共查询到19条相似文献,搜索用时 156 毫秒
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快堆堆芯抗震分析是堆芯设计的重要组成部分,它将为堆芯在地震作用下的结构完整性评价和堆芯反应性变化分析提供必要的数据,同时为控制棒的可插入性评价提供参考。本文采用日本有限元程序FINAS,以中国实验快堆为例,对快堆堆芯水平抗震的计算方法和模型进行了研究,完成了单组件预分析,其中包括模态分析、自由振动分析和与刚性墙壁的碰撞分析,为堆芯多组件水平抗震分析作好了准备。 相似文献
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对快堆堆芯组件进行的抗震分析需要考虑冷却剂与堆芯组件之间的流固耦合作用。在之前的分析中,大多数人将流体附加阻尼处理为定值。实际上冷却剂对组件的作用还随着组件间的间隙变化而变化,其带来的附加阻尼应为变量。为更准确地模拟堆芯组件的振动,本文采用变化附加阻尼对快堆堆芯组件的抗震分析方法进行了研究。建立了快堆堆芯单排(5根)堆芯组件的抗震分析计算模型,对该模型进行了附加阻尼为定值和随间隙变化两种情况下的抗震分析,结果显示了考虑变化附加阻尼的堆芯组件抗震分析方法的可行性与有效性。本文所使用的模拟方法更为贴近堆芯组件的振动情况,为更为真实地模拟快堆堆芯组件的地震响应打下基础,这也有助于减少结构设计的保守性,具有一定的工程价值。 相似文献
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针对熔盐快堆中子物理与水力强耦合的特点,使用开发的熔盐堆三维多物理耦合程序TMSR3D,分析了稳态情况下锕系元素再循环嬗变熔盐堆(MOSART)缓发中子先驱核守恒方程中湍流扩散项对熔盐快堆堆芯物理参数的影响。结果表明:在稳态情况下,湍流扩散项对堆芯有效增殖因数影响很小,对堆芯快中子和热中子通量密度的影响也很小,但湍流扩散项对堆芯缓发中子先驱核分布的影响大,且影响程度与具体的湍流运动黏度分布、湍流施密特数和不同的缓发中子先驱核群相关。 相似文献
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运用MCNP与ORIGEN2耦合计算程序COUPLE,对加速器驱动的次临界系统(ADS)钠冷金属燃料快堆堆芯进行稳态与燃耗计算,比较分析次锕系核素(MA)非均匀布置堆芯与均匀布置堆芯在MA嬗变效果与反应性参数方面的差异。计算结果表明,对比均匀布置,非均匀布置具有更高的MA嬗变率与嬗变支持比,在反应性参数方面导致多普勒效应与有效缓发中子分额降低,钠空泡效应增大,在堆芯功率分布与加速器束流功率方面没有明显变化。 相似文献
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次锕系核素(Minor Actinides,MA)作为长寿命高放射性核废料,在乏燃料放射性中占据主导位置。乏燃料最小化是保证核能可持续发展的重要环节,而嬗变是安全处置乏燃料的有效途径。小型模块化增殖焚烧(Breed and Burn,BB)快堆的中子经济性较好,燃烧寿期长,装料方式灵活多样,可用于增殖产生易裂变核燃料、嬗变长寿命核废料,从而解决核电发展中前端核燃料供给和后端乏燃料处理问题。本文分析对比了U3-MA和U5-MA燃料装载模式的临界、燃耗和安全性能,并系统研究了两种装料模式在BB快堆上嬗变MA的性能。结果表明:两种装料方式均能达到较好的嬗变性能,且MA的添加还能使反应堆寿期更长,堆芯中子经济性更高;此外,从安全性能上来看,添加MA对钍铀燃料循环的缓发中子份额影响较弱,但是对其燃料多普勒系数影响较强,这为后续钍铀、铀钚燃料循环选取合理的MA装载份额提供了参考依据。 相似文献
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热工水力分析软件的验证是安全审查重点关注的问题。为了实现不同设计软件间的对比验证,本工作开发出具有自主知识产权的钠冷快堆堆芯子通道分析程序SSCFR,进行中国实验快堆(CEFR)全堆芯稳态分析、子通道稳态分析及全堆芯瞬态分析,并将分析结果与CEFR运行和设计值进行对比。结果表明,SSCFR程序的计算结果与CEFR运行值及安全分析报告中的设计计算值符合较好,可用于钠冷快堆后续的软件对比验证及设计计算工作。 相似文献
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《Journal of Nuclear Science and Technology》2013,50(6):525-536
The 1,000kWe metal fueled sodium-cooled fast reactor concept “RAPID” to achieve highly automated reactor operation has been demonstrated. RAPID (Refueling by All Pins Integrated Design) is designed for a terrestrial power system which enables quick and simplified refueling. It is one of the successors of the RAPID-L, the operator-free fast reactor concept designed for lunar base power system. The essential feature of the RAPID concept is that the reactor core consists of an integrated fuel assembly instead of conventional fuel subassemblies. In this small-size reactor core, 14,000 fuel pins are integrated and encased in a fuel cartridge. Refueling is conducted by replacing a fuel cartridge. The reactor can be operated without refueling for up to 10 years. Unique challenges in reactivity control systems design have been addressed in the RAPID concept. The reactor has no control rod but involves the following innovative reactivity control systems: lithium expansion modules (LEM) for inherent reactivity feedback, lithium injection modules (LIM) for inherent ultimate shutdown, and lithium release modules (LRM) for automated reactor startup. All these systems adopt 6Li as a liquid poison instead of B4C rods. In combination with LEMs, LIMs and LRMs, RAPID can be operated without an operator. In this paper, the RAPID reactor concept and its transient characteristics are presented. 相似文献
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堆芯热通道因子是堆芯热工设计及安全分析的一项重要参数,确定热通道因子需用中子学计算给出较准确的燃料组件内元件棒功率分布。在三维六角形几何节块扩散理论基础上,使用多项式重构的方法计算节块内中子通量密度分布和功率密度分布。针对快堆六角形燃料组件的特点,用小六角形积分的方法计算组件内元件棒功率,得到组件内各元件棒功率分布。在NAS程序基础上,编制了元件棒功率分布计算模块NAS PIN。通过与蒙特卡罗程序的校验可发现,二者计算结果符合较好,计算精度可满足工程设计的需要。 相似文献
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V. Prakash M. Thirumalai M. Anandaraj P. Anup Kumar D. Ramdasu G.K. Pandey G. Padmakumar C. Anandbabu P. Kalyanasundaram 《Nuclear Engineering and Design》2011,241(8):3325-3332
The Prototype Fast Breeder Reactor (PFBR) which is under construction at Kalpakkam, India, is a 500 MWe sodium cooled pool type reactor. The core of the PFBR consists of 1758 free standing subassemblies supported on the grid plate. The entire core is divided into 15 different flow zones and the flow rate required through each zone is calculated based on the fission heat generation. The coolant sodium flows from the bottom of the subassembly to top and the design of the subassembly for each flow zone is quite complex. There are 181 fuel subassemblies in PFBR core with 217 fuel pins in each subassembly, vertically held in the form of bundle within a hexagonal wrapper tube. The pins are separated by spacer wires wound around the pins helically. Analytical prediction of subassembly pressure drop, vibration and determination of inception of cavitation for this complex geometry is very difficult. So experiments were conducted extensively to get a more accurate evaluation of the design and for its qualification for the use in PFBR, which is designed for 40 years of operation.Pressure drop and cavitation experiments were carried out in water on full scale (1:1) subassemblies of all flow zones. The overall pressure drop of the subassembly determines the ratings of the pump. Cavitation of the pressure drop devices lead to erosion damage of fuelpins and may also result in reactivity fluctuation due to sodium-void effect. So it is essential to confirm that the subassembly is not cavitating in the operating regime of the reactor. Subassembly can vibrate in cantilever mode due to the turbulence in the flow and can result in reactivity fluctuation, reactor control problem and can even lead to the failure of the fuel pins. So vibration measurements were carried out in water on the maximum rated subassembly. This paper discusses various experiments carried out on PFBR subassembly, the similarity criteria followed, instrumentation, results and conclusion. 相似文献
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《Journal of Nuclear Science and Technology》2013,50(12):1066-1069
In most of the calculations using analytical methods a reactor core is approximated as cylinder and the reactor parameters are calculated using two-dimensional computer codes. While such calculations are useful in scoping studies in view of azimuthal asymmetry in the actual reactor core these calculations could entail errors of unknown magnitude. The present study reports our estimate of such errors in K eff with the instance of fast reactor having 22 and 23 fuel subassemblies. The K eff are calculated using Monte Carlo code KENO and Hansen-Roach cross section set, modelling the core in two different ways, (1) by approximating the core to a cylinder (2-D calculation), (2) by near exact representation of the core (3-D calculation). The difference in K eff is appreciable between 2-D and 3-D calculations. Experimental values are adduced in support of these calculations. 相似文献
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《Journal of Nuclear Science and Technology》2013,50(4):693-700
Core modification has been investigated to further increase the core burnup and to improve the irradiation efficiency of the experimental fast reactor Joyo. This modification enables the core to accommodate more irradiation test subassemblies that have lower fissile material contents compared with the driver fuel. The design calculations showed that the replacement of the radial reflector elements made of stainless steel with those made of zirconium alloy or nickel-based alloy is effective in improving neutron efficiency. The irradiation test capacity can be increased by reducing the number of control rods based on a reevaluation of the design margin in the control rod worth calculation. The design calculation results show that these modifications, without any change in fuel specification, will be useful for conserving driver fuels and enhancing the irradiation capability of Joyo. 相似文献
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与临界反应堆相比,ADS次临界反应堆的外源中子和裂变中子的空间分布具有严重的不均匀性,对应的中子价值也不同。本工作对次临界反应堆的稳态输运方程作分群扩散近似,得到了多群方程,进一步推导出按堆芯功率归一化的中子共轭方程表达式和与功率相关的中子价值函数表达式,给出了次临界反应堆中子价值的物理意义。由稳态中子共轭方程组出发,给出了两种带外加中子源的次临界反应堆增殖因数的表达式。 相似文献
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增殖燃烧一体化快堆插花式倒料方案研究 总被引:1,自引:1,他引:0
增殖燃烧一体化快堆利用快堆的增殖特性,通过倒料完成从增殖组件向燃烧组件的过渡,从而实现增殖和燃烧过程的一体化。全寿期内燃烧组件提供堆芯的绝大部分功率,而在燃烧组件周围的贫铀组件则将其中的238U转化为239Pu,实现增殖功能。通过定期倒料,堆芯在一次装料后可实现长期自持临界,维持几十年的稳定运行。合理的堆芯布置与倒料方案可更好地平衡燃料的燃烧和增殖过程。插花式的堆芯布置与倒料方案是将一部分增殖组件分散布置在堆芯高通量区,保证了增殖组件的快速增殖,同时可保持堆芯在整个反应堆寿期内具有稳定的功率分布。另外,插花式堆芯布置与倒料方案最终的组件卸料燃耗是相对均衡的,所有从燃烧区倒出的组件均具有相近的燃耗,一般在250~300 GW•d/t左右。这使得增殖燃烧一体化快堆可在不进行燃料后处理的条件下,实现铀资源的高效利用。 相似文献
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目前商用压水堆积累了大量的长寿命高放废物,放射毒性强,衰变时间漫长,对环境和人类构成了长期威胁,作为6种第四代核能系统堆型中的一种,铅基冷却快堆在减少长寿命高放废物产生方面具有优势。基于此本文提出了一种热功率为300 MW的铅-铋合金冷却快堆设计。利用MCNP程序对反应堆堆芯进行建模并计算了堆芯在寿期初的主要物理参数,详细分析了燃耗过程中长寿命高放核素的积累量,并与一般压水堆长寿命高放核素的积累量进行了比较。结果表明,对主要关心的次锕系核素,铅-铋合金冷却快堆的产生量远小于压水堆的,而长寿命裂变产物的产生量与压水堆的相当。总体来说,铅-铋合金冷却快堆产生的长寿命高放废物总量小于压水堆的,可看出铅-铋合金冷却快堆在减少长寿命高放废物产生方面更具有竞争性。 相似文献