芯块年龄对燃料棒富集度检测的影响研究

为保证核电站运行安全,反应堆装入核燃料组件的²³⁵U富集度具有严格的设计要求。因此,反应堆燃料棒均必须经过100%富集度及装料均匀性无损检测。已有富集度检测包括有源法和无源法,通过对有源法、无源法两种富集度检测方法中UO₂芯块年龄现象的研究,发现有源检测法在较低活度下,受UO₂芯块年龄影响,使得燃料棒富集度无法正确检测,需采用被动放置等待的方式解决芯块年龄问题。无源法可在检测中直接校正芯块年龄,满足生产检测要求,已得到工程化应用。     

Neutron-physical and thermohydraulic calculations of VVR-SM with high-and low-enrichment uranium fuel assemblies

The VVR-SM reactor at the Institute of Nuclear Physics of the Academy of Sciences of Uzbekistan is being converted from fuel assemblies with high-enrichment uranium (36% ²³⁵U) to fuel assemblies with low-enrichment uranium (19.7% ²³⁵U). During the conversion process consisting of nine cycles, the IRT-3M fuel assemblies with high-enrichment uranium, which are removed at the end of each cycle, will be replaced with IRT-4M fuel assemblies with low-enrichment uranium. This will require increasing the core size up to 20 fuel assemblies and increasing the power of the reactor to 11 MW. The methods used for and the results of neutron-physical calculations (burnup, power distribution, subcriticality), thermohydraulic analysis, and calculations of the kinetic parameters of a stable state are described for a core with high-enrichment uranium, a mixed core, and the first full core with low-enrichment uranium. __________ Translated from Atomnaya énergiya, Vol. 104, No. 5, pp. 269–273, May, 2008.     

通过分析放射性核素活度鉴别燃料初始富集度的可行性研究

高温气冷堆核电站示范工程（HTR-PM）的反应堆在达到平衡状态前要经过一个较长时间的过渡过程。该过程中堆芯将装入两类燃料球,它们在设计上只有燃料初始富集度不同。反应堆运行要求在过渡过程中要鉴别出装有低富集度燃料的燃料球,并按其燃耗水平不同将其卸出。本文针对此问题,讨论了通过分析燃料球中放射性核素活度（或其比值）以鉴别两类燃料球的方法。堆物理分析软件和KORIGEN软件针对过渡过程的计算结果初步肯定了该方法的理论可行性,并可看出最有可能的鉴别指征量是¹³⁴Cs活度、¹²⁵Sb与¹³⁷Cs的活度比值和¹³⁴Cs活度与¹³⁷Cs活度平方的比值。     

CZT探测器测量核燃料组件铀丰度实验研究

采用CZT探测器、数字谱仪、准直器等组成了1套便携式CZT探测器铀丰度测量装置。该装置可对燃料组件铀丰度进行测定,以便确定相应铀产品丰度符合规定要求。实验研究中,对几类燃料组件丰度进了测量,建立了CZT探测器测量燃料组件铀丰度的方法。现场测量结果表明,铀丰度测量结果相对偏差小于3%,方法简单可靠,装置简便,能满足核材料保障监督和核设施现场测量中的需求。     

Core Thermohydraulic Design with 20% LEU Fuel for Upgraded Research Reactor JRR-3

This paper presents the outline of the core thermohydraulic design and analysis of the research reactor JRR-3, which is to be upgraded to a 20 MWt pool-type, light water-cooled reactor with 20% low enriched uranium (LEU) plate-type fuel. For the condition of normal operation, the upgraded JRR-3 core is planned to be cooled by two cooling modes of forced-convection at high power and natural-convection at low power. The major feature of core thermohydraulics is that at the forced-convection cooling mode the core flow is a downflow, under which fuel plates are exposed to a severer condition than an upflow in cases of operational transients and accidents. The core thermohydraulic design was, therefore, done for the condition of normal operation so that fuel plates may have enough safety margins both against the onset of nucleate boiling (ONR) not to allow the nucleate boiling anywhere in the core and against the departure from nucleate boiling (DNB). The safety margins against ONB and DNB were evaluated. The core velocity thus designed is at the optimum condition where fuel plates have the maximum margin against the ONB, and the minimum DNB ratio (ratio of DNB heat flux to the maximum heat flux) was evaluated to be about 2.1, which gives a sufficient margin against the DNB. The core thermohydraulic characteristics were also clarified for the natural-convection cooling mode.     

The Operation of Nuclear Electric Irradiated Fuel Transport FlaskS

Abstract

The current flask designs are outlined and the Quality Assurance arrangements necessary to secure safe flask operation are described. Flask despatch procedures, transport logistics, maintenance requirements, emergency arrangements and the operational experience feedback system are described. The enviable safety record of over 14,000 loaded movements, shipping over 21,000 tonnes of spent fuel over 30 years without significant mishap and without any accident involving the release of radioactivity is contrasted with the public concern over the safety record of the nuclear industry.     

Optimization of Axial Enrichment and Gadolinia Distributions for BWR Fuel under Control Rod Programming

The axial enrichment and gadolinia distributions of BWR (boiling water reactor) fuel are optimized under control rod programming. The objective of the problem is to minimize the average enrichment required to reach a planned EOC (end-of-cycle) with criticality condition and axial power peaking constraint.

A method of approximation programming is employed as the basis for the solution method. Resulting linear programming problem at each iteration step is solved by means of goal programming algorithm. The method is applied to the initial fuel for a typical BWR/5 represented by an axial one-dimensional core model

Two-region analysis leads to the conclusion that the core bottom should be depleted during the cycle so that the power shifts to the core top at EOC. The enrichment and gadolinia distributions are determined to maximize EOC power peaking within a limit. The optimal solution of a 24-region fuel with a power peaking limit of 1.4 saves 10.6% in uranium ore compared with a uniform fuel depleted with a Haling power shape. Half the saving comes from an optimal natural uranium blanket implementation.     

252Cf中子活化核燃料棒235U富集度均匀性检测装置

采用^252Cf中子活化方法研制燃料棒^235U富集度均匀性检测设备，用慢化后的^252Cf中子照射燃料棒，使燃料棒UO2芯块中的^235U发生裂变，通过测量其裂变产物的γ射线总强度对燃料棒^235U富集度及其均匀性进行在线检测。采用1．2mg的^252Cf中子源，能检测出燃料棒中^235U富集度相对偏差土10％的单个混料芯块，单根燃料棒的检测速度可达7m/min，检测结果的置信概率为97．74％。     

循环工况转换过程中的燃料棒热应变分析

利用RELAP5程序建立分析模型,结合强迫循环工况转换自然循环工况的试验数据,对转换过程中的冷却剂温度变化速率和燃料棒的热应变进行计算分析.计算结果表明:结合RELAP5程序进行燃料棒的热应变分析是合适的;自然循环转换过程中燃料棒热应变的变化剧烈,对燃料棒的机械性能产生影响;结合RELAP5瞬态分析结果,也可以分析其他工况变化过程对燃料棒的热应力冲击问题.     

压水堆燃料组件安全运行保障措施分析

通过对燃料组件进行设计、制造、运行、辐照后检查多方面的分析,提出当前保障燃料安全运行体系改进的方向,特别提出加强燃料组件拆卸检查、池边检查、热室分析的建议。对于自主化燃料组件,从燃料棒下端部防止磨蚀、改进防异物设计、增强组件结构强度等方面提出了相关建议。     

New Fresh Mox Fuel Packages: From Design to Operation

Abstract

The MX packages developed by COGEMA Logistics according to TS-R-1 requirements will replace older Current packaging to transport fresh pressurised water reactor (PWR) and boiling water reactor (BWR) mixed oxide (MOX) fuels in Europe. Two types of package have been developed: (1) MX8 with a capacity of eight 17 × 17 900 MWe PWR fuel assemblies for dry loading and underwater unloading operations; and (2) MX6 for dry loading and unloading operations. The capacity of the MX6 is six 16 × 16 or 18 × 18 PWR fuel assemblies or sixteen 10 × 10 BWR fuel assemblies. To meet these capacities requirements, an innovative and optimised design has led to ‘mid-weight’ packages with original solutions for the body, the baskets and the fuel restraining system. To cope with both capacity and legal weight transport requirements, a new high-security transport system has been developed simultaneously. The first shipment with MX8 was made in December 2001, and the first use of MX6 packages is scheduled for the end of 2003.     

Effect of the VVR-SM neutron spectrum on the radioactivity and color of natural topazes

The neutron flux density from 0.025 eV to 12 MeV has been measured experimentally in all channels of the VVR-SM core by the activation method using threshold monitors (Au, Ni, Fe. Ti, Mg, Y). Comparing with a calculation of the neutron flux density at different energy using the IRT-2D computer code showed agreement to within 5%. The distribution of the neutron fluxes and spectra in the core, which is of practical utility for radiation technologies, was obtained. A series of irradiations has been conducted and experimental dependences of the irradiation time on the channel position in the core as well as on the size of the stones for obtaining a standard light blue and dark blue color have been obtained. The irradiation conditions making it possible to lower the induced radioactivity of the minerals three-fold as a result of increasing the ratio of the fast to thermal neutron fluxes are found.     

252Cf中子活化核燃料棒235U富集度检测设备的软件设计

介绍了252Cf中子活化核燃料棒235U富集度检测设备的软件设计,该软件采用多线程技术控制研华PCI-1780采集卡定时采集六路探测器输出的经252Cf中子活化后235U裂变产物的γ射线信号,针对采集数据的特性,进行相应的处理和分析,可以检测出核燃料棒的实际235U富集度以及有无异常芯块.该软件经过实验验证在检测速度为6时,能够准确测量核燃料棒的实际235U富集度值并判断棒中是否混有异常芯块,同时向PLC发送相应信号实现自动分选.目前已应用在核燃料元件厂的核燃料棒235U富集度无损检测设备上.     

锆基弥散微封装燃料在稳态运行条件下的失效机理研究

为实现锆基弥散微封装燃料（M3燃料）的优化设计，进一步提升其在轻水堆（LWR）运行环境下的可靠性，需对其在稳态运行条件下的失效机理进行研究。本研究借助于ABAQUS有限元软件，通过二次开发建立了M3燃料的辐照-热-力耦合性能三维数值模拟分析方法，并基于此分析方法对M3燃料在稳态运行条件下的失效机理进行了研究。研究结果表明，稳态运行期间M3燃料的失效主要以辐照初期内致密热解碳层（IPyC层）的失效、辐照中后期疏松热解碳层（Buffer层）与IPyC层分开再接触后导致的碳化硅层失效为主。该研究结果可为后续M3燃料的优化设计提供指导。      

18个月换料延伸运行一次侧超压保护验证

在一次侧平均温度和稳压器水位的整个变化范围内，应用CATIA2程序，对大亚湾18个月燃料循环延伸运行中的主蒸汽流量全部丧失事故进行了分析计算。结果表明：一次侧功率，平均温度和稳压器水位变化均能满足运行图的要求；事故中一次侧最大压力不会超过超压保护准则值。本文在运行图上给出了包络的运行区域，并从几十种计算工况中选出了两种具有代表性的工况进行计算分析。     

300#堆燃料组件运行历史数据库

为了配合ORIGEN2计算原始数据准备,采用Microsoft Access建立300#堆运行历史数据库.简化后只用一个表记录反应堆的运行历史.每盒燃料组件的表单只记录其经历的装载历史,最多不超过20条.表单之间用字段“装载ID“联接.对统计和录入中可能出现的两类错误,各建立一个查询,用于自动检索错误.对每盒燃料组件,建立查询,根据组件装载历史表单从总反应堆运行历史表单中采集数据,并将数据以文件文件形式输出,用编制的运行历史数据处理程序,将数据转换为0RIGEN2计算需要的运行历史输入数据.