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首页> 外文会议>ASME(American Society of Mechanical Engineers) Pressure Vessels and Piping Conference: Aging Management and Component Analysis; 20030720-20030724; Cleveland,OH; US >SIGNIFICANT CORROSION OF THE DAVIS-BESSE NUCLEAR REACTOR PRESSURE VESSEL HEAD
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SIGNIFICANT CORROSION OF THE DAVIS-BESSE NUCLEAR REACTOR PRESSURE VESSEL HEAD

机译：DAVIS-BESSE核反应堆压力容器头的严重腐蚀

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摘要

In March of 2002, significant corrosion of the Davis-Besse reactor head was discovered. The Davis-Besse reactor head is of standard construction, composed of low alloy steel and clad with stainless steel. Alloy 600 control rod nozzles penetrate the reactor head, attached with J-groove welds. During an ultrasonic inspection, three of these nozzles were found to have through-wall cracks induced by Primary Water Stress Corrosion Cracking (PWSCC). Undiscovered leakage of borated water over the course of several operating cycles from one of these nozzles led to localized cooling and wastage of the reactor head near the nozzle. This leakage, less than 0.2 gpm (0.8 l/min), was small in comparison to allowable unidentified leakage, but larger than typical PWSCC leakage. The greatest damage to the low alloy steel reactor pressure vessel head was an oblong cavity, approximately 7x5 inches (18 x 13 cm), penetrating to the stainless steel cladding. The cracks in this nozzle were axially oriented, which would previously have been considered low risk because they would not have caused control rod ejection. However, the damage led to an increase in risk of a loss of coolant accident, prolonged loss of generation, and replacement of the reactor pressure vessel head. In addition to the industry wide regulatory impact of this event, the Nuclear Regulatory Commission has indicated that there may be a need to revise the inservice inspection requirements in Section XI of the ASME Code. This paper provides a brief synopsis of PWSCC in Control Rod Drive Mechanism nozzles, describes the inspection activities that led to the discovery of both the cracking and the corrosion, and describes the extent and technical cause of the damage. Management and human performance issues that allowed the damage to progress to an advanced state are discussed, since this event would not have been noteworthy if administrative controls and programs had been properly implemented.
机译：2002年3月，发现戴维斯-贝斯反应堆头受到严重腐蚀。戴维斯-贝斯反应器头为标准构造，由低合金钢和不锈钢制成。 600型合金控制棒喷嘴穿透反应堆头部，并带有J型槽焊缝。在超声检查过程中，发现其中的三个喷嘴有因一次水应力腐蚀裂纹（PWSCC）引起的通孔裂纹。在这些喷嘴之一的几个操作循环过程中，未发现硼酸水的泄漏导致局部冷却和喷嘴附近反应堆头部的浪费。小于0.2 gpm（0.8 l / min）的泄漏，与允许的未识别泄漏相比较小，但大于典型的PWSCC泄漏。低合金钢反应堆压力容器顶部受到的最大破坏是一个长方形的腔，穿透不锈钢包壳，大约7x5英寸（18 x 13厘米）。该喷嘴中的裂纹是轴向取向的，以前不会被认为是低风险，因为它们不会引起控制棒弹出。但是，这种损坏导致损失冷却剂事故，延长发电时间以及更换反应堆压力容器头的风险增加。除了该事件对整个行业的监管影响之外，核监管委员会还表示，可能有必要修改ASME规则第XI节中的在役检查要求。本文简要介绍了控制杆驱动机构喷嘴中的PWSCC，描述了导致发现裂纹和腐蚀的检查活动，并描述了损坏的程度和技术原因。讨论了导致损害发展到高级状态的管理和人员绩效问题，因为如果适当地实施了行政控制和程序，此事件就不会引起注意。

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