共查询到20条相似文献,搜索用时 46 毫秒
1.
三门核电AP1000机组辐射防护设计分析 总被引:1,自引:0,他引:1
三门核电AP1000机组为第三代核电机组,在辐射防护设计中采用了一回路加锌、较高pH值运行、停堆氧化操作、蒸汽发生器一回路水室电解抛光、优化设备维修、优化屏蔽设计、无线剂量监测等措施,以期降低机组辐射水平和职业照射剂量。本文介绍了三门核电AP1000机组在功率运行及大修期间的辐射水平和职业照射剂量数据,并与国内CPR1000机组的相关数据进行了对比,对AP1000机组的辐射防护设计进行分析,给出了三门核电AP1000机组在辐射防护运行管理及技术改进方面的建议。 相似文献
2.
介绍了10MW高温气冷实验堆吸收球停堆控制系统的设计原则和调试过程,试验证明,在不同运行工况下该系统能实现设计功能,从而保证了反应堆运行的安全性和可靠性。 相似文献
3.
王继东 《核标准计量与质量》2011,(3):2-7
根据美国联邦法规的要求,核电厂必须针对SSE(安全停堆地震)进行设计,OBE(运行基准地震)是否作为设计输入,取决于许可证申请者确定的OBE加速度数值.介绍了美国法规、导则关于核电厂的抗震设计要求,调查了AP1000的抗震设计情况,并就AP1000抗震设计与我国抗震要求进行了对比.经分析对比可得出结论:AP1000的抗... 相似文献
4.
5.
10MW高温气冷实验堆吸收球停堆系统的设计 总被引:2,自引:1,他引:1
介绍了吸收球停堆系统的设计原则,分析了不同参数对系统设计的影响,并对吸收球停堆系统的最大可信事故进行了分析。分析表明,本吸收球停堆系统的设计能实现在任何工况下的启动和运行,不会发生失效。 相似文献
6.
沉积于一回路系统设备内壁的活化腐蚀产物是压水堆核电厂停堆工况下的主要放射性来源.文中选择CPR1000停堆换料期间放射性浓度较高的活化腐蚀产物58Co作为研究对象,分析该核素在停堆开盖过程中放射性浓度变化的影响因素,并建立相应的放射性浓度计算模型.计算结果表明,一回路净化流量和附着于设备内壁的58Co释放率是影响停堆期间一回路冷却剂58Co放射性浓度变化的主要因素,同时从理论上得出了CPR1000机组停堆净化工序能够使得一回路冷却剂内58Co放射性浓度降至相关停堆放化控制限值内的结论. 相似文献
7.
8.
9.
核电厂为运行人员提供了主控制室(MCR)作为电厂集中监控中心,并提供了与MCR实体隔离和电气隔离的远程停堆站(RSS)作为辅助控制室,以在MCR不可用时投入使用,对电厂实施监控,并将电厂带入停堆状态和导出余热。根据核安全法规、导则及标准要求,来自MCR和来自RSS的电厂控制功能须相互闭锁,不能同时执行。本文通过比较分析,研究CPR1000、EPR及AP1000堆型核电厂控制室操作模式切换方案的特点与不足,在详细研究的基础上给出控制室切换功能设计的几个基本原则,供新的核电厂控制室功能切换方案设计时参考,以设计出更为实用、简洁、安全、便利的方案。 相似文献
10.
12.
《Journal of Nuclear Science and Technology》2013,50(9):1199-1209
The ECCS performance, which mitigates a postulated catastrophic failure of the main reactor coolant piping during the full power operation, is judged to cover the consequences of LOCA occurring in other plant operational states. During Mode 3 with an accumulator isolated and Mode 4, since the normal alignment of ECCS equipments is changed from that which is available during the power operation, a potential safety issue, which involved the performance of ECCS for LOCA during Mode 3 with the accumulator isolated and Mode 4, was identified in 1985. This study is performed as the plant specific shutdown LOCA program for the power uprated Kori-3 and 4, of which the nominal core power is planned to increase by 4.5%. We determine and verify the operator action time to initiate SI following a small break LOCA in order that the peak clad temperature of fuel does not exceed the 10CFR50.46 limit of 1,477.6 K. We evaluate the 0.1524 m (6 inches) pipe break in the cold leg to develop the SI initiation time. There is a considerable margin to the 10CFR50.46 limit of 1,477.6 K in the case that the SI is manually initiated at 25 min after an operator identifies the symptom of a small break LOCA. However, in respect of the safe plant operation, we decide the operator SI initiation time as 15 min in order that the SI water is supplied to prevent the fuel heat-up during the blowdown phase of a small break LOCA. After then, we evaluate the applicability of the pre-determined SI initiation time to other small break LOCAs, which have a smaller break size, a lower initial decay heat level or a different break location. Since the peak clad temperatures of applicability evaluation cases are lower than those of the umbrella case, we confirm that the pre-determined SI initiation time can be applied to mitigate the small break LOCAs during the plant shutdown operation. The SI initiation time developed in this study will be used in the Abnormal Operating Procedure of the power uprated Kori-3 and 4 for the small break LOCAs during the plant shutdown operation. 相似文献
13.
14.
15.
16.
快中子反应堆非能动停堆系统中控制棒驱动杆磁性连接件的设计与有限元分析 总被引:2,自引:2,他引:0
为确保快中子反应堆的安全运行,提出了一种非能动性的智能触发停堆系统,完成了对该系统中的永久磁铁的设计,采用ANSYS软件对永久磁铁进行热分析并进行安全评估,验证了该系统的安全性和有效性. 相似文献
17.
高温气冷堆紧急停堆后需要快速冷却堆芯,使其达到重新启动条件,制定合理的冷却方案对于减少电厂运行成本和保护设备安全具有重要意义。本文建立了冷却系统的数学模型,对冷却过程中关键设备的传热传质过程进行了动态数值模拟。首先分析了德国高温气冷堆采用的直接冷却方案,结果表明,此方案无法避免对设备形成冷冲击或热冲击,风险性较大。进而提出了适用于我国高温气冷堆的新方案,新方案包括4个步骤:蒸汽发生器排水-卸压-预冷-冷却堆芯。动态分析表明,新方案成功地避免了冷/热冲击,大幅提高了安全性,冷却时间也在可接受范围内。 相似文献
18.
本文针对兆瓦级高温气冷堆布雷顿循环系统,采用Fortran语言开发系统分析程序TASS,包括堆芯、透平-发电机-压气机、回热器、冷却器和热管式辐射散热器等模型。通过设计值与程序计算值对比对TASS进行验证,并利用TASS对系统启动、停堆瞬态工况进行数值模拟。结果显示,通过分两阶段、阶梯式引入正反应性和提高涡轮机械的转轴速度,堆芯流量和功率匹配良好,系统可在3.5 h内完成启动过程,达到反应堆功率3 406 kW、流量14.2 kg/s的稳态运行。系统停堆过程中,反应堆可依靠自身的非能动余热排出能力,确保芯块和包壳温度与熔点间存在较大安全裕量,实现安全停堆。 相似文献
19.
《Journal of Nuclear Science and Technology》2013,50(12):1212-1223
In PWRs, loss of decay heat removal (DHR) during reactor shutdown with the reactor coolant system (RCS) partially drained may result in core boiling in a short time. The subsequent RCS pressurization could prevent water flow into the RCS by gravity feed and consequently the core would be uncovered. This paper analyzes U.S. PWR operating experience involving the DHR loss in such reduced inventory conditions. Between 1976 and 1990, reported were a total of 63 loss of DHR events which occurred during reactor shutdown with the RCS inventory reduced. Review of the event reports indicated that many loss of DHR events in reduced inventory conditions resulted from air entrainment into the DHR pumps due to lowering the reactor water level too far, loss of coolant inventory, increased pump flow and so on. The coolant heatup rates were evaluated for 12 events with use of the data such as the time elapsed from reactor shutdown actually reported. The calculated results were in reasonably good agreement with the observed ones and showed that core boiling would take place within 1 h even if the DHR loss would occur in the late stage of shutdown (for example, 30 days after the shutdown). 相似文献