Issue 40: Safety Concerns Associated with Pipe Breaks in the BWR Scram System

DESCRIPTION

Historical Background

On April 3, 1981, AEOD published draft NUREG-0785, "Safety Concerns Associated with Pipe Breaks in the

BWR Scram System."[1]As a result of the development of these safety concerns and the findings presented in the report, the NRC staff met with representatives of the BWR Regulatory Response Group and GE on April 9,

1981. A letter[2]was issued on April 10, 1981 to all BWR licensees requiring a generic evaluation of the safety concerns within 45 days of receipt and a plant-specific evaluation within 120 days of receipt.

A meeting was held with GE on April 28, 1981 to discuss the status of its generic evaluation. Subsequently, NEDO-24342[3]was submitted to the NRC by letter dated April 30, 1981.[4]

A multidisciplinary group from NRR was assembled to review the generic evaluation. A three-phase approach was developed to identify generic review objectives and describe review termination points. It was agreed that this approach would be based on establishing either: (1) a low probability for the event, (2) acceptable consequences for the event, or (3) alternate cooling systems and mitigation equipment for the event.

As the review progressed, it became evident that a sufficient data base did not exist to conservatively terminate the generic review on the basis of a quantitative risk assessment. It was equally difficult to show acceptable consequences for all scram initiators, considering the potential for an unisolable leak from the reactor coolant system into the reactor building. Thus, it was necessary to generically evaluate the mitigation capability for this scenario.

As the evaluation proceeeded, several suggestions for improving and verifying piping integrity, mitigation capability, and environmental qualifications of essential equipment were made. These suggestions are discussed

in NUREG-0803[5]which begins with a review of the licensing design basis for the SDV piping system. An evaluation of the SDV piping system integrity and an assessment of the mitigation capability follow. Finally,

each suggestion for improvement is evaluated in NUREG-0803[6]and the final guidance for resolution of this problem is presented. NUREG-0803[7]was transmitted to the BWR licensees, CP applicants, CP holders, and OL applicants by letters.[8]^,[9]These letters also requested appropriate responses to the safety concerns and guidelines presented in NUREG-0803.[10]In these letters, it has been noted that an acceptable plant specific response for this issue will conform to the final approved guidance provided in NUREG-0803.[11]

However, an additional submittal[12]was forwarded to the NRC staff by GE and the BWR Owners' Group in August, 1982 in which an analysis was presented to demonstrate the probability of a pipe break in the scram discharge volume system was negligibly small and that, therefore, this issue should not be regarded as a significant safety issue. On the basis of its review of the August 1982 submittal, the NRC staff concluded that the results of the submittal were unacceptable. However, before the submittal was formally rejected by the staff, GE and the BWR Owners' Group provided additional material which amplified the August 1982 submittal with

supporting information[13]which was presented at a meeting with the staff on February 8, 1983.

A study[14]was completed which describes the predicted response of Unit 1 at the Browns Ferry Nuclear Plant to a postulated small-break LOCA outside of the primary containment. This study is contained in the first volume of a two-volume study in which a detailed analysis of the accident sequence is presented. An estimate of the magnitude and timing of the concomitant release of the noble gas, cesium, and iodine-based fission products to the environment will be provided in Volume 2 of the study.

Safety Significance

If a break or leak exists or develops in the SDV piping during a reactor scram, this would result in the release of water and steam at 212F into the reactor building at a maximum flow rate of 550 gpm and is postulated to result in 100% relative humidity in the reactor building. The principal means of isolating this break would be to close the scram exhaust valves which are located on the hydraulic control units; however, this is dependent upon the ability to reset scram, which cannot be absolutely ensured immediately following the scram. Therefore, a rupture of the SDV could result in an unisolable break outside of primary containment, which is postulated to threaten emergency core cooling equipment by flooding areas in which this equipment is located and by causing ambient temperature and relative humidity conditions for which this equipment is not qualified.

Solution

NUREG-0803[15]provides guidance to ensure pipe integrity, detection capability, mitigation capability and qualification of the emergency equipment to the expected environment.

CONCLUSION

This issue was RESOLVED, requirements were established, and MPA B-65 was established by DL for implementation purposes.[16]