4 CARR电气系统施工设计

CARR供电系统由中压、低压、备用、直流、应急及接地等子系统构成。中压系统由两段母线、20面开关柜及6面电容补偿柜组成，其主要负荷是4台主循环泵、5台二次水泵和4台10／0．41000kVA的主变；低压系统由四段母线及33面开关柜组成，其电源来自4台主变的低压侧，其主要功能是向所有的低压负荷提供电力并作为应急和备用供电系统的悠闲电源；直流系统由三面开关柜组成，主要向中压系统提供操作电源。     

压水堆核电站主循环泵的试验研究

压水堆动力装置的主循环泵是使一回路冷却剂循环流动的设备,是核电站中重要的转动机械,必须具有良好的工作性能和安全可靠性。本公司应用美国西屋公司提供的技术和自己的试验研究成果,于1979年在日本首次制成了93A 型主循环泵.两台主循环泵已用于九州电力公司玄海核电站2号机组.     

CARR电气系统施工设计

CARR供电系统由中压、低压、备用、直流、应急及接地等子系统构成。中压系统由两段母线、20面开关柜及6面电容补偿柜组成,其主要负荷是4台主循环泵、5台二次水泵和4台10/0.41000kVA的主变;低压系统由四段母线及33面开关柜组成,其电源来自4台主变的低压侧,其主要功能是向所有的低压负荷提供电力并作为应急和备用供电系统的悠闲电源;直流系统由三面开关柜组成,主要向中压系统提供操作电源;备用供电系统由2台互为冗余的柴油机和配电装置组成,其主要作用是向允许短时中断的负荷提供电力。应急供电系统由3套独立的UPS、3组蓄电池组及开关柜组…     

CARR堆芯燃料管理技术研究

本课题选用WIMSD一4和CITATION两个通用物理计算程序,建立针对CARR堆芯的物理计算模型,研发了CARR专用燃耗计算程序和CARR堆芯换料专用程序,完成程序计算验证和改进。     

主循环泵重量和体积的优化设计

主循环泵是核动力装置中的重要设备,其重量、尺寸是影响核动力装置整体重量、体积及布置的重要因素。本文采用自主开发的复合形-遗传优化算法,以主循环泵的重量和体积作为优化目标,对主循环泵进行了优化设计。结果显示：优化方案与母型相比重量减小了10.82%,体积减小了4.28%,优化效果显著。同时对主循环泵重量和体积受一回路运行参数以及泵转速影响的敏感性进行了分析,其结果可供工程设计参考。     

CARR控制棒物理设计计算

中国先进研究堆(CARR)是一座高性能、多用途新型研究堆。CARR的控制棒物理计算是一个难点,本文采用Monte Carlo方法计算CARR的控制棒物理特性。物理计算的结果直接指导控制棒的结构设计和加工。     

中国先进研究堆稳态自然循环能力分析

针对中国先进研究堆(CARR)的结构和运行特点,开发了CARR自然循环能力计算程序,计算得到了不同池水温度条件下CARR自然循环能力,并分析了池水温度对CARR自然循环特性的影响:自然循环冷却剂流量随池水温度的升高而增大,但自然循环能力(带走的堆芯功率)随池水温度升高而降低.基于理论推导和程序计算结果,提出了一个适用于预测不同池水温度下CARR自然循环流量和堆芯功率的简单关系式,该关系式预测值与程序计算结果误差小于±10%.     

CARR热工水力与安全分析程序TSACC的开发与验证

针对中国先进研究堆(CARR)的具体结构和运行特点,利用Fortran程序设计语言开发了CARR热工水力安全分析程序TSACC(Thermal-hydraulic and Safety Analysis Code for CARR). TSACC完全采用模块化结构设计,便于二次开发,可应用于多种事故工况及其他堆型的分析计算.基于程序验证的基本思想,分别利用TSACC和商用程序RELAP5/Mod3对CARR丧失厂外电源事故工况进行了计算.得到了堆芯平均通道以及最热通道内冷却剂流量、温度和最小偏离泡核沸腾比(MDNBR)等参数的瞬态响应.将TSACC计算结果与RELAP5/Mod3计算结果进行比较、分析后发现:除冷却剂发生倒流前后二者计算结果相差较大外,总体吻合较好.局部值差别较大的主要原因是两个程序在低流速区域选用的换热公式不同.程序验证结果表明了TSACC的准确性和适用性.     

白俄罗斯首台核电机组完成热试

正【俄罗斯国家原子能集团公司网站2020年4月15日报道】俄罗斯国家原子能集团公司(Rosatom) 2020年4月15日宣布,白俄罗斯首台核电机组即奥斯特罗韦茨1号机组已完成热试。奥斯特罗韦茨正在建设两台VVER-1200机组,计划分别于2020年和2021年投入运行。热试主要包括以下流程:在主回路压力为160 kg/cm2且温度超过280℃的情况下,测试所有四台主循环泵的性能;冲洗主蒸汽管道;检查反应堆控制和保护系统及辅助电源系统;对蒸     

屏蔽电机有效材料重量的优化设计

屏蔽电机作为屏蔽式主循环泵中的主要大型部件之一，其重量、尺寸直接影响主循环泵重量和体积。本文基于自主开发的映射交叉遗传优化算法，以屏蔽电机的有效材料重量作为优化目标，研究了屏蔽电机的概念设计优化。结果显示：优化方案与原设计方案相比有效材料重量减小了9.933%，优化效果显著。同时，对屏蔽电机有效材料重量受相关尺寸参数、导线参数及槽型参数影响的敏感性进行了分析，其结果可供工程设计参考。     

核电站主泵停运控制核素的选择

主泵是核电站必不可少的设备,除维系机组的安全运行外,它还对辐射源项的控制起着不可替代的作用。选用合适的放射性核素控制主泵的停运对于压水堆机组大修时解决工期进度和减少辐射源项这一矛盾有着重要指导意义。本文结合大亚湾核电站的运行经验,指出＾１１０ｍＡｇ是对反应堆换料水池表面辐射水平有主要贡献的核素,并给出了停运主泵时＾１１０ｍＡｇ比活度的建议限值。     

中国一体化反应堆核电厂创新安全壳设计研究

中国一体化反应堆核电厂(CIP)是中国核反应堆系统设计技术国家重点实验室正在开发的新一代革新型、完全一体化的压水堆,其电功率约为300 MW.CIP采用堆内一体化布置,反应堆冷却剂系统设备以及控制棒驱动机构全部布置在反应堆压力容器内.这种一体化设计消除了传统的冷却剂回路管道,消除了大LOCA事故,具有更高的安全性.本文介绍了CIP安全壳系统方案选择、安全壳设计、安全壳设计压力的确定以及安全壳结构的计算分析.     

基于Labview的相关法测量钠流量系统设计

为满足中国实验快堆(CEFR)一回路主泵旁路钠流量计校准的需求,设计了1套基于Labview软件的相关钠流量测量系统。本文介绍了相关法的测量原理、设计的相关钠流量测量系统、对该系统的仿真试验和钠回路上的验证试验。试验结果表明,这套基于Labview的相关钠流量测量系统是可行的。本文还进行了该系统的测量误差分析,给出了减小误差的方法。该系统及其试验为CEFR一回路主泵旁路钠流量计在役校准装置的设计、调试和运行提供了依据。     

Possible design improvements and a high power density fuel design for integral type small modular pressurized water reactors

The study evaluates potential weaknesses and possible improvements for integral type small modular pressurized water reactor designs. By taking International Reactor Innovative and Secure (IRIS) as the reference design and keeping the power output as the same, a new fuel and reactor design were proposed. The proposed design relocates the primary coolant pumps and the pressurizer outside the reactor pressure vessel (RPV). Three recirculation lines and jet pumps/centrifugal pumps are introduced to provide the coolant circulation similar to Boiling Water Reactor designs. The pressurizer component is expected to be similar to the AP600 design. It is located at one of the recirculation lines. The new fuel assembly adopts 264 solid cylindrical fuel pins with 10 mm diameter and 2.3 m height, arranged at a hexagonal tight lattice configuration. Large water rods are introduced to preserve the moderating power and to accommodate finger type control rods. The resulting fuel can operate with 104.5 kW/l power density while having substantially higher margin for boiling crisis compared to typical large PWRs. Full core neutronic analysis shows that 24-month cycle length and 50 MWd/kg burnup is achievable with a two-batch refueling scheme. Furthermore, the fuel behavior study shows that the new fuel with M5 type Zircaloy cladding show fairly acceptable steady state performance. A preliminary Loss of Coolant analysis shows that the new design could be advantageous over IRIS due to its low ratio of the water inventory below the top of the active fuel to total RPV water inventory. The proposed reactor pressure vessel height and the containment volume are 30% lower than the reference IRIS design.     

49-2游泳池式反应堆全场断电事故与自然循环能力分析

在全场断电事故下,采用RELAP5/MOD3.3程序对49-2游泳池式反应堆系统热工水力参数瞬态特性进行计算分析,验证反应堆利用自然循环和自身负反应性对事故的缓解能力,并简要讨论了堆芯通道和主泵惰转对事故后果及进程的影响。计算结果表明,在49-2反应堆发生全场断电事故且紧急停堆系统失效后,反应堆可依靠自身的负反应性使反应堆处于停堆状态,并能形成稳定的自然循环,导出堆芯余热,验证了49-2反应堆在全场断电超设计基准事故中是安全的。     

Design options for safety depressurization system

The Korean Next Generation Reactor (KNGR) adopted an advanced design feature, a safety depressurization system (SDS) to rapidly depressurize the primary system in case of events beyond the design basis. Two design approaches are considered for the SDS design. The use of bleed valves similar to the ABB-CE System 80+ is design option 1, while in design option 2, the French Sebim valve is considered to provide the combined function of overpressure protection and rapid depressurization. In this paper, thermal hydraulic analysis using a best-estimate version of CEFLASH-4AS/REM is performed for a total loss of feedwater (TLOFW) event to investigate the feasibility of those two design options. For each design option, various feed and bleed (F and B) procedures are investigated for a TLOFW event. For design option 1, the required bleed capacity is determined from the CEFLASH-4AS/REM simulation according to the EPRI Advanced Light Water Reactor (ALWR) requirements. The analysis results demonstrate that the TLOFW event can be mitigated in a proper manner with a sufficient margin using design option 1. For design option 2, the operator action times for initiating the F and B are investigated by varying the number of Sebim valves and high pressure safety injection (HPSI) pumps. If the operator opens two out of the three Sebim valves in conjunction with the four HPSI pumps before a hot leg saturation condition, the decay heat removal and core inventory make-up function can be successfully accomplished. The results of the present investigation demonstrate that the two design options are both feasible.     

Electron dumps for ITER HNB and DNB beamlines

The negative ion accelerators that produce the high-energy particle beams for the neutral injection systems for the International Tokamak Experimental Reactor (ITER) also produce unwanted particles such as electrons. These electrons are emitted in a wide angular spectrum that allows some of them to directly intercept sensitive beamline components such as the cryogenic pumps. As the electrons are also subject to backscattering, indirect interception always occurs. In this article the electron spectra produced by the Heating Neutral Beam (HNB) and Diagnostic Neutral Beam (DNB) accelerators are calculated. It is shown that these are very different. It is proposed to install electron dumps in the beamlines to intercept electron power directed towards inconvenient places in the HNB and DNB beamlines.     

Assessment of the CATHARE 1D pump model

The development of an advanced model to determine the dynamic pump performance under two-phase flow conditions is presented. This model is included in CATHARE 2, version V1.3. It is based on the two-fluid six-equation CATHARE model which describes the mechanical and thermal non-equilibria.In a previous review (P. Van den Hove and G. Geffraye, The CATHARE code— one-dimension pump model, Fifth Int. Topical Meet. on Nuclear Reactor Thermal Hydraulics (NURETH-5), Salt Lake City, USA, September, 1992), various calculations were presented concerning Eva single-phase and two-phase steam–water test results in the first three quadrants. Here, the range of assessment of the first quadrant is enlarged with Eva air–water tests and Bethsy pump steam–water tests. Both pumps are mixed flow pumps, the Bethsy one being radial at the impeller outlet.Some improvements suggested in the above cited paper are tested against all single-phase liquid, single-phase vapor, two-phase steam–water, and two-phase air–water data in the first quadrant. They concern a new deviation model and head losses model, and the model of mechanical interaction between phases.     

船用堆严重事故下熔融物堆内滞留能力分析

堆芯熔化的严重事故,可能导致船用堆下封头失效、熔融物进入堆坑,危害人员及船体安全。本文采用严重事故一体化程序MAAP4,以船用堆全船断电事故为研究对象,针对低压安全注射系统投入时机、低压安全注射水流量,研究下腔室熔池形成后,投入低压安全注射系统对熔融物堆内滞留的作用。结果表明:在下腔室熔池形成后1576?s时,投入两台安全注射泵仍能有效阻止压力容器失效,实现熔融物堆内滞留;在下腔室熔池形成2646?s后,投入低压安全注射系统不能阻止压力容器失效。      

LOCA Analysis of Super Fast Reactor

This paper describes loss of coolant accident (LOCA) analyses of the Supercritical-pressure Water-Cooled Fast Reactor (Super Fast Reactor). The features of the Super Fast Reactor are high power density and downward flow cooled fuel channels for the improvement of the economic potential of the Super Fast Reactor with high outlet steam temperature. The LOCA induces large pressure and coolant density change in the core. This change influences the flow distribution among the downward flow parallel channels. It will affect the safety of the Super Fast Reactor. LOCA analysis of Super Fast Reactor is important to understand the safety features of the Super Fast Reactor. Keeping the flow rate in the core is important for the safety of the Super Fast Reactor. In LOCA, it is difficult to maintain an adequate flow rate due to the once-through coolant cycle and the downward flow cooled fuel assemblies. Therefore, the early actuation of the Automatic Depressurization System (ADS) and reduction of the maximum linear heat generation rates of the downward flow seed fuel assemblies and Low-Pressure Core Spray (LPCS) system are necessary for the Super Fast Reactor to cool the core under LOCA. Analysis results show that the Super Fast Reactor can satisfy the safety criteria with these systems.