海南核电机组一回路冷却剂泄漏率试验优化

Experimental Optimization of Coolant Leakage Rate in the Primary Circuit of Hainan Nuclear Power Plant

核电厂反应堆一回路冷却剂的泄漏率是机组正常运行期间监视的重要参数,运行技术规范对一回路冷却剂的泄漏率有严格要求。反应堆冷却剂系统开口后泄漏率试验(PT1/2RCP002)是超运行技术规范试验,试验压力为15.8 MPa(a),超出运行技术规范对一回路冷却剂压力的限制为(15.5±0.1)MPa,有一回路意外超压导致压力边界破裂损坏风险。对PT1/2RCP002泄漏率计算公式进行分析完善,以获得更加准确的计算结果,同时探讨用PT1/2RCP010替代PT1/2RCP002的可行性,以降低机组意外超压风险。

The leakage rate of the primary coolant in the reactor of nuclear power plant is an important parameter monitored during the normal operation of the unit.The operating technical specifications have strict requirements on the leakage rate of the primary circuit coolant.The leakage rate test after opening of reactor coolant system(PT1/2RCP002)is a test exceeding the technical specifications for operation.When the test pressure is 15.8 MPa(a),exceeding the limit of primary circuit coolant pressure(15.5±0.1)MPa in the operation technical specifications,there is a risk of pressure boundary rupture and damage caused by accidental overpressure of the primary circuit.The PT1/2RCP002 leakage rate calculation formula is analyzed and improved to obtain more accurate calculation results.At the same time,the feasibility of replacing PT1/2RCP002 with PT1/2RCP010 is discussed to reduce the risk of accidental overpressure of the unit.