its geometrical parameters using the results of the performed investigations. Static analytical and experimental investigations are presented in the article. The finite element code BRIGADE/Plus was used for the analytical analysis. The calculation model was verified by comparing the experimental investigation and simulation results for further employment of this finite element model in the development of an optimum design of shock absorber. Static simulation was used to perform primary optimization of design and dimension of the shock absorber.     

Implementation of burn-up credit approach for transport and storage cask

Abstract

TN International currently uses burn-up credit methodology for the design of casks dedicated to the transport of pressurised water reactor uranium oxide spent fuel assemblies. As long as the fuel enrichment of the pressurised water reactor fuel assemblies was sufficiently low, a burn-up credit methodology based on the sole consideration of actinides and the use of a partial burn-up was satisfactory to cover the needs without necessity to design new casks. Nevertheless, the continuous increase in the fuel enrichment during the last decade has led TN International to continue the investigations on the burn-up credit methodology to limit both the increase in the neutron poison content in the new basket designs and the burn-up constraints attached to the acceptability of the fuel assemblies for transport. The strategy of TN International was then to take benefit of the large negative reactivity reserves, which might be gained by the consideration of the fission products coming from the fuel irradiation. A big step forward has recently been reached by TN International on this field with the definition of an advanced burn-up credit methodology based on the consideration of relevant fission products recommended by OECD. In the meantime, TN International has taken the opportunity to use such burn-up credit approach in the design of the TN 24 E transport and storage cask developed for the German nuclear power plants. The relevant task has been carried out according to the German standard DIN 25712 for burn-up credit application. The present paper will describe the basic principles of the burn-up credit methodology implemented by TN International such as:

(i) the current state of the art concerning the burn-up credit application in the criticality assessment

(ii) the basic approach used for the implementation of the advanced burn-up credit methodology (bounding axial burn-up profiles, fuel irradiation parameters, fission products, etc.)

(iii) the area of validity of the TN International burn-up credit approach with fission products

(iv) example of application of the burn-up credit methodology for the design of the TN 24 E transport and storage cask under licensing in Germany

(v) the perspectives of development of the burn-up credit methodology.     

核燃料的燃耗测量方法综述

介绍了非破坏性分析和破坏性分析的燃耗测量现状，论述了包括使用γ谱仪和中子探测等的多探头的乏燃料测量系统是燃耗测量的发展趋势。     

Time dependent burn-up and fission products inventory calculations in the discharged fuel of the Syrian MNSR

The Syrian Miniature Neutron Source Reactor (MNSR), a 30 kW, 89.8% HEU fueled (U-Al), went critical in March, 1996. By operating the reactor at nominal power for 2.5 h/day, the estimated core life is 10 years. This paper presents the results of fuel burn-up and depletion analysis of the MNSR fuel lattice using the ORIGEN 2 code. A one-group cross-section data base for the ORIGEN 2 computer code was developed for the Syrian MNSR research reactor. The ORIGEN 2 predicted burn-up dependent actinide compositions of MNSR spent fuel using the newly developed data base show a good agreement with the published results in the literature. In addition, the burn-up characteristics of MNSR spent fuel was analyzed with the new data base. Finally, to study the effect of burn-up on the reactivity, the microscopic cross-sections of the fission products calculated by the WlMS code (using the number densities of fission products generated by the ORIGEN 2 code as a function of burn-up time), were used as an input for the CITATION code calculations. The results contained in this paper could be used in performing criticality safety analysis and shielding calculations for the design of a spent fuel storage cask for the MNSR core.     

Mechanical safety analysis for high burn-up spent fuel assemblies under accident transport conditions

Abstract

Transport packages for spent fuel have to meet the requirements concerning containment, shielding and criticality as specified in the International Atomic Energy Agency regulations for different transport conditions. Physical state of spent fuel and fuel rod cladding as well as geometric configuration of fuel assemblies are, among others, important inputs for the evaluation of correspondent package capabilities under these conditions. The kind, accuracy and completeness of such information depend upon purpose of the specific problem. In this paper, the mechanical behaviour of spent fuel assemblies under accident conditions of transport will be analysed with regard to assumptions to be used in the criticality safety analysis. In particular the potential rearrangement of the fissile content within the package cavity, including the amount of the fuel released from broken rods has to be properly considered in these assumptions. In view of the complexity of interactions between the fuel rods of each fuel assembly among themselves as well as between fuel assemblies, basket, and cask body or cask lid, the exact mechanical analysis of such phenomena under drop test conditions is nearly impossible. The application of sophisticated numerical models requires extensive experimental data for model verification, which are in general not available. The gaps in information concerning the material properties of cladding and pellets, especially for the high burn-up fuel, make the analysis more complicated additionally. In this context a simplified analytical methodology for conservative estimation of fuel rod failures and spent fuel release is described. This methodology is based on experiences of BAM acting as the responsible German authority within safety assessment of packages for transport of spent fuel.