Information Notice No. 87-10, Supplement 1: Potential for Water Hammer during Restart of Residual Heat Removal
UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C. 20555-0001 May 15, 1997 NRC INFORMATION NOTICE 87-10, Supplement 1: POTENTIAL FOR WATER HAMMER DURING RESTART OF RESIDUAL HEAT REMOVAL PUMPS Addressees All holders of operating licenses or construction permits for boiling-water reactors (BWRs). Purpose The U.S. Nuclear Regulatory Commission (NRC) is issuing this supplement to Information Notice (IN) 87-10 to alert addressees to the continuing potential for water hammer in the residual heat removal (RHR) system of BWRs during a design-basis loss-of-coolant accident (LOCA) coincident with a loss of offsite power (LOOP) if the RHR system is aligned in the suppression pool cooling (SPC) mode of operation. This supplement also addresses the increased use of RHR pumps in the SPC mode due to leaking safety relief valves (SRV). It is expected that licensees will review this information for applicability to their facilities and consider actions, as appropriate, to avoid similar problems. However, suggestions contained in this information notice are not NRC requirements; therefore, no specific action or written response is required. Background Most BWRs are designed with a multifunction RHR system that includes a suppression pool cooling capability for long-term cooling following a design- basis accident. To ensure the design basis-heat removal capacity, BWRs also have technical specification limits on the temperature of the suppression pool during normal operation. Information Notice 87-10, "Potential for Water Hammer During Restart of Residual Heat Removal Pumps," dated February 11, 1987, informed licensees that the RHR system at the Susquehanna facility was susceptible to water hammer loads that could exceed allowable stresses. Such loads could result from a design-basis LOCA coincident with a LOOP if the RHR system is in the SPC mode of operation. In these circumstances, the LOOP, subsequent valve realignment, and large elevation differences may allow portions of the system to drain down to the suppression pool, leaving voids in the RHR piping. When the diesel generator reenergizes the buses in response to the LOOP, the RHR pumps will start and possibly cause water hammer damage in the voided RHR loop. In addition to concerns about possible water hammer, IN 87-10 also identified that the RHR system was being operated in the suppression pool cooling mode more than was assumed in 9705130215 IN 87-10, Supp. 1 May 15, 1997 Page 2 of 4 the design basis. The Susquehanna design basis for LOCA/LOOP assumes that the RHR system is aligned in the standby configuration (suppression pool cooling flow path valves closed) approximately 99 percent of the time and in the suppression pool cooling mode only 1 percent of the time. Contrary to this assumption, the RHR system was aligned in the suppression pool cooling mode nearly 25 percent of the time during some operating cycles because of the leaking SRVs. As an interim corrective action, the licensee modified its operating procedures to allow only one loop of RHR at a time to be operated in the suppression pool cooling mode of operation. In addition, the licensee planned to revise plant procedures to address restart of an RHR pump if it trips while operating in the suppression pool cooling mode. Although not discussed in this supplement, IN 87-10 also noted that Susquehanna had identified that the core spray (CS) system may be susceptible to water hammer damage if a LOOP occurs when the CS system is aligned in the suppression pool mixing mode of operation. To compensate, the licensee prohibited operation of the system in the suppression pool mixing mode except to perform the required surveillance testing. Description of Circumstances Experience since IN 87-10 was issued in February 1987 indicates that the suppression pool cooling mode of RHR is now frequently used at many BWRs during normal operation to remove heat from leaking SRVs and to maintain the suppression pool temperature within technical specification limits. Although the leakage appears to be more frequent in BWRs with two-stage Target Rock SRVs, other model SRVs have had similar leakage problems. Half of the 10 BWRs with Target Rock two-stage SRVs (FitzPatrick, Hatch, Hope Creek, Limerick, and Millstone 1) have experienced problems. The increasing amount of time in the SPC mode increases the probability that response to a LOCA will require realignment of RHR from the SPC mode to the low-pressure coolant injection (LPCI) mode which therefore increases the possibility of water hammer damage. Analyses by several licensees have revealed potential problems similar to those described in IN 87-10. Millstone 1 In Licensee Event Report (LER) 96-050-00 (Accession No. 9610160456), dated October 10, 1996, the licensee submitted its evaluation of a LOCA concurrent with a LOOP and a loss of direct current (dc) power. Such a scenario prevents closure of LPCI suppression pool test return valves (which are open for suppression pool cooling) and allows the LPCI flow to be diverted from the core to the suppression pool. Since the licensee's existing LOCA analysis did not consider the impact of a failure of the dc bus with the suppression pool test return valves in the open position, the plant was considered to be in an unanalyzed condition. Also, since sections of LPCI piping would be voided at the time LPCI initiates, there is the potential for damage to the system as a result of water hammer IN 87-10, Supp. 1 May 15, 1997 Page 3 of 4 The licensee also reported that SPC was operated more frequently during periods in which SRVs were leaking. For example, during the most recent period when SRVs were leaking, SPC was operated for approximately 320 hours during a six-month period. The average duration of operation was 12 hours, with both LPCI subsystems operating. WNP-2 In LER 93-001-02 (Accession No. 9406130009), dated June 3, 1994, the licensee reported that water hammer could fail the train of RHR in the SPC mode of operation if a LOOP occurred coincident with a LOCA when both trains of RHR are being operated in the SPC mode. Typically, SPC was operated to limit the suppression pool temperature increase caused by leaking SRVs. For example, operators placed two trains of the RHR system in the SPC mode on September 30, 1991, for almost 3 hours; on July 6, 1992, for more than 6 hours; and on July 11, 1992, for more than 2 hours. Limerick IN 95-47, Revision 1, "Unexpected Opening of a Safety/Relief Valve and Complications Involving Suppression Pool Cooling Strainer Blockage" (Accession No. 9511270084), reported that shortly after starting up Limerick Unit 1 from a refueling outage, elevated tailpipe temperatures indicated that three SRVs ("F," "M," and "S") were leaking. SRVs "D" and "L" were also observed to be leaking at some time during the cycle. Reactor operation continued from March 1994 until September 1995, except for two short mini-outages. The licensee frequently operated one or both trains of SPC in order to remove heat from the leaking SRVs. Discussion In addition to the long-term post-accident running of RHR pumps in the SPC mode as described in the final safety analysis report (FSAR), running RHR pumps occasionally for short durations in the SPC mode may be described in the system design basis for BWRs. However, experience indicates that some licensees may be running the RHR pumps in the SPC mode more often than was assumed in their safety analysis. Extended use (increased frequency and long duration) of the RHR system in the SPC mode during normal operation may be outside the original design-basis analysis (LOCA) assumptions. For many BWRs, the original design closing speeds of the valves in the system's cooling/test lines were specified as the standard speed (12 inches/minute) and not the fast-closing valves such as the LPCI injection valves. Because the cooling/test return valves take longer to close than the LPCI injection valves take to open, there is a potential for the core injection flow to be diverted to the suppression pool in the event of a LOCA. The emergency core cooling system performance analysis does not always consider the longer closing time of the test line valves since they are assumed to be normally closed. As the amount of time that the test valves are kept open increases, the likelihood that the valves will be open at the time of an accident increases, thereby increasing the possibility that the LPCI flow could be diverted to the suppression pool. This may be an unanalyzed condition for some BWRs IN 87-10, Supp. 1 May 15, 1997 Page 4 of 4 When operating in the SPC mode, the RHR system is more likely to undergo a water hammer event if there is a loss of station power. Since the probability of a water hammer event increases as the amount of time the system is operated in the SPC mode increases, and the likelihood of damage to the system increases with the frequency of water hammer events, operating in the SPC mode more often than assumed in the FSAR may be an unreviewed safety question as defined in 10 CFR 50.59(a)(2)(i). In addition, a significant increase in the amount of time the RHR system is operated may affect the amount and types of preventive maintenance and monitoring activities that are required to ensure that it is capable of performing its intended function. This information notice requires no specific action or written response. If you have any questions about the information in this notice, please contact one of the technical contacts listed below or the appropriate Office of Nuclear Reactor Regulation (NRR) project manager. signed by S.H. Weiss for Marylee M Slosson, Acting Director Division of Reactor Program Management Office of Nuclear Reactor Regulation Technical contacts: George Thomas, NRR (301) 415-1814 E-mail: gxt@nrc.gov Charles D. Petrone, NRR (301) 415-1027 E-mail: cdp@nrc.gov
Page Last Reviewed/Updated Tuesday, March 09, 2021
Page Last Reviewed/Updated Tuesday, March 09, 2021