共查询到18条相似文献,搜索用时 62 毫秒
1.
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
2.
在压水堆核电机组功率运行状态下,反应堆冷却剂系统内始终保持氢覆盖,然而机组在进行停堆氧化过程中,因反应堆需开口,为避免氢氧混合爆炸,需要首先除去氢气,将一回路的溶解氢含量降低到规范值以下才能开展氧化运行工作.在压水堆核电机组停堆氧化过程中,一回路溶解氢的有效控制能够决定化学控制过程是否会成为大修下行的关键路径.福清核电... 相似文献
3.
为评估压水堆核电厂燃料包壳破损时的工作人员辐射风险和燃料包壳破损程度,基于特征物理量建立一回路冷却剂系统中锕系核素质量评估方法。本文基于锕系核素的生成和迁移机理,建立了一回路冷却剂系统中锕系核素的平衡方程组,并选取3种易监测的特征物理量用以评估锕系核素向一回路冷却剂系统的释放量及其分布,并建立了一回路冷却剂系统中锕系核素质量的评估方法。然后分别采用国内在役压水堆核电厂无燃料包壳破损和有燃料包壳破损的实测数据对建立的评估方法进行了验证,验证结果表明:建立的评估方法可在无燃料包壳破损和有燃料包壳破损的情况下对一回路冷却剂系统中锕系核素质量进行评估,评估结果和预期符合。本文研究成果可为压水堆核电厂运行期间一回路冷却剂系统中锕系核素质量及其分布评估提供指导,从而优化后端的工作人员防护措施,降低辐射风险。 相似文献
4.
5.
6.
7.
冷却剂流量降低停堆保护系统整定值分析 总被引:1,自引:0,他引:1
在确保反应堆安全的基础上 ,尽量扩大电厂的运行区域是反应堆停堆保护系统设计以及整定值确定的原则。本文通过对电网运行要求的分析 ,得到了恰希玛核电厂主泵低转速和一回路低流量停堆整定值 ,随后的安全验证表明了其对冷却剂流量降低事故保护的有效性 相似文献
8.
9.
对百万千瓦级核电厂的停堆运行事故风险进行内部事件1级概率安全评价(PSA),并根据不同的停堆进程分别建立停堆PSA模型,分析经历LOI-RRA水位对电厂风险水平构成的影响。分析结果表明停堆工况下的电厂风险不可忽视,在冷停堆工况下经历LOI-RRA水位导致堆芯损坏频率明显增加。 相似文献
10.
11.
介绍了一回路冷却剂净化系统(KBE)的结构及陛能特点,研究分析了氨对硼酸型态及阴阳树脂的影响,冷却剂贮存系统(KBB)的设计缺陷。整理绘制了机组运行过程中碱金属、溶解氢的趋势图,结合机组在实际运行中出现的阴棚旨排带造成冷却剂氯离子超标、总碱金属偏离、溶解氢浓度下降等实际案例,总结优化了阳树脂氨钾饱和的开始时间、加钾量和氨浓度的控制;以及在不改变KBE初始设计的基础上增加KBE除碱金属功能,优化碱金属偏离的纠正措施。并根据实际运行结果对PUROLITE和BAYER两家公司生产的核级树脂性能进行了对比。 相似文献
12.
压水堆主回路冷却剂流经堆芯时,水中固有及特加核素受中子辐照后会产生氚,氚几乎全部以气体和液体的形式排入环境,造成氚污染。因此,氚是压水堆辐射环境影响评价的主要关注内容之一。本文以AP1000为例,根据压水堆主回路冷却剂中氚的产生途径及其随时间的变化情况建立详细的计算模型,计算压水堆主回路冷却剂中的氚活度并分析各产氚途径对氚产生量的贡献。计算结果表明:主回路冷却剂中的氚主要来源于可溶性硼的中子活化和铀裂变,对氚产生量的贡献达80%以上;在7Li纯度为99.9%时,AP1000主回路中的年产氚量为5.23×1013 Bq,锂产氚量占总量的14.01%,随7Li纯度的增加,锂产氚量的贡献呈线性减小,在7Li纯度为99.99%时,锂产氚量占总量的3.18%。其他途径对氚的产生量贡献很小,可忽略。根据以上结果,可通过控制主回路冷却剂中添加的初始硼浓度、提高燃料包壳质量、增加LiOH中7Li的纯度等多种途径来降低主冷却剂中氚的产生量,从而减少氚对环境的放射性污染。 相似文献
13.
~(16)N是压水堆一回路冷却剂中的主要活化产物,也是一回路中的主要辐射源。本文在传统~(16)N源项计算模型的基础上,根据堆芯内冷却剂的流向,考虑堆芯区域以及下降段区域的中子通量差异,将堆芯划分为活化区域以及反射区域,并建立了相应的计算模型,以典型三代压水堆核电站为例进行了计算与验证,计算结果与技术文件吻合良好,偏差在10%以内,验证了模型的正确性。最后分析了一回路典型部位的~(16)N平衡放射性活度浓度,发现在反应堆堆芯出口处最高,随着冷却剂流向逐步减少。研究结果表明,优化的计算模型可更准确计算压水堆核电站冷却剂的~(16)N源项,为分析反应堆一回路的辐射源项提供参考依据。 相似文献
14.
Guo-Qing Zhang Shuai Wang Hai-Qing Zhang Xing-Wang Zhu Chao Peng Jun Cai Zhao-Zhong He Kun Chen 《核技术(英文版)》2017,28(3)
The release of fission products from coated particle fuel to primary coolant,as well as the activation of coolant and impurities,were analysed for a fluoride salt-cooled high-temperature reactor (FHR) system,and the activity of radionuclides accumulated in the coolant during normal operation was calculated.The release rate (release fraction per unit time) of fission products was calculated with STACY code,which is modelled mainly based on the Fick's law,while the activation of coolant and impurities was calculated with SCALE code.The accumulation of radionuclides in the coolant has been calculated with a simplified model,which is generally a time integration considering the generation and decay of radionuclides.The results show that activation products are the dominant gamma source in the primary coolant system during normal operation of the FHR while fission products become the dominant source after shutdown.In operation condition,health-impacts related nuclides such as 3H,and 14C originate from the activation of lithium and coolant impurities including carbon,nitrogen,and oxygen.According to the calculated effective cross sections of neutron activation,6Li and 14N are the dominant 3H production source and 14C production source,respectively.Considering the high production rate,3H and 14C should be treated before being released to the environment. 相似文献
15.
Young-Seok Son Jee-Young Shin Ho-Gon Lim Jin-Hee Park Seung-Cheol Jang 《Nuclear Engineering and Design》2005,235(15):4021-1581
The methods developed for full-power probabilistic safety assessment, including thermal-hydraulic methods, have been widely applied to low power and shutdown conditions. Experience from current low power and shutdown probabilistic safety assessments, however, indicates that the thermal-hydraulic methods developed for full-power probabilistic safety assessments are not always reliable when applied to low power and shutdown conditions and consequently may yield misleading and inaccurate risk insights. To increase the usefulness of the low power and shutdown risk insights, the current methods and tools used for thermal-hydraulic calculations should be examined to ascertain whether they function effectively for low power and shutdown conditions. In this study, a platform for relatively detailed thermal-hydraulic calculations applied to low power and shutdown conditions in a pressurized water reactor was developed based on the best estimate thermal-hydraulic analysis code, MARS2.1. To confirm the applicability of the MARS platform to low power and shutdown conditions, many thermal-hydraulic analyses were performed for the selected topic, i.e. the loss of shutdown cooling events for various plant operating states at the Korean standard nuclear power plant. The platform developed in this study can deal effectively with low power and shutdown conditions, as well as assist the accident sequence analysis in low power and shutdown probabilistic safety assessments by providing fundamental data. Consequently, the resulting analyses may yield more realistic and accurate low power and shutdown risk insights. 相似文献
16.
Guo-Qing Zhang Shuai Wang Hai-Qing Zhang Xing-Wang Zhu Chao Peng Jun Cai Zhao-Zhong He Kun Chen 《核技术(英文版)》2017,28(2)
The release of fission products from coated particle fuel to primary coolant,as well as the activation of coolant and impurities,were analysed for a fluoride saltcooled high-temperature reactor (FHR) system,and the activity of radionuclides accumulated in the coolant during normal operation was calculated.The release rate (release fraction per unit time) of fission products was calculated with STACY code,which is modelled mainly based on the Fick's law,while the activation of coolant and impurities was calculated with SCALE code.The accumulation of radionuclides in the coolant has been calculated with a simplified model,which is generally a time integration considering the generation and decay of radionuclides.The results show that activation products are the dominant gamma source in the primary coolant system during normal operation of the FHR while fission products become the dominant source after shutdown.In operation condition,health-impacts related nuclides such as 3H,and 14C originate from the activation of lithium and coolant impurities including carbon,nitrogen,and oxygen.According to the calculated effective cross sections of neutron activation,6Li and 14N are the dominant 3H production source and 14C production source,respectively.Considering the high production rate,3H and 14C should be treated before being released to the environment. 相似文献
17.
18.
It is necessary to develop PSA methodology and integrated accident management technology during low power/shutdown operations. To develop this technology, thermal-hydraulic analysis is necessarily required to access the trend of plant process parameters and operator's grace time after initiation of the accident. In this study, the thermal-hydraulic behavior in the loss of shutdown cooling system accident during low power/shutdown operations at the Korean standard nuclear power plant was analyzed using the best-estimate thermal-hydraulic analysis code, MARS2.1. The effects of operator's action and initiation of accident mitigation system, such as safety injection and gravity feed on mitigation of the accident during shutdown operations are also analyzed.When steam generators are unavailable or vent paths with large cross-sectional area are open in the accident, the core damage occurs earlier than the cases of steam generators available or vent paths with small cross-sectional area. If an operator takes an action to mitigate the accident, the accident can be mitigated considerably. To mitigate the accident, high-pressure safety injection is more effective in POS4B and gravity feed is more effective in POS5. The results of this study can contribute to the plant safety improvement because those can provide the time for an operator to take an action to mitigate the accident by providing quantitative time of core damage. The results of this study can also provide information in developing operating procedure and accident management technology. 相似文献