秦山核电站二期反应堆堆芯流量分配数值分析

获得可靠的堆芯入口流量分配数据是改善压水堆堆芯热工水力性能的需要。应用计算流体力学方法研究了反应堆压力壳内复杂的流动现象,得到了秦山核电站二期600 MW反应堆1/4整体水力模型的堆芯入口流量分配状况,并对下腔室几何结构、冷管段入口流量等影响因素进行了敏感性研究。分析结果表明,双环路工况入口流量对堆芯入口流量分配影响较小;与双环路工况相比,单环路运行工况时流动特性显著变化,导致流量分配状况差异较大;堆芯入口流量再分配因子为0.05,与原型设计参数吻合。计算结果证实所采用研究方法有效,可为相关反应堆工程设计验证提供依据。     

373 MW液态燃料熔盐堆流量分配设计的优化研究

流量分配设计是熔盐堆热工水力学研究的重点内容之一,流量分布直接决定了堆芯局部热点值和位置,制约着反应堆运行的安全性可靠性。以热功率373 MW钍基熔盐堆液态燃料(Thorium Based Molten Salt Reactor-Liquid Fuel,TMSR-LF)为研究对象,使用计算流体力学通用程序Fluent 15.0对堆芯流场进行了数值模拟。基于相应热工水力参数的分析结果,对原堆上下腔室的结构提出了若干分步式改进方案:下腔室设置折流板、增加上腔室高度、调整下腔室流量支撑板的孔径。优化结果表明:下腔室设置圆柱折流板能有效解决由局部涡流遮蔽部分通道引起的质量流量不均匀的问题;增加上腔室高度可以平衡径向方向内外侧通道出入口的压降,改善流量的均匀性;减小流量较高的内侧通道对应的金属支撑板孔道的半径可有效降低该处流量,使总体分布趋势更加平缓。最终确定的优化方案为在下腔室中央位置设置半径240 cm圆柱型折流板,半椭球型上腔室高度调整至45 cm,下腔室流量支撑板2～3排孔径由2.3 cm减至2.1 cm,4～5排由2.3 cm减至2.2 cm。以上结果对TMSR-LF2堆芯上下腔室的工程优化设计具有重要的参考价值。     

一种液态燃料熔盐堆堆芯流量分配设计

在液态燃料熔盐堆(Molten salt reactor,MSR)热工水力设计中,为实现堆芯径向功率展平需对堆芯流量分配进行设计,使得堆芯进口流量分布正比于释热量分布,而下腔室结构和流场分布对堆芯流量分配起决定性作用。利用FLUENT软件对堆芯三维流场进行模拟,通过调节下腔室结构和流量分配装置,对下腔室流场分布进行优化,最终实现堆芯流量合理分配。数值模拟结果表明,喇叭状下腔室比椭球形下腔室熔盐通道流量标准差降低4.2%,设置流量分配板熔盐通道流量标准差降低29.2%;改变下腔室结构和设置流量分配装置能够较好调节流量分配和功率分布匹配性,该结果可为液态熔盐堆堆芯优化设计提供依据。     

VVER型反应堆堆芯流量分配与组件冷却剂温升CFD分析

利用计算流体力学程序CFX对VVER-1000型反应堆三维流场进行模拟,并计算了热组件冷却剂的温升。结果表明,燃料组件流量分配系数最大值为1.12,最小值为0.92;热组件流量分配系数约为0.97;偏离工况条件下,热组件冷却剂温升均高于当前温升预警限值ΔTt。该分析结果可为核电站运行中ΔTt的设定提供参考。     

CAP1400反应堆堆芯入口流量分配试验研究

反应堆堆芯入口流量分配是反应堆水力性能研究的重要内容之一,其与堆芯热裕量和燃料组件燃料棒的流致振动密切相关,从而影响反应堆的运行。CAP1400反应堆堆芯入口流量分配试验是验证CAP1400反应堆结构设计与分析的一个重要环节,旨在验证CAP1400反应堆堆芯入口流量分配的均匀程度。本文通过1/6比例模型试验,获得无均流板结构工况和带均流板结构3种工况（均匀流量工况、非均匀流量工况、偏回路流量工况）下CAP1400反应堆堆芯入口流量分配结果,并进行了各工况下流量分配均匀程度的分析。试验结果表明,CAP1400反应堆堆芯入口具有较好的流量分配效果。     

中国先进研究堆全堆芯流致振动及流量分配试验研究

由于设计和安全评价的需要,对中国先进研究堆(CARR)进行了1∶1全堆芯的流致振动和流量分配试验研究.文章分别介绍了流致振动和流量分配试验研究的技术路线、模型的设计、试验研究的内容、试验方法、试验结果分析和得到的结论等.试验中发现了结构设计的部分问题,设计方根据试验结果改进和优化了最终设计.试验验证了CARR堆内部件和堆芯的设计,为CARR工程通过安全审评提供了依据.     

板状燃料组件流量分配CFD研究与优化

板状燃料组件被广泛应用于研究堆中,组件内的流量分配是设计时需要考虑的一项重要内容。计算流体动力学(Computational Fluid Dynamic,CFD)方法是研究流量分配的重要手段,但有限的计算资源限制了其在板状燃料组件流量分配研究中的推广。针对板状燃料组件冷却剂流道狭长、封闭的特点,提出了部分建模迭代求解的计算方式,将无流量分配组件与有流量分配组件两种工况下各流道流量的计算值与直接完整建模的结果进行了对比,最大误差分别为0.56%与0.81%。鉴于前者对计算资源的需求远小于后者,部分建模迭代求解可以作为板状燃料组件流量分配CFD研究的合理可信的优化方案。     

CSR1000堆芯流量分配的研究

以中国百万千瓦级超临界水冷堆(CSR1000)堆芯为研究对象,建立热工水力计算模型,计算出冷却剂和慢化剂温度分布、堆芯功率分布、燃料组件出口压力及流量分配等参数。计算结果表明,适当增加堆芯内部燃料组件流量比例,可以有利于径向功率展平,内外燃料组件通道出口压降,呈现"N"型变化,增大内部燃料组件的堆芯入口功率,内部组件内的流量分配也将减少,而外部燃料组件通道中的流量将增加,适当调整堆芯入口流量初始分配比例,可以使各通道功率分布展平。     

中国实验快堆全堆芯流量分配计算与试验

针对中国实验快堆(CEFR)堆芯和一回路的设计特点,开发水力特性计算程序DAEMON,完成不同工况下的全堆芯流量分配计算,给出流量分配不均匀性等参数。在反应堆调试阶段,进行全堆芯流量分配试验。结果表明,程序计算值与试验值符合较好。在此基础上,验证了CEFR堆芯的流体力学设计,并为反应堆调试和运行提供了基础数据。     

Multipurpose electron accelerator driven electronuclear system based on a subcritical cascade reactor

An electronuclear system consisting of an electron accelerator, a neutron-producing target, and dual-zone subcritical blanket with fast and thermal neutron spectra is proposed. Some general mechanisms of cascade multiplication of neutrons in a dual-zone subcritical blanket are examined. The results of calculations of the electron-photon-neutron interactions of an electron beam with the target material and of the neutron-physical and heat-engineering characteristics of the system are presented. It is shown that a cascade neutron amplification factor of 2.5–3 can be obtained with system subcriticality 2%. The power of the system reaches 50 MW with electron beam power 4 MW. __________ Translated from Atomnaya énergiya, Vol. 102, No. 2, pp. 92–98, February, 2007.     

ADS原理验证装置快热耦合次临界堆设计

作为加速器驱动洁净核能系统(ADS)原理验证装置"启明星"一号的次临界驱动堆,堆芯采用快-热耦合方式组成,由天然金属铀组成快中子能谱区能有效地嬗变锕系元素(MA),低浓铀元件组成热中子能谱区能有效地嬗变裂变产物(FP).使用MCNP程序对次临界实验装置进行设计计算,确保keff在0.90～1.00之间.     

用于ADS次临界反应堆Keff测量微电流放大器的设计

设计了一种基于噪声分析方法,利用反应堆堆芯中子涨落特性测量加速器驱动洁净核能系统（Accelerator driven system,ADS）次临界反应堆动态参数的微电流放大器,并且根据需要设计了微电流源,运用Protel实时电路仿真工具对放大器进行仿真调试,使其达到设计要求。     

Simulation and optimization for a 30-MeV electron accelerator driven neutron source

A neutron source driven by electron accelerator is proposed in Shanghai Institute of Applied Physics (SINAP). The facility is planned for the study of nuclear data in Thorium-Uranium cycling system, and for material research. A detailed simulation of the neutron source is performed for the program to get the neutron generation maximum economically. Several parameters of the facility, which affect the neutron yield and the neutron escape from outer surface of the target, are analyzed respectively. Besides, the yielding neutron spectrum and the escaping neutron angular distribution are calculated and discussed.     

Testing a system prototype for out-of-reactor monitoring of reactor power and its distribution in the core volume

Conclusion Prototype tests were performed for a representative set of reactor conditions: the power distribution was performed by a wide range of changes in the position of three groups of control and protection system regulatory devices, the A0 values were varied in the range from –0.40 to 0.16, and the coefficient of non-uniformity k_v in the range from 1.7 to 2.6. The results showed the high accuracy and effectiveness of out-of-reactor monitoring of the power and its distribution throughout the core volume.For an LWR, the out-of-reactor monitoring system assemblies can be placed in radiation shielding channels; for this three assemblies, each having three detectors, is sufficient. Chambers having an energy range of standard neutron flux monitoring equipment can be used as detectors.Determining the thermal power and the coefficients of nonuniformity of its distribution in the core does not begin to exhaust the possibilities for out-of-reactor monitoring. Algorithms already exist or are being developed which would allow increased accuracy in the monitoring of power and its distribution, localization of the region or fuel element with the greatest energy loading, detection of stuck control rods and nonfunctional thermal monitoring sensors, and diagnosis of fluctuations and position shifts of internal reactor vessel structures. Thus a reliable, cheap, rapdily responding core condition diagnostics systems can be constructed on the base of out-of-reactor detectors.Translated from Atomnaya Énergiya, Vol. 64, No. 3, pp. 174–180, March, 1988.     

Transient analyses for accelerator driven system PDS-XADS using the extended SIMMER-III code

Transient analyses for Preliminary Design Studies of an Experimental Accelerator Driven System (PDS-XADS) were performed with the reactor safety analysis code SIMMER-III, which was originally developed for the safety assessment of sodium-cooled fast reactors and recently extended by the authors so as to describe the XADS specifics such as subcritical core, strong external neutron source and lead–bismuth–eutectic (LBE) coolant. As transient scenarios, the following cases were analyzed in accordance with the PDS-XADS program: spurious beam trip (BT), unprotected beam overpower (UBOP), unprotected transient overpower (UTOP), unprotected loss of flow (ULOF) and unprotected blockage (UBL) in a single fuel assembly. In addition, to cover some core-melt situations and investigate the potential for recriticalities, so-called snap-shot analyses with ad hoc postulated severe blockage conditions were also investigated.The simulation results for BT and UBOP showed that immediate fuel damage might not take place under short-time beam interruption or a 100% increase of the external neutron source. Concerning UTOP, it was found that a reactivity jump of 1 $ would not lead to damage of the fuel and the cladding. The ULOF simulation showed that the remaining natural convection of the coolant would prevent the cladding from disruptions. In the simulation of UBL in a single fuel assembly, it was shown that no cladding failure might be expected, due to the radial heat transfer and the coolant flow in the hexcan gap. Under an artificial suppression of the radial heat transfer for this UBL case, a pin failure occurred in the simulation but subsequent fuel sweep-out into the upper plenum region would bring a reactivity reduction and no power excursion. The severe accident simulations starting from postulated blockage above an already disrupted core showed that a severe recriticality could be avoided by the fuel sweep-out into the dummy-assembly or hexcan gap regions.The present simulation results showed that the current PDS-XADS design has a remarkable resistance against severe transient scenarios even in core-degradation conditions.     

The study of aeroball system for measuring 3D neutron flux distribution in reactor core

Aeroball system is attractive in several aspects because it can easily transport the map of neutron flux distribution to be measured from incore to outside of a reactor vessel.However,before the aeroball system is put to practical use in the heating reactor.there are four topics that have to be further studied.They are the stability of the activated positions,enhancement of signal/noise(S/N)ratio,distributed control and data-acquisition system and on-lin nbeutron flux distribution reconstruction.Besides describing the rasons for them,this paper gives out the theory,concept and solution about the first two topics and it is helptul to give the possibility to enhance the reactor-power.     

QACO_Alpha applied to the nuclear reactor core fuel reload optimization

The nuclear reload optimization is an important issue to nuclear engineering. It consists on maximizing the length of the operation cycle of the power plant. The aim is to find a configuration of the fresh fuel assemblies and remnants in order to keep the power plant running at full power by the largest time as possible.Quantum-inspired evolutionary algorithms are optimization tools based in artificial intelligence developed to simulate the quantum processing in classical computers. In this work is introduced one of these tools, which adds quantum concepts to the biological metaphor of collective learning of the real ants, named QACO_Alpha. It uses mechanisms developed in order to avoid premature convergence problems like a pheromone evaporation step besides a new updating method.To show its effectiveness, QACO_Alpha was applied to the optimization of 7th cycle of Angra 1. Experimental results were confronted to that obtained with other optimization methods, qualifying QACO_Alpha as a valid optimization tool for this kind of problem.     

Efficient code system of core characteristics for small cylindrical thermal reactor

A new code system for the overall neutronic calculation of a thermal reactor by a simple and effective way is presented. The code covers microscopic library compilation, macroscopic constant generation, cell calculations by multi-group treatment for neutron transport equation and core calculations over three zones for fuel and one zone for moderator. The Dancoff correction factor required in the interpolation of the self-shielding factors of resonance nuclides is automatically calculated by the installed collision probability routines. The burn-up calculation and Garrison and Ross model of fission product have been included. Also the effect of control rod on the reactivity of the reactor with special treatment for the control rod based on the homogenization technique has been included. Making a comparison with SRAC95 code system has checked the adopted code.