Resolution of Generic Safety Issues: Item B-55: Improved Reliability of Target Rock Safety Relief Valves (Rev. 1) ( NUREG-0933, Main Report with Supplements 1–34 )
The BWR pressure relief system is designed to prevent overpressurization of the reactor coolant pressure boundary (RCPB) under the most severe abnormal operational transient: closure of the main steam line isolation valves (MSIVs) with failure of the MSIV position switches to scram the reactor. This design function is accomplished through the use of a plant-unique combination of safety valves (SVs), power actuated relief valves (PARVs), and dual function safety/relief valves (SRVs). The majority of the valves in BWRs are commonly referred to as Target Rock SRVs.
In addition to the RCPB overpressure protection design functions of the BWR pressure relief system, a specified number of the PARVs or SRVs utilized in the pressure relief system of each BWR facility are used in the automatic depressurization system (ADS), which is one of the emergency core cooling systems. In the event of certain postulated small-break LOCAs, the ADS is designed to reduce reactor coolant system pressure to permit the low pressure emergency core spray and/or low pressure coolant injection systems to function. The ADS performs this design function by automatically actuating certain preselected PARVs or SRVs following receipt of specific signals from the protection system.
Certain safety concerns result when: (1) a valve fails to open properly on demand; (2) a valve opens spuriously and then fails to properly reseat; and (3) a valve opens properly but fails to properly reseat. The failure of a pressure relief system valve to open on demand results in a decrease in the total available pressure-relieving capacity of the system. Spurious openings of pressure relief system valves, or failures of valves to properly reseat after opening, can result in inadvertent reactor coolant system blowdown with unnecessary thermal transients on the reactor vessel and the vessel internals, unnecessary hydrodynamic loading of the containment systems' pressure suppression chamber (torus) and its internal components, and potential increases in the release of radioactivity to the environs. In addition, if the failed valve also serves as part of the ADS, a degradation in the capability of the ADS to perform its emergency core cooling function could result. This issue was documented in NUREG-0471.3
At the time of the evaluation of this issue in 1983, approximately 160 RY of operating experience had accumulated with a significant number of failures of the Target Rock valves occurring due to various causes. Studies and testing of these valves by the Owners' Group, in some cases at the suggestion of the NRC, have resulted in design changes in the valves and the issuance of several formal generic installation, operating, and maintenance instructions.3
In 1978, it was concluded223 by the staff that the inadvertent blowdown events that had occurred to date, as a result of pressure relief system valve malfunctions, had neither significantly affected the structural integrity or capability of the reactor vessel, the reactor vessel internals, or the pressure suppression containment system, nor resulted in any significant radiation releases to the environment. The staff concluded that such events, even if they were to occur at a more frequent rate than that indicated by operating experience, would not likely have any significant effects on the reactor vessel or the vessel internals. It was also concluded that pressure relief valve blowdown events would not result in offsite radiological consequences appreciably different from those encountered during a normal reactor shutdown.
With respect to the pressure-suppression containment system, the slowly progressive nature of the material fatigue mode of failure associated with the dynamic loading conditions resulting from pressure relief valve blowdown events, and the substantial fatigue life margin available in the affected structures led the staff to conclude that additional short-term actions were not required to ensure that the integrity and functional capability of the system would be maintained. In addition, existing programs to provide additional containment system structural safety margins for the long- term (i.e., the anticipated 40-year lifetime of the BWR facilities) were acceptable. The performance of these valves, however, was under continuous surveillance and the consequences of their failures were subject to review.
For potential core-melt frequency reduction, the Grand Gulf-1 BWR risk parameters were used in an analysis64 of this issue. It was assumed that a final solution (negligible frequency of Target Rock valve malfunction) had not yet been achieved. Hence, failure rate data on these valves on existing reactors were applicable to this analysis. It was presumed that reactors with MARK III containments for which full operating licenses were pending did not use Target Rock valves.
Analyses of the effects of malfunctioning valves as separate failures indicated that, for the short-term, public safety was not of concern. The resulting thermal transients, even at the current rate of these events, were not likely to create concerns over pressurized thermal shock. The potential for radioactive release to the public following a malfunction resulting in an unplanned blowdown was no greater than for a normal shutdown. However, when a valve fails to reseat simultaneously with failures on other systems, some potential for a core-melt exists. Analysis of the dominant accident sequences at Grand Gulf-1 for these events was done as part of this evaluation.
All minimal cut sets in the following four Grand Gulf accident sequences were affected: T1PQI (loss of offsite power with failure of the SRV to reseat and failures of the power conversion and RHR systems); T23PQI (normal transient with SRV reseat failure and failures of the power conversion and RHR systems); T1PQE (loss of offsite power with SRV reseat failure and failures of the power conversion and ECCS); and T23PQE (normal transient with SRV reseat failure and failures of the power conversion and ECCS).
It was assumed that the resolution of the issue would result in a reduction in the frequency of valves failing to reseat by a factor of 4. This assumption was based on the continued success of the existing remedial programs for these valves that were underway at existing BWRs. The estimated change in core-melt frequency was 4.7 x 10-6/RY.
When the frequencies for the individual release categories are multiplied by the appropriate public dose and the products are summed, the resulting estimated change in public risk was 30 man-rem/RY. Assuming 10 reactors with an average remaining life of 26.7 years affected by the issue, the total risk reduction was estimated to be 8,000 man-rem.
Industry Cost: Modifying or refurbishing SRVs on high-temperature, high-pressure steam lines was expected to require engineering, design drawings, license review, testing, travel, labor, material, QA control, and management review. This cost was estimated64 to be $75,000. In addition, new top works were estimated to cost $60,000 each and there were usually 11 SRVs/plant. Finally, it was estimated that 50 man-hours/RY would be required for operation (testing) and maintenance. There were 20 BWRs with Target Rock SRVs with an average remaining life of 26.7 years. Of these, about half had already installed new SRV top works. Thus, the total cost was estimated to be about $800,000/reactor for the remaining 10 reactors, or $8M.
NRC Cost: The NRC cost was reduced since the issue had been defined and partial solutions had been achieved. It was estimated that 4 staff-weeks/plant would be needed to support implementation. Thus, NRC cost was estimated to be about $150,000.
Total Cost: The total industry and NRC cost associated with the possible solution to the issue was estimated to be $(8 + 0.15)M or $8.15M.
Based on a potential public risk reduction of 8,000 man-rem and an estimated cost of $8.15M for a possible solution, the value/impact score was given by:
Based on the above safety priority score, the issue was given a medium priority ranking (see Appendix C). In resolving the issue, the staff found that licensees had significantly improved the performance of Target Rock SRVs and continued to evaluate and improve their performance. Licensee compliance with existing regulations, such as 10 CFR 50 Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," and 10 CFR 50.65, "Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," were sufficient for the staff to pursue additional improvements on a plant-specific basis, if needed. Thus, the issue was RESOLVED with no new or revised requirements.1765