首页> 外文会议>International conference on nuclear engineering;ASME power conference >IMPORTANCE OF REACTOR HEAT TRANSPORT SYSTEM OVERPRESSURE PROTECTION SYSTEM UNDER SEVERE ACCIDENT CONDITIONS WITH SPECIAL REFERENCE TO CANDU REACTORS
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IMPORTANCE OF REACTOR HEAT TRANSPORT SYSTEM OVERPRESSURE PROTECTION SYSTEM UNDER SEVERE ACCIDENT CONDITIONS WITH SPECIAL REFERENCE TO CANDU REACTORS

机译:在特殊事故条件下反应堆传热系统超压保护系统的重要性,特别参考坎杜反应堆

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After Fukushima, any Station Blackout capability reviews must be carried out with diligence and integrity beyond any reasonable doubt. This is especially true for nuclear reactors that did not consider a sustained loss of AC power and severe accidents in their original design basis. Integrity of the engineered barriers to release of activity must be demonstrated to showcase the extension of defense in depth to a Station Blackout and reasonable provisions for effective interventions. An important aspect of nuclear reactor heat transport system pressure boundary integrity is the ability of its over-pressure protection system to meet the challenges upon a sustained loss of AC power.The primary heat transport systems in water cooled nuclear power reactors have at least two passive safety relief valves so that at least one is available to act with the generally mandated consideration of a single failure. Design criteria for these relief valves vary but their steam relief capacity, reliability and performance must conform to the relevant ASME code or equivalent requirements. A typical PWR of about 3000 MW thermal power may usually have 3-5 such valves able to relieve upto ~200 kg/s of steam, which may be about an order of magnitude higher than required. A typical single unit CANDU reactor, on the other hand, has about 30% less thermal power but only 2 such safety relief valves with a combined steam relief capacity of about 4 kg/s or about 4 MW of thermal power equivalent at the relief set-point and a time when the decay power is ~20 MW. Installation of these valves in a CANDU reactor is also atypical; as they do not provide a direct and unobstructed path from the heat transport system but are installed downstream of another series of isolating Liquid Relief Valves emptying into an unpressurized and small degasser condenser vessel downstream of which they are mounted. These valves become critical when there is a sustained loss/depletion of engineered heat removal systems following multiple failures as in Fukushima. An unmitigated increase in heat transport pressure and a consequential breech in pressure boundary become inevitable if the core decay heat exceeds the safety valve steam relief capacity. If the ensuing failure is in the boiler tubes, a containment bypass and release of activity into atmosphere is possible.The present design of CANDU reactor Primary Heat Transport system does not seem to allow the anticipated energy relief through the safety relief valves following a sustained loss of all engineered heat sinks. This may result in uncontrolled primary heat transport system pressurization and a potential for boiler tube ruptures such that activity releases bypass the containment and expose the population to dangerously high radiation well before any evacuation. If the valves are to conform to requirements of the ASME codes under early stages of a loss of heat sinks scenario just as they must for design basis accidents (energy relief capacity greater than the heat load), it may seem that many clauses of the applicable ASME code sections for the subject valves are violated in abundance and with impunity.A containment bypass resulting from boiler tube failure caused by the faulty overpressure protection can cause fatalities that can be high with astronomical economic consequences, especially after the fuel begins to overheat. Such a containment bypass is considered to present the highest risk to public. The probability of a sustained loss of heat sinks is not insignificant and the overall risk is yet to be quantified for all instigators for any CANDU power plant. The national regulators have not required that utilities do so in a timely manner and as a condition of operating license.The paper examines the CANDU safety relief valve design criteria, lists design challenges and potential consequences during a station blackout severe accident scenario that could be well mitigated by an otherwise robust design.
机译:福岛事故发生后,任何车站停电能力检查都必须在没有任何合理怀疑的情况下以勤勉和正直的态度进行。对于没有在其原始设计基础中考虑持续损失交流电源和发生严重事故的核反应堆而言,尤其如此。必须证明工程释放障碍的完整性,以展示国防深度扩展到车站停电以及有效干预措施的合理规定。核反应堆传热系统压力边界完整性的一个重要方面是其过压保护系统能够应对持续失去交流电的挑战。水冷核电反应堆的一次传热系统至少有两个被动式安全溢流阀,以便至少有一个可以在一般性考虑单个故障的前提下起作用。这些泄压阀的设计标准各不相同,但其泄压能力,可靠性和性能必须符合相关的ASME规范或等效要求。典型的约3000 MW火力的PWR通常可以安装3-5个这样的阀门,这些阀门可以释放高达〜200 kg / s的蒸汽,这可能比所需的蒸汽量高大约一个数量级。另一方面,典型的单单元CANDU反应堆的热功率要少大约30%,但只有2个这样的安全泄压阀,其总的蒸汽释放量约为4 kg / s,或在释放装置处的等效热功率为4 MW点和衰减功率为〜20 MW的时间。这些阀门在CANDU反应器中的安装也是非典型的。因为它们不会从传热系统提供直接且畅通的路径,而是安装在另一系列的隔离液卸压阀的下游,这些阀排空到未安装压力的小型脱气冷凝器容器中,并在其下游安装。当在福岛发生多次故障后工程散热系统持续损耗/耗尽时,这些阀门就变得至关重要。如果堆芯衰减热量超过安全阀的蒸汽释放能力,那么不可避免的传热压力升高和必然的压力边界泄漏将不可避免。如果随之而来的故障是锅炉管中的故障,则有可能发生安全壳旁路并将活性释放到大气中。CANDU反应堆的一次热传输系统的当前设计似乎无法在持续损失后通过安全释放阀释放预期的能量。所有工程散热器。这可能会导致无法控制的主热传输系统增压,并可能导致锅炉管破裂,从而导致活动释放绕开安全壳,并使人员在疏散之前早已暴露在危险的高辐射下。如果阀门在散热片丢失的早期阶段符合ASME规范的要求,就像设计基准事故(节能能力大于热负荷)所必需的那样,则似乎适用的许多条款大量违反了该阀的ASME规范章节,没有受到惩罚。由于过压保护故障而导致的锅炉管故障而导致的安全壳旁路可能导致死亡人数众多,对天文经济造成重大影响,尤其是在燃料开始过热之后。这种围堵措施被认为对公众构成最高风险。散热器持续损失的可能性并不微不足道,对于任何CANDU电厂的所有激励器而言,总风险尚待量化。国家监管机构没有要求公用事业公司及时并作为操作许可的条件这样做。本文研究了CANDU安全溢流阀的设计标准,列出了在车站停电严重事故情况下的设计挑战和潜在后果,这可能是一个很好的选择。通过本来坚固的设计减轻了压力。

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