研究堆新燃料组件铁路运输货包固定系统安全分析

研究堆新燃料组件放置在专用货包内采用铁路运输,货包用木板、螺栓等进行紧固。对紧固系统在正常情况和极端情况下的受力情况进行了分析。强度校核计算表明,货包的固定方式能够保证其即使在10g的加速度下也不会在各个方向移动。稳定性校核计算表明,固定系统结构在此情况下也是稳定性的。实际运输监测中发现,全程最大加速度未超过10g,固定系统能够保障燃料组件运输容器的铁路运输安全。     

CNSC乏燃料组件运输容器临界安全分析

临界安全作为乏燃料组件运输容器的一项重要安全指标,需经过计算和分析以判断其是否满足法规标准。为分析中国核工业集团有限公司(China National Nuclear Corporation,CNSC)乏燃料组件运输容器临界安全设计是否满足《放射性物品安全运输规程》的要求,使用蒙特卡罗程序MCNP(Monte Carlo N Particle Transport Code)构建了保守临界计算模型,对正常和事故工况下CNSC乏燃料组件运输容器进行了临界计算分析。分析表明:正常运输条件下单个货包和货包阵列的k_(eff)最大值为0.804 25,小于次临界限值,临界安全指数为0;事故工况下单个货包和货包阵列的k_(eff)最大值为0.813 17,小于次临界限值,临界安全指数为0。可见,正常和事故工况下,CNSC乏燃料组件运输容器的keff最大值均小于0.94的次临界限值,临界安全指数为0,满足法规标准要求。     

AP1000新燃料组件运输货包的临界安全计算

新燃料组件运输过程中最主要的核安全问题是临界安全。在对运输货包进行临界安全分析中必须要同时考虑多货包阵列形式、事故后货包损伤情况、最佳水慢化条件等因素。本文采用MCNP程序针对美国西屋公司XL型运输容器装载AP1000新燃料组件货包的实例进行了临界安全计算。结果表明,在XL型运输容器设计许可书中允许装载货包数N=75的限制条件下,临界安全是有保障的。     

高温气冷堆新燃料运输货包的撞击事故分析

介绍了高温气冷堆新燃料运输货包严重撞击事故的仿真计算分析方法。根据实际货包结构及运输条件,确定了分析的严重撞击事故景象。通过有限元法计算分析了货包在不同姿态、不同速度下的碰撞结果,给出了容器不同部分及所装载的燃料组件的损坏情况。在此基础上,计算了严重事故景象下有效增殖因子k_eff。     

中国先进研究堆新燃料组件运输的临界安全分析

在新燃料组件运输过程中,临界安全是重点。使用MCNP程序对中国先进研究堆新燃料组件的运输进行临界安全计算分析,通过选取最不利临界安全的次临界限值、组件模型参数、事故工况来保证计算结果的保守性。结果表明,运输货包的临界安全指数可确定为0。该结果可为中国先进研究堆（CARR）的新燃料组件运输容器的研发提供参考依据。     

CEFR-MOX新燃料组件运输货包临界安全计算

易裂变材料运输过程中重要的安全问题之一是临界安全。在对运输货包进行临界安全分析中必须要同时考虑多货包阵列形式、事故后货包损伤对临界安全影响、最佳水慢化条件等因素。本文采用MCNP 程序针对CEFR-MOX新燃料组件运输货包进行了临界安全计算。计算结果表明:MCNP程序(采用核截面库为ENDF/B-V库)对本问题的次临界限值为0.924 6;正常运输条件下无限个运输货包的最大k_eff值为0.574 4,运输事故条件下无限个运输货包的最大k_eff值为0.659 7。根据临界安全指数的定义,确定CEFR-MOX新燃料组件运输货包的临界安全指数为0。     

放射源运输容器安全设计有限元分析方法

放射性物质运输容器力学试验是证明货包安全设计满足法规标准要求的重要工作之一。根据法规标准要求,应采用能够导致货包产生最严重损坏的姿态进行力学试验,评价力学试验后容器的安全性能。通过有限元分析来确定容器最严重损坏的姿态是目前国际上通常采用的方法,能够极大地节约时间和成本。本工作针对某型号放射源运输容器,通过分析容器力学有限元计算结果,确定容器最严重损坏的姿态,分析比较有限元计算结果和试验结果,证明放射源运输容器安全设计满足法规标准要求。     

重水运输容器货包自由下落分析

本文采用ANSYS有限元程序,对重水运输容器货包进行了自由下落分析,计算模型包括3种下落方式:水平下落、垂直下落和倾斜下落.根据ASME规范NB分卷进行了应力强度评定.结果表明,重水运输容器满足强度与密封要求.     

CEFR新燃料组件贮存临界安全分析

为论证新燃料组件贮存是否满足核临界安全要求,对CEFR新燃料贮存系统进行核临界安全分析计算。 新燃料组件贮存系统由1、2号格架组成,每个格架可贮存56个(7层×8个)组件容器,一个组件容器里装2盒组件,即每个格架可贮存包括控制组件、燃料组件等在内的112盒组件。新燃料组件首先     

AP1000型燃料组件水平放置对格架弹簧影响分析

核电厂建设的工期延误和组件制造厂的燃料组件存贮场地不足,不能按期交付的首炉燃料组件被迫存放在组件运输容器内水平放置,有的燃料组件在运输容器内水平放置时间约1年以上.针对燃料组件在运输容器内长期水平放置是否对组件格架弹簧的力学特性有影响,以AP1000型燃料组件为例,对格架弹簧进行了力学特性影响分析.分析认为,燃料组件在...     

Comprehensive evaluation of the state of spent VVR fuel assemblies from first-generation nuclear submarines

New, unconventional methods developed at the Russian Science Center Kurchatov Institute for evaluating the state of spent VVR fuel assemblies from submarines are described. They were used in combination with other methods during inspections of spent fuel assemblies held in TK-6 and -11 containers at the point of storage of spent nuclear fuel and radioactive wastes in the town of Gremikha in Murmansk Oblast.     

乏燃料组件燃料棒更换装置抓爪结构的数值分析和试验研究

燃料组件是核电站核反应堆的关键设备之一,涉及燃料组件的维修特别是乏燃料组件破损棒更换维修属于高风险作业。本文主要针对乏燃料组件燃料棒更换装置的核心零件燃料棒抓爪的结构进行研究,通过结构力学分析得到抓爪较优壁厚数值,然后通过有限元计算抓爪的强度固化结构参数,最终进行抓爪试制,并通过抓爪试验台模拟抓爪的实际工况对抓爪进行性能测定,确保抓爪满足使用要求。     

快堆燃料组件抗震分析流体附加质量计算方法研究

浸没在液态钠中的快堆堆芯组件在地震作用下发生振动,可能导致组件结构损坏或堆芯结构变形,从而影响反应堆结构完整性和安全.流体使该振动表现为强烈的非线性,因此,研究地震引起的流固耦合效应对快堆抗震分析十分重要.本文主要研究流固耦合问题中附加质量的计算方法,该方法由Westergaard首先提出,是一种考虑水体对结构作用的简化动力学计算方法,它将动水压力等效成质量附加在结构上,质量等效原则自提出在各行业得到广泛应用,但缺乏详细理论推导.本文首先推导出附加质量公式,并对该公式进行有效性分析;接着对单根和两根组件用CASTEM在空气和水中进行建模;最后将频率、碰撞力分别与试验值比较.结果表明,计算值和试验值吻合.     

Handling spent nuclear fuel and radioactive wastes on the shore servicing base in Gremikha

The situation which has developed at the shore base in Gremikha involving fuel assemblies in the removed cores of water-moderated water-cooled reactors (first-generation submarines), which are located on an open site in containers and receiving chambers, and involving solid and liquid radioactive wastes present at the base is examined. Data are presented on the number of fuel assemblies and their technical state and on the state and amount of solid and liquid radioactive wastes. Suggestions on what should be done with the fuel assemblies and wastes are discussed. __________ Translated from Atomnaya énergiya, Vol. 101, No. 1, pp. 61–65, July, 2006.     

An economically optimum PWR reload core for a 36-month cycle

Cycle length extension in currently operating PWRs may be economically interesting if the benefits stemming from capacity factor improvement offset the higher fuel costs of the longer cycle. A PWR reload core is presented that meets current physics and fuel performance design limits for a cycle of 33.9 EFPM or 36 calendar months when operating at a capacity factor of 94.1%. Fuel is enriched to 6.5% U-235 and selected pins use gadolinia as burnable absorber mixed with UO₂. The power is evenly distributed over a broad region of the core by including pins with two different concentrations of gadolinia in the assemblies. The core periphery is loaded with reused assemblies. The rest of the assemblies are discharged after one cycle in the core. The fuel performance is acceptable, although the parameters analyzed are closer to the limits than in a contemporary reference 18-month cycle multibatch loading strategy. The 36-month core is economically competitive with an 18-month reference core under certain operational conditions. Potential reductions in fuel enrichment costs would make the 36-month cycle cost competitive with the 18-month reference cycle under a wide range of conditions.     

Application of Fire-Retardant Treatment to the Wood in Type A Unirradiated Nuclear Fuel Outer Containers

Abstract

Packagings for transporting unirradiated nuclear fuel assemblies in the United States are commonly constructed as rectangular boxes consisting of a metal inner container, a wooden outer container, and cushioning material separating the two. The wood in the outer container is a potential source of fuel for fire. Use of a fireretardant treatment on the wood may reduce or eliminate the damage to nuclear fuel assemblies in some types of accidents involving fire. The applicability of using fire-retardant treatments on the wood of outer containers is addressed. An approximate cost-benefit analysis to determine if fire-retardant treatments are economically justified is presented.     

Certification test for Safety of New Fuel Transport Package (Fire Test)

Abstract

Fire tests of actual transport systems for both BWR type and PWR type new fuel assemblies were carried out to clarify their safety margin, create a public perception of the transport programme and offer a data base to verify analytical codes. The supposed scenario of traffic accidents was considered by first reviewing them to select the plan of this work. Next, heating tests of simulated fuel rod elements and fuel assemblies of 1 m length were conducted in an electric furnace to examine their deformation and oxidisation behaviour under high temperature conditions. An analytical code including a simple model was verified by comparison with data obtained in a furnace test simulating a fire accident. Furthermore, simulated fuel assemblies and actual containers of both BWR and PWR types were manufactured. Actual 11 ton trucks carrying them were put in a burn pit 5 m in width, 11·5 m in length and 0·5 m in depth. The fire tests were carried out by burning kerosene for 30 min. These experimental results were analysed by the code.     

Design and thermal modeling of radially configured nuclear waste packages

A series of radial design configurations for packaging nuclear wastes are described. These radial arrangements for used nuclear fuel assemblies in containers are effective techniques for packaging significantly more radioactive waste in the available internal container volume. The radial package designs can be applied to packaging the nuclear waste for permanent storage at the Yucca Mountain (YM) repository. The radially configured containers will have high degree of structural strength and will be efficient in transferring heat from the waste form to the package surface due to the minimization of internal gaps. Radial configurations are reported for packaging the Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) used fuel assemblies. These configurations can be varied for co-packaging the colder, i.e. vitrified high level waste (HLW), canisters. Details of the geometry and the materials selected are discussed. Thermal analysis of the radial designs was conducted which confirm the feasibility of the designs demonstrating that no over-heating occurs in the contained nuclear waste in spite of the significantly extra amount of waste. The larger amount of packaged waste per container coupled with efficient heat transfer characteristics of these designs favor hotter and drier conditions for container surfaces in the YM emplacement drifts.     

Handling spent nuclear fuel from the research reactors of the Russian Science Center Kurchatov Institute

Shipping out the spent fuel of the research reactors at the Institute for reprocessing is examined. The spent fuel is characterized by a great diversity of structural characteristics of the fuel assemblies and fuel elements, fuel compositions, and the enrichment, burnup, and cool-down times of the fuel as well as the state of the components of the assemblies and the structural materials. A classification and quantitative indicators of the accumulated spent fuel from the standpoint of the modern state of its reprocessing technology and the requirements for delivery to the Mayak Industrial Association are presented. The structural features of the TKU-19 and -128 shipment containers are presented, and the loading of spent fuel assemblies into them for shipment to reprocessing is described. The plans and goals of further work on the removal of spent fuel from the Institute’s territory are presented. Translated from Atomnaya énergiya, Vol. 106, No. 2, pp. 99–105, February, 2009.