ML20211P904

Engineering Evaluation Rept AEOD/E608, Reexam of Water Hammer Occurrences

ML20211P904
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Issue date:	07/14/1986
From:	Leeds E NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD)
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ML20211P894	List: ML20211P890 ML20211P897 ML20211P904
References
REF-GTECI-A-01, REF-GTECI-PI, TASK-A-01, TASK-A-1, TASK-AE, TASK-E608, TASK-OR AEOD-E608, NUDOCS 8607230416
Download: ML20211P904 (38)

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AE00 ENGINEERING EVALUATION REPORT

• UNIT: Multiple EE REPORT N0.: AE0D/E608 DATE: July 14, 1986 DOCKET N0: Multiple EVALUATOR / CONTACT: E. Leeds LICENSEES: Multiple

SUBJECT:

RE-EXAMINATION OF WATER HAMMER OCCURRENCES

SUMMARY

On November 21, 1985, Southern California Edison's San Onofre Nuclear Generating Station, Unit I experienced a partial loss of inplant ac electrical power while the plant was operating at 60 percent power. One of the most significant aspects of the event involved the failure of five safety-related check valves in the feedwater system which contributed to a severe, condensation-induced water hammer. The water hammer caused a leak in the feedwater system, damaged plant equipment, and challenged the' integrity gf the plant's heat sink. Prompted by the San Onofre event, a limited scope study was initiated to review water hammer events which have occurred since the resolution of Unresolved Safety Issue (USI) A-1, " Water Hamer," to determine if check valves are a common generic contributor to the recent water hammer events.

The study found that the underlying causes and general nature of the water hammer events which have occurred over the past five years do not appear to indicate any new generic concern not already identified and examined by the staff. Check valves were involved in only 2 of the 40 water hammer events evaluated. Furthermore, check valves were found to have been specifically cited as contributing to only five of the almost 200 water hammer events evaluated since 1969. Therefore, the study concludes that check valve leakage or failure is not a common generic cause of water hammer.

The study also found that the occurrence of steam generator water hammer (SGWH) events has declined substantially during the past five years as predicted by fhe staff in the resolution of USI A-1. This appears to be the result of the implementation of modifications described in Branch Technical Position (BTP)

ASB 10.2 at many domestic pressurized water reactors (PWRs). However, 3 of the 4 SGWH events reported since the resolution of USI A-1 have occurred at PWRs which had not implemented the J-tube modification described in BTP ASB 10.2. Presently,13 operating PWRs apparently have not yet modified the bottom discharge steam generator feedrings. This study concludes that implementation of the J-tube modification at these plants could aid in pre-venting SGWH events in the future.

As a result of the water hammer occurrence at San Onofre Unit 1, the Office of Nuclear Reactor Regulation is conducting a review of water hammer events reported since 1981. This review is being performed in an effort to assess the need to reopen USI A-1. The study suggests that NRR use the information, analysis and evaluation contained in this report to support their assessment of the issue.

• This document supports ongoing AEOD and NRC activities and does not represent the position or requirements of the responsible NRC program office.

8607230416 860714 PDR ORG NEXD 1 PDR

INTRODUCTION On November 21, 1985, Southern California Edison's San Onofre Nuclear Generating Station, Unit 1 (SONGS-1) experienced a partial loss of inplant ac electrical power while the plant was operatir.g at 60 percent power (Ref.1). Following a manual reactor trip, the plant lost all inplant ac power for 4 minutes. One of the most significant aspects of the event involved the failure of five safety-related check valves in the feedwater system whose prior degradation had occurred without detection. The backleakage allowed by the failed check valves caused voiding of long horizontal sections of the feedwater lines.

Steam from the steam generators (SGs) filled the voided feedwater lines and when cold auxiliary feedwater flow was initiated a condensation-induced SGWH occurred. The SGWH caused a leak in the feedwater system, damaged plant equipment, and challenged the integrity of the plant's heat sink. Prompted by the SONGS-1 water hammer event caused by the multiple check valve failures, a limited scope study was initiated to review water hammer events which have occurred since the resolution of USI A-1, " Water Hammer," to determine if check valves are a common generic contributor to the recent water hammer events.

In the resolution of USI A-1, the staff noted that line voiding was the single most frequent cause of water hammer events (Ref. 3). However, the staff'did not specifically identify check valve leakage as a significant contributing cause of water hammer. Check valve leakage is only one of many mechanisms which can lead to line voiding and the staff 2.cknowledged that line voiding can occur by many means. The operational data base used by the staff in the resolution of USI A-1 is contained in NUREG/CR-2059. This NUREG document provides a compilation of data concerning known and suspected water hammer events which occurred in domestic nuclear power plants from January 1, 1969 to May 1, 1981 (Ref. 2). Of the 148 reported water hamer events summarized in NUREG/CR-2059, check valves were cited as contributing to water hammers in only 3 of the events. Based on the pre-1981 operational data, it did not appear evident that check valve leakage or failure was a significant common generic cause of water hammer.

DISCUSSION Operational Data To perform the study, a search of the Sequence Coding and Search System (SCSS)

Licensing Event Report (LER) data base was conducted. The LER data base was searched for water hammer events which have been reported at all domestic light water reactors (LWRs) from January 1,1981 to December 31, 1985. However, not all water hammer events must be reported via an LER. Additional data sources (preliminary notifications and daily reports) were also searched for water hammer events. A total of 40 water hammer events at various LWRs were found from all sources for analysis and evaluation. A summary of each of the events collected from these searches is presented in Appendix A. The summary informa-tion for each event includes the plant name, the event date and a brief description of the event.

Analysis of Events At the completion of USI A-1, the staff issued NUREG-0927, Revision 1, "Evalua-tion of Water Hammer Occurrence in Nuclear Power Plants," which set forth the technical findings relevant to the issue of water hammer. NUREG-0927 contains a set of tables in which the staff categorized water hammer events by plant 2

type [i.e., PWR or boiling water reactor (BWR)], the system affected, number ,

of events per system and, the underlying cause of each water hammer event. The e

underlying cause categorization described in NUREG-0927 provides the initiating mechanism of the event (i.e., flow into voided lines, steam buoble collapse, steam-water entrainment) but does not provide the component level cause of the event. 1 l

For ease of comparison of the data compiled in this study (1981-1985) with the l data evaluated in the resolution of USI A-1 (1969-1980), tables similar te those presented in NUREG-0927 were constructed and are presented in Table 1, "BWR Water Hammer Events," and Table 2, "PWR Water Hammer Events." In addition to identifying the underlying cause of each water hammer event, the tables also include a " Comments" column which describes significant contributing factors in each event. It should be noted, however, that detailed information for each event was not always available. In several cases, the events were reported as " suspected" water hammers. In these cases the damage incurred was discovered sometime after the event and was attributed to a probable water hammer based on the nature of the damage (e.g., bent hangers and frozen snubbers).

As shown in Tables 1 and 2,18 of the reported water hammer events occurred at BWRs and 22 of the events occurred at PWRs. This data indicates a significant difference in one of the key findings reported in NUREG-0927. In NUREG-0927, the staff found that BWRs reported a higher frequency of water hammer events than PWRs (81 BWR events to 67 PWR events). This was attributed to two factors; (1) the susceptibility of BWR emergency core cooling system (ECCS) lines to leakage-caused voiding because of the low elevation of the suppression pool (which is the ECCS water source); and (2) the presence of more steam-water interfaces in BWRs than in PWRs. The higher frequency of water hammer events experienced at PWRs than BWRs over the past 5 years may be partly explained by a finding in NUREG-0927. The staff found that approximately half of all reported water hammer events occurred during preoperational testing and the first year of commercial operation. They suggested that the learning process and increased operator awareness of the potential for water hammers may result in a decrease in the frequency of the events as a plant's staff became more experienced. In the time period evaluated in this study, more PWR events were reported at preoperational/first year of commercial operation plants (7) than at similar BWRs (2).

As previously identified in NUREG-0927, Table 1 shows that the BWR systems most susceptible to water hammers continue to be the residual heat removal (RHR) and high pressure coolant injection (HPCI) systems. The table also shows that flow into voided (or partially voided) lines and steam-water entrainment continue to be the most frequently cited underlying causes for water hammer events.

Although leakage past an isolation valve was identified as a contributor in one of the RHR system water hammer events, check valve leakage or failure was not identified as contributing to any of the BWR events. In general, therefore, Table 1 does not appear to indicate any new generic implications concerning BWR water hammers beyond those previously identified in NUREG-0927.

Table 2 shows that 22 water hammer events were reported at PWRs from 1981 through 1985. The PWR systems most susceptible to water hammers were the feedwater system (4 events) and the steam generators (4 events). As previously

! identified in NUREG-0927, line voiding and steam-water entrainment continue to be the most frequent water hammer initiating mechanisms. Isolation valve (other than check valve) leakage was cited as a contributor in two of the 22 water hammer events at PWRs. Only one reported water hammer event (other than the SONGS-1 event) involved check valve backleakage. The other event involved 3

a feedline water hammer event which occurred at Salem-2 on April 6, 1984 and was

, reported in LER 84-011. At Salem-2, a motor-operated stop check valve located between the loop 3 feedwater regulating valve and its associated steam generator (SG) failed to close against 1000 psig SG pressure. The partially open check valve allowed backleakage to void the upstream section of the feedline, which eventually led to the water hammer event which occured during stroke time testing of the feedwater regulating valve. It is believed that the check valve failed to fully close because of crud buildup. Magnetite was found in the bowl section of the valve in the post-event inspection. (A more detailed description of the Salem-2 event is presented in Appendix A.)

Of the 40 reported water hammer events evaluated in this study, only two events (the November 21, 1985 event at 50NGS-1 and the April 6, 1984 event at Salem-2) involved check valve leakage or failure. The data compiled in NUREG/CR-2059 indicates that check valves were involved in only 3 of 148 water hammer events.

Therefore, check valves were specifically cited as contributing to only 5 of the almost 200 water hammer events evaluated since 1969. Operational data appears to continue to indicate, therefore, that check valve leakage or failure is not a significant common generic cause of water hammer.

Steam Generator Water Hammer (SGWH) Events ,

In NUREG-0927, the staff reported that the single most important type of water hammer events (27 of 67 PWR events) occurring in PWRs from 1969 through 1980 were SGWHs. The substantial decrease of SGWH events over the past 5 years (4 of 22 PWR events) was expected by the staff and stated in a key finding of NUREG-0927 which says in part that:

"Following the implementation of design features and testing contained in Branch Technical Position (BTP) ASB 10.2, the frequency of steam generator water hammer in top feedring design steam generators has been essentially eliminated."

In NUREG-0927, the staff specifically states that SGWH has been essentially eliminated only in plants conforming to the measures contained in BTP ASB 10.2.

Therefore, it is significant to note that 3 of the 4 recently reported SGWH events occurred at plants which did not modify the SG feedrings with the J-tube

.fix described in BTP ASB 10.2 (Ref. E The four SGWH events were reported to have occurred at SONGS-1 (two events: a suspected SGWH reported in LER 81-006 and the SGWH event of November 21,1985),

Maine Yankee and Calvert Cliffs-2. Calvert Cliffs-2 was the only plant of the three which had installed J-tubes on the SG feedrings prior to a SGWH event. In the April 21, 1984 event at Calvert Cliffs, the SG level had been drained below the level of the feedring for over 30 minutes during testing of a newly installed motor driven auxiliary feedwater pump. During the 30 minutes SG water level remained below the level of the feedring, water in the feedring drained through the slip fit joint where the feedring enters the thermal sleeve, allowing steam to fill the feedring. Leakage past the feedwater regulating valve was significant enough that when the feedwater isolation valve was reopened at the conclusion of the test, a severe SGWH occurred. Thus, it appears that the J-tube modification only extends the time required for the feedring to drain once uncovered and will not, in and of itself, prevent SGWHs 4

from occurring. A more completa descrip2 ion of the Calvert Cliffs SGWH event

, which includes a discussion of the root cause of the event (personnel not following established procedures) is presented in Appendix A.

Prior to the SGWH event of January 25, 1983, Maine Yankee had not experienced a SGWH since commercial operation began in 1972. Therefore, plant modifications to prevent SGWHs were not considered necessary (Ref. 4). At the time of the SGWH event, the Maine Yankee SGs each used an unmodified, bottom discharge feedring. Immediately prior to the SGWH, feedwater flow to the SGs was lost for approximately 15 minutes. Data indicates that the SG feedrings were uncovered for about ten minutes, which is more than sufficient time for the feedrings to empty through the bottom discharge nozzles (Ref. 5). It has been fcund that a bottom discharge feedring can be drained of water and filled with steam within 1 or 2 minutes after the feedring is uncovered if feedwater flow has been terminated (Ref. 4). The subsequent water hammer, which occurred after the manual initiation of auxiliary feedwater flow, caused a pre-existing crack in the main feedwater line to the No. 2 SG to become a through-wall crack (see Appendix A for a more detailed event description). The licensee's corrective actions included sealing the feedring bottom discharge nozzles and installing twenty-eight J-tubes on the top of the feedrings.

At SONGS-1, damage to the thermal sleeve and feedring assembly discovered in the "B" SG on May 4, 1981, was suspected to have been caused by water hammer, flow induced vibration and thermal cycling. Specifically, the licensee reported that the thermal sleeve deformation was attributed to water hammer, which caused the sleeve to expand; reducing the radial gap between the sleeve and generator wall. In regard to the concern for SGWH at SONGS-1, the staff stated, in part, that:

" San Onofre 1 has short horizontal feedwater pipe (less than 3 feet) leading to the SG inlet. Steam generators still use the ' unmodified' feedring with bottom discharge holes. The auxiliary feedwater flow at the plant can only be started manually; this allows the plant operator to feed the SGs with heated main feedwater whenever possible."

The staff has accepted the present implementation at San Onofre 1. However, this matter will be re-examined if any SGWHs occur at the plant in the future (Ref. 4).

A regulatory review of the potential for SGWH at SONGS-1 is presented in NUREG-1190, " Loss of Power and Water Hammer Event at San Onofre, Unit 1, on November 21, 1985." In the regulatory review, the staff notes that the thrust of the NRC concern was directed at the prevention and mitigation of consequences of SGWH and not at preventing gross voiding of the feedwater lines which occurred during the November 21, 1985 SONGS-1 event.

Although the five check valve failures contributed significantly to the magni-tude of the SGWH event on November 21, 1985, at SONGS-1 and the subsequent damage incurred, the conditions for a limited severity SGWH may have existed even if the check valves had not failed. During the event, when the unit tripped, SG 1evel dropped below the level of the feedring, initiating the auxiliary feedwater (AFW) system. However, the electric AFW pump was de-energized due to the partial loss of implant ac electrical power and, because of a programmed warmup period for the turbine, the steam-driven AFW pump took 3 1/2 minutes to deliver flow. Thus, for 3 to 4 minutes no flow was being provided to the SGs (Ref. 1). Since, as previously stated, only 1 to 2 5

' minutes are required to allow a bottom discharge feedring to drain and refill with steam, it is possible that a limited severity SGWH would have occurred,

$ even if the check valves did not permit gross voiding of the horizontal feedwater lines. The J-tube modification described in PTB ASB 10.2 is designed to prolong the tine necessary to drain the feedring during a loss of feedwater event. If this postulated scenario had actually occurred, the J-tube modification would probably have prevented a SGWH.

NUREG-0918, " Prevention and Mitigation of Steam Generation Water Hammer Events in PWR Plants," contains a summary of the SGWH modifications implemented at 38 operating plants. When NUREG-0918 was published (1982), 14 PWR units still used unmodified, bottom discharge feedrings in the SGs. Since the publication of NUREG-0918, at least one additional plant, (i.e. Maine Yankee) implemented the J-tube modification. The staff approved the SGWH modifications and arrangements implemented at all Westinghouse and Combustion Engineering designed plants on an individual plant-by-plant basis. The approval was conditional, however, and the staff stated that the situation would be reexamined at any plant which experienced a SGWH in the future.

The damage caused by SGWH events has generally been confined to the feedring and its supports and to the SG feedwater nozzle region. Damage to feedwater line snubbers and supports has also occurred at some plants. However, more' significant damage caused by SGWHs include a fractured weld in a feedwater line at Indian Point-2 in 1972 and a through-wall crack in a main feedline at Maine Yankee in 1983. Although the staff determined that water hammer events are of limited safety significance, SGWHs can challenge the integrity of a plant's heat sink and can cause substantial equipment damage.

FINDINGS AND CONCLUSIONS Forty actual or suspected water hammer events have been reported to have occurred at domestic LWRs during the period from January 1, 1981 through December 31, 1985. Of these, 18 events occurred at BWRs and 22 events occurred at PWRs. The underlying causes and general nature of the water hammer events do not appear to indicate any new generic concern beyond those already identified and examined by the staff. In the resolution of USI A-1, the staff stated that the total elimination of water hammer occurrence is not feasible, due to the co-existence of steam, water and voids in various nuclear plant Tystems. The specific hardware and operating procedures previously identified by the staff for the resolution of the water hammer issue remain important measures which should be taken to reduce the incidence and mitigate the consequences of severe water hammers in PWR and BWR systems.

In the resolution of USI A-1, the staff identified line voiding as the single most frequent cause of water hammer events. This study also found that line voiding continues to be the most frequently cited underlying cause of water hammer events. Check valve leakage or failure is one of many mechanisms that can create voids in piping. Operational data indicates, however, that check valve leakage or failure is a relatively minor contributor to the cause of water hammer events. Only 2 of 40 events reported from January 1, 1981 through December 31, 1985 specifically cite check valve leakage as a contributing factor in the water hammer event. A review of a previous study, which evaluated water hammer events reported from January 2,1969 through May 1,1981, found that 6

only 3 of 148 events involved check valve leakage. Therefore, it appears that check valve leakage or failure is not a significant common generic cause of water hammer.

The study also found that the occurrence of SGWH events has declined substan-tially during the past five years as predicted by the staff in the resolution of USI A-1. This appears to be the result of the implementation of modifications described in BTP ASB 10.2 at many domestic PWRs. Three of the four SGWH events reported since the resolution of USI A-1, however, have occurred at PWRs which had not implemented the J-tube modification described in BTP ASB 10.2. Presently, approximately 13 operating FWRs have unmodified, bottom discharge SG feedrings. Most of the plants with unmodified SG feedrings have never experienced a SGWH. Nevertheless, implementation of the J-tube modification at these plants could prove to be an effective measure in pre-venting SGWH events in the future.

SUGGESTION As a result of the water hammer occurrence at San Onofre Unit 1 on November 21, 1985, the Office of Nuclear Reactor Regulation is conducting a review of water hammer tccurrences reported since 1981. This review is being perfonned in an effort to assess the need to reopen USI A-1. It is suggested that NRR use the information, analysis and evaluation contained in this study to support their assessment of this issue.

7

REFERENCES

1. U. S. Nuclear Regulatory Commission, " Loss of Power and Water Hammer Event at San Onofre, Unit 1, on November 21, 1985," USNRC Report NUREG-1190, January, 1986.

2. Chapman, R.L. et. al, EG&G Idaho, Inc. " Compilation of Data Concerning Known and Suspected Water Hammer Events in Nuclear Power Plants," USNRC Report NUREG/CR-2059, May 1982.

3. U. S. Nuclear Regulatory Commission, " Evaluation of Water Hammer Occurrence in Nuclear Power Plants," USNRC Report NUREG-0927, Revision 1, March 1984.

4. U. S. Nuclear Regulatory Commission, " Prevention and Mitigation of Steam Generator Water Hammer Events in PWR Plants," USNRC Report NUREG-0918, November, 1982.

5. E. V. Imbro, " Technical Review Report: Water Hammer in the Main Feedwater System.Resulting in a Feedwater Line Crack," USNRC Report AE00/T337, '

November, 1983.

8

Table 1. BWR Water Hammer Events Number of System Events Underlying Causes Comments RHR Shutdown cooling 4 Procedures (1) Isolation valve Voided lines (3) leakage (1)

LPCI 1 Unknown Pool Cooling 1 Unknown Condensing 1 Unknown HPCI 5 Design & procedures (2)

Steam-water entrainment(1)

Voided lines (1)

Unknown , ,

Main Steam 1 Steam-water entrainment Inadequate draining procedures Core Spray 1 Unknown Isolation Condenser 1 Steam-water entrainment Inadeauate draining procedures Service Water 3_ Voidedlines(3) Procedures (2)

Total 18 9

, Table 2. PWR Water Hammer Events Number of System Events Underlying Causes Comments Feedwater 4 Unknown (1) Stuck open check Steam bubble valve at Salem-2 collapse (1) 4/6/84 Voided lines (1)

Procedures (1)

Steam Generator 4 Unknown (1) Line voiding due Procedures (2) to check valve Voided lines (1) leakage at SONGS-1 11/21/85 RHR 3 Voided lines (3) Inadequate procedures (1)

. Isolation valve leakage (1)

Design deficiency (1)

SG Blowdown 3 Voided lines (1) Isolation valve Piping Procedures (1) leakage (1)

Unknown (1)

CVCS 2 Voided line (1) Maintenance error (1)

Steam bubble collapse (1)

Steam supply 3 Steam water Drain pot operation (2) to auxiliary entrainment (3) Failed heat tracing (1) feedpump Main Steam 1 Steam water Operator error entrainment Steam supply 1 Steam-water Drain pot operation to Main Feed entrainment pump Service Water 1_ Voided lines l Total 22 10

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1. 9 APPENDIX A l

I

.i 1

'l

EVENT DATE 03/28/81 _

00CKET:325 BRUNSWICK 1 TYPE:BWR REGION: 2 NSSS:GE ARCHITECTURAL ENGINEER: UECX FACILITY OPERATOR: CAROLINA POWER & LIGHT CO.

ABSTRACT:

WHILE PERFORMING A REVIEW 0F THE PIPE SUPPORTS ON THE HPCI INJECTION LINE, FOUR SUPPORTS (2 SNUBBERS AND 2 PIPE GUIDES) WERE FOUND DAMAGED.

AS A RESULT OF THIS DISCOVERY, ALL ECCS LINES ON BOTH UNITS WERE INSPECTED TO IDENTIFY ANY OTHER POSSIBLE DAMAGE. THE ENCLOSED TABLE IDENTIFIES THE 25 PROBLEMS FOUND. ALL DAMAGED SUPPORTS WERE ATTRIBUTED EITHER DIRECTLY OR INDIRECTLY TO WATER HAMMER IN THEIR RESPECTIVE SYSTEM. ALL SUPPORTS WERE REPAIRED AND STRENGTHENED AS REQUIRED. AN ATTEMPT IS BEING MADE TO DETERMINE THE CAUSE OF THE WATER HAMMER EVENTS S0 THAT PERMANENT CORRECTIVE ACTION CAN BE -

DETERMINED AND IMPLEMENTED.

EVENT DATE 04/14/81 DOCKET:325 BRUNSWICK 1 TYPE:BWR REGION: 2 NSSS:GE ARCHITECTURAL ENGINEER: UECX FACILITY OPERATOR: CAROLINA POWER & LIGHT C0.

ABSTRACT:

DURING UNIT POWER OPERATION, THE SNUBBER SHAFT OF HYDRAULIC SNUBBER 1-E11-47SS326 BROKE AS A RESULT OF WATER HAMMER OF THE A RESIDUAL HEAT REMOVAL (RHR) SYSTEM STEAM CONDENSING PIPING, WHICH OCCURRED WHEN THE SYSTEM WAS STARTED TO RECIRCULATE THE SUPPRESSION P0OL FOR SAMPLING.

THE SNUBBER IS LOCATED DOWNSTREAM 0F THE SUBJECT PIPING INLET PRESSURE CONTROL VALVE, 1-E11-F051A. TECH SPEC 3.7.5, 6.9.1.81. A STEAM P0CKET IN THE PIPING, DUE TO LEAKAGE PAST THE F051A RESPECTIVE UPSTREAM ISOLATION VALVE, E11-F025A, CAUSED THE WATER HAMMER. 4755326 WAS REPAIRED, TESTED AND RETURNED TO SERVICE. TO PRECLUDE FUTURE SIMILAR EVENTS, THE RHR STEAM CONDENSING MODE OF UNIT 1 WAS ELIMINATED DURING THE 1985 REFUEL / MAINTENANCE OUTAGE, PER PLANT MODIFICATION.

l THE RESPECTIVE PIPING OF UNIT 2 WILL BE MODIFIED IN A LIKEWISE MANNER.

EVENT DATE 07/04/81 DOCKET:220 NINE MILE POINT 1 TYPE:BWR REGION: 1 NSSS:GE ARCHITECTURAL ENGINEER: SWXX FACILITY OPERATOR: NIAGARA M0 HAWK POWER CORPORATION ABSTRACT:

WHILE MODE SWITCH WAS IN STARTUP AND REACTOR WAS AT LOW PRESSURE THE EMERGENCY CONDENSOR LOOP #11 WAS REMOVED FROM SERVICE BECAUSE OF A WATER HAMMER. THE CAUSE IS THE RESULT OF MAIN STEAM ISOLATION VALVES CLOSED, LOW REACTOR PRESSURE AND NO CONDENSOR VACUUM CAUSING INADEQUATE DRAINING 0F #11 EMERGENCY CONDENSOR SYSTEM. THE ACTION 4 TAKEN BY THE SSS WAS TO ISOLATE #11 E.C. LOOP. THE LOOP COOLED DOWN AND WAS RESTORED TO SERVICE ABOUT 11 HOURS LATER. FUTURE ACTION IS BEING INVESTIGATED TO CHANGE PROCEDURES WHEN THE MSIV'S ARE CLOSED AND '

REACTOR IS ISDLATED, TO OPEN THE MAIN STEAM LINE DRAIN VALVE TO #11 TURBINE BUILDING EQUIPMENT SUMP TO DRAIN CONDENSATE FROM LINES.

EVENT DATE 09/03/81 DOCKET:237 DRESDEN 2 TYPE:BWR REGION: 3 NSSS:GE ARCHITECTURAL ENGINEER: SLXX FACILITY OPERATOR: COMMONWEALTH EDISON CO.

REFERENCE LERS:

1 237/70-035 ABSTRACT:

MEINTENANCEPERSONNELNOTICEDSEVERALBROKENSWAYSTRUTS,ANDHIGH ENERGY LINE BREAK RESTRAINTS ON THE HPCI STEAM LINE 2-2305-10. HPCI SYSTEM WAS DECLARED IN0PERABLE. PREVIOUS OCCURRENCE 50-237/70-35. '

SUSPECTED CAUSE OF THE BROKEN HANGERS IS EXCESSIVE LINE VIBRATION OR WATER HAMMER. WELDS WERE BROKEN ON THE SWAY STRUT ADJUSTING SCREW AND THE SUPPORT BOLTS WERE PULLED OUT ON THE HIGH ENERGY LINE BREAK RESTRAINT. THE HANGERS WERE REINSTALLED AND ADJUSTED. A DRAIN LINE WAS INSTALLED ON THE LOW POINT OF THE PIPE RUN TO ELIMINATE ANY WATER BUILDUP IN THE LINE. ANALYSIS FOR LONG TERM FIX IS BEING PERFORMED BY EDS NUC.

1 l

EVENT DATE 10/05/81 DOCKET:237 DRESDEN 2 TYPE:BWR REGION: 3 NSSS:GE ARCHITECTURAL ENGINEER: SLXX FACILITY OPERATOR: COMMONWEALTH EDISON C0.

REFERENCE LERS:

1 237/81-057 ABSTRACT: .

EDS NUCLEAR PERSONNEL NOTICED A BROKEN PIPE RESTRAINT, 1510-12 ON CCSW LINE 2-1510-16. THE 2A CCSW LOOP WAS DECLARED INOPERABLE. ALL REQUIRED SURVEILLANCES WERE PERFORMED PER TECH. SPEC. 4.5.B.3.

PREVIOUS OCCURRENCE: 50-237/81-57. PROBABLE CAUSE WAS DUE TO HYDRODYNAMIC LOAD DUE TO AN AIR VOID OR VACUUM IN THE SYSTEM DURING '

STARTUP 0F THE CCSW PUMP (S). THE HANGER WAS REDESIGNED AND REINSTALLED WITH NEW ANCHOR BOLTS. MODIFICATIONS HAVE BEEN INITIATED TO KEEP THE CCSW LINES FULL 0F WATER AT ALL TIMES TO PREVENT WATER HAMMER.

EVENT DATE 10/28/81 DOCKET:366 HATCH 2 TYPE:BWR REGION: 2 NSSS:GE ARCHITECTURAL ENGINEER: BESS FACILITY OPERATOR: GECRGIA POWER CO.

REFERENCE LERS:

1 366/81-120 ABITRACT:

ON 10-28-81 AND 11-12-81, A WALKDOWN OF THE HPCI SYSTEM WAS PERFORMED AND IT WAS OBSERVED ON 10-28-81 THAT A RIGID RESTRAINT ON THE HFCI PUMP DISCHARGE LINE WAS BROKEN AND ON 11-12-81 THAT A HPCI PUMP DISCHARGE LINE HANGER WAS BENT. THE A/E'S EVALUATION IN EACH CASE, FOUND ONLY A REDUCTION OF THE STRESS SAFETY FACTOR; HPCI WAS NOT DECLARED INOPERABLE. AN A/E INVESTIGATION CONCLUDED THAT THE STRUT FAILURE WAS A RESULT OF A DYNAMIC LOAD ASSOCIATED WITH A WATER HAMMER.

THE RSSA SUPPORTS WERE REPAIRED & THE SYSTEM WAS RETURNED TO ITS ORIGINAL DESIGN STATUS. A VISUAL INSPECTION OF ALL ACCESSIBLE ECCS SUPPORTS WAS PERFORMED BY THE SITE AND NO FAILURES WERE OBSERVED.

EVENT DATE 01/21/82 DOCKET:331 ARN0LD TYPE:BWR REGION: 3 NSSS:GE ARCHITECTURAL ENGINEER: BECH FACILITY OPERATOR: IOWA ELECTRIC LIGHT & POWER C0.

SYMBOL: IEL REFERENCE LERS:

1 331/77-008 2 331/77-029 3 331/79-025 ABSTRACT:

THE LPCI SYSTEM AND THE "A" RHR SERVICE WATER SYSTEM WERE DECLARED INOPERABLE AS A RESULT OF AN ENGINEERING EVALUATION WHICH CONCLUDED A PORTION OF THE "A" RHR (LPCI) SYSTEM PIPING MAY HAVE BEEN STRESSED IN EXCESS OF CODE ALLOWABLES. RHRSW SYSTEM SNUBBER GRC-1-SS-56 WAS FOUND INOPERABLE. A 7 DAY LC0 WAS ENTERED. THERE HAVE BEEN 3 PREVIOUS SIMILAR EVENTS IN A DIFFERENT SECTION OF RHR PIPING. (R0 77-8, 29, 79-25). EXACT CAUSE UNKNOWN. WATER HAMMER IS THE MOST PROBABLE CAUSE. SNUBBER GBC-1-SS-56 WAS REPLACED AND AN RHR SYSTEM VALVE WHICH WAS FOUND DAMAGED WAS REPAIRED. BOTH SYSTEMS WERE TESTED AND DECLARED OPERABLE.

EVENT DATE 02/19/82 DOCKET:331 ARNOLD TYPE:BWR REGION: 3 NSSS:GE ARCHITECTURAL ENGINEER: BECH FACILITY OPERATOR: IOWA ELECTRIC LIGHT & POWER C0.

REFERENCE LERS:

1 331/77-008 2 331/77-029 3 331/79-025 4 331/82-008 ABSTRACT:

"A" RHR LOOP WAS DECLARED INOPERABLE DUE TO FAILED SNUBBER GBB-4-SS-211 AND HANGER GBB-4-H10. THIS MADE LPCI INOPERABLE REQUIRING A 7-DAY LC0 PER TECH SPEC 3.5.A.5. RHR CROSS-TIE LINE WAS PROMPTLY CLOSED. IMMEDIATELY AND DAILY THEREAFTER, THE OPERABILITY OF CONTAINMENT SPRAY, BOTH CORE SPRAYS, AND DIESEL GENERATORS WAS DEMONSTRATED. "A" RHR WAS MADE OPERABLE AFTER 6.5 DAYS. FOUR SIMILAR OCCURRENCES OF WATER HAMMER DAMAGE TO RHR. (R0 77-08, 77-29, 79-25, 82-08). INSPECTION REVEALED ADDITIONAL MINOR DAMAGE AND SIGNS OF PIPE MOVEMENT ON "A" RHR. INADEQUATE FILL SYSTEM PERFORMANCE AND INADEQUATE VENTING OF "A" RHR PERMITTED ACCUMULATION OF NONCONDENSABLES RESULTING IN WATER HAMMER., SNUBBER REPLACED AND HANGER REPAIRED.

SYSTEM FILLING AND VENTING PROCEDURES HAVE BEEN MODIFIED.

EVENT DATE 06/04/83 DOCKET:333 FITZPATRICK TYPE:BWR REGION: 1 NSSS:GE ARCHITECTURAL ENGINEER: SWXX FACILITY OPERATOR: POWER AUTHORITY OF THE STATE OF NY ABSTRACT:

DURING A ROUTINE SNUBBER INSPECTION AT THE BEGINNING 0F A REFUELING OUTAGE, THE BASEPLATE OF A SNUBBER SUPPORT ON THE TORUS C0OLING LINE AT THE A RHR SYSTEM WAS FOUND WITH CONCRETE ANCHOR BOLTS PULLED FROM THE CONCRETE EMBEDMENT. THIS SNUBBER IS PEQUIRED OPERABLE DURING PLANT OPERATION BY TECH SPEC SECTION 3.6.I. REPAIRS TO THIS SUPPORT PFSK-1952 WERE COMPLETED PRIOR TO STARTUP. A REDUNDANT RHR SYSTEM WAS AVAILABLE. THE APPARENT CAUSE OF THE FAILED SH0CK SUPPRESSOR WAS WATER HAMMER ALTHOUGH N0 SPECIFIC EVENT WAS IDENTIFIED AS CAUS.ING THE DAMAGE. THE PIPING SYSTEM CONTAINING THE SNUBBER WAS INSPECTED. AN .

ADJACENT SUPP' ORT WAS FOUND WITH MINOR DAMAGE. ALL DISCREPANCIES WERE REPAIRED PRIOR TO PLANT STARTUP. LICENSED OPERATORS HAVE REVIEWED THIS REPORT AND HAVE BEEN CAUTIONED ABOUT WATER HAMMER EVENTS.

EVENT DATE 04/19/84 DOCKET:298 COOPER TYPE:BWR REGION: 4 NSSS:GE ARCHITECTURAL ENGINEER: BNR0 FACILITY OPERATOR: NEBRASKA PUBLIC POWER DISTRICT ABSTRACT:

POWER LEVEL - 070%. WHILE SCHEDULED CONSTRUCTION WORK WAS IN PROGRESS, A.BULLD0ZER INADVERTENTLY SHEARED A HYDRANT FROM THE FIRE PROTECTION SYSTEM WITHIN THE COOPER NUCLEAR STATION RESTRICTED SECURITY AREA.

THE STATION FIRE PUMPS AUTOMATICALLY STARTED, BUT WERE LATER TEMPORARILY SECURED WHILE THE HYDRANT WAS BEING ISOLATED FROM THE SYSTEM. AT THIS POINT, THE FIRE PROTECTION SYSTEM HEADER PRESSURE HAD DROPPED FROM 140 PSIG TO APPROX. 10 PSIG. AFTER THE HYDRANT WAS ISOLATED, THE FIRE PROTECTION SYSTEM WAS REPRESSURIZED BY USING THE ELECTRIC FIRE PUMP, STARTING 0F THE ELECTRIC FIRE PUMP CAUSED A PRESSURE SURGE WHICH RESULTED IN A SYSTEM WATER HAMMER. THIS WATER HAMMER FORCED OPEN THE CLAPPERS ON THE SBGTS AUTOMATIC DELUGE VALVES WHICH FLOODED THE CHARC0AL FILTERS ON THE SBGTS TRAINS, RENDERING BOTH TRAINS IN0PERABLE. THIS PLACED THE PLANT IN A TECH SPEC LC0 REQUIRING COLD SHUTDOWN. THE REACTOR WAS PLACED IN A COLD SHUTDOWN CONDITION UNTIL THE INOPERABLE SBGT TRAINS WERE MADE OPERABLE. LACK OF ATTENTION PAID BY THE BULLD0ZER OPERATOR TO HIS WORKING ENVIRONMENT, AND THE FAILURE OF THE CONTROL ROOM OPERATORS TO RESTOPE SYSTEM PRESSURE GRADUALLY, WERE PERSONNEL ERRORS WHICH WERE IDENTIFIED.

• w EVENT DATE 09/06/84 00CKET:325 BRUNSWICK 1 TYPE:BWR REGION: 2 NSSS:GE ARCHITECTURAL ENGINEER: UECX FACILITY OPERATOR: CAROLINA POWER & LIGHT CO.

ABSTRACT:

POWER LEVEL - 100%. ON 9-6-84, IT WAS DISCOVERED THAT THE COMMON SUPPORT EMBED PLATE OF UNIT 2 A CORE SPRAY SUBSYSTEM HYDRAULIC SNUBBERS 2-E21-2SS31 AND 2SS32 WAS LOOSE AND THE PLATE CONCRETE BASE WAS CRACKED AND BROKEN. SUBSEQUENT INSPECTIONS REVEALED THE CONDITION EXISTED ON THE SAME CORRESPONDING PIPING SUPPORT EMBED PLATE OF THE UNIT 1 A CORE SPRAY SUBSYSTEM. AT THE TIME OF THIS EVENT UNIT 2 WAS IN A REFUEL / MAINTENANCE OUTAGE. IN ADDITION, THE B CORE SPRAY .

SUBSYSTEM ON BOTH UNITS AND THE LPCI SUBSYSTEMS, THE HPCI SYSTEM, AND '

THE RCIC SYSTEM ON UNIT 1 WERE OPERABLE. AFTER PRELIMINARY PLANT ENGINEERING INVESTIGATION AND EVALUATION OF THIS EVENT, THE A CORE SPRAY SUBSYSTEM ON EACH UNIT WAS RESPECTIVELY DECLARED IN0PERABLE.

THE SUBJECT HYDRAULIC SNUBBER EMBED PLATE ON BOTH UNITS WAS STRENGTHENED WITH WING PLATES AND WEDGE ANCHORS. A WALKDOWN OF THE LOOP PIPING 0F BOTH CORE SPRAY SUBSYSTEMS ON EACH UNIT WAS CONDUCTED TO DETERMINE IF OTHER SUBSYSTEMS' PIPING SUPPORTS WERE AFFECTED. NO OTHER PROBLEMS WERE FOUND THAT AFFECT SYSTEM OPERABILITY.

EVENT DATE 09/08/84 00CKET:416 GRAND GULF 1 TYPE:BWR REGION: 2 NSSS:GE ARCHITECTURAL ENGINEER: BECH

_ FACILITY OPERATOR: MISSISSIPPI POWER & LIGHT CO.

ABSTRACT:

POWER LEVEL - 004%. DURING HANGER INSPECTIONS ON THE STANDBY ' SERVICE WATER SYSTEM, 12 PIPE SUPPORTS WERE FOUND DAMAGED ON PIPING TO AND FROM THE FUEL POOL COOLING HEAT EXCHANGERS. AN EVALUATION REVEALED THAT A WATER HAMMER TRANSIENT COULD OCCUR ON THIS HIGH, VERTICAL LENGTH OF PIPE. ON A LOP /LOCA SIGNAL, THE SSW PUMP DISCHARGE VALVE ,

OPENS PRIOR TO THE PUMP START ALLOWING A SLIGHT DRAIN DOWN WITH VOID FORMATIONS IN THE HIGHER ELEVATION PIPING. THUS. LOADS HIGHER THAN THE ORIGINAL DESIGN VALUES MAY OCCUR ON PUMP STARTS IN THIS SECTION OF PIPE. THE CONFIGURATION OF THE 'A' LOOP IS SUCH THAT THIS CONDITION WOULD NOT BE EXPERIENCED.

EVENT DATE 11/27/84 DOCKET:324 BRUNSWICK 2 TYPE:BWR REGION: 2 NSSS:GE ARCHITECTURAL ENGINEER: UECX FACILITY OPERATOR: CAROLINA POWER & LIGHT CO.

ABSTRACT:

POWER LEVEL - 000%. WHILE ATTEMPTING TO ESTABLISH SHUTDOWN COOLING ON THE A LOOP 0F RHR SYSTEM, 2 SEPARATE EVENTS OCCURRED: 1) A WATER HAMMER EVENT DAMAGING SUPPORTS ON THE STEAM CONDENSING LINE, AND 2) A RPS TRIP ON LOW VESSEL LEVEL. AT THE TIME OF THESE EVENTS, THE REACTOR WAS SHUT DOWN AT 90 PSI. FOLLOWING BOTH EVENTS, THE PLANT WAS RESTORED TO ITS NORMAL CONDITION. BOTH EVENTS OCCURRED WHEN THE CONTROL SIGNAL TO THE E11-F053A VALVE CAUSED THE VALVE TO TRAVEL TO ITS '0 PEN' POSITION INSTEAD OF THE DESIRED 'CLOSE' POSITION. THERE IS NO POSITION INDICATION OF THE F053A VALVE ON THE RTGB. AN -

INVESTIGATION DETERMINED THAT THE OUTPUT JACK WAS IN THE 'HIGH' POSITION INSTEAD OF THE ' LOW' POSITION. THIS MISPOSITIONING CAUSED THE F053 CONTROLLER TO SELECT THE HIGHER OF 2 SIGNALS (RHR HEAT EXCHANGER PRESSURE / LEVEL) INSTEAD OF THE LOWER. WHEN THE CONTROLLER WAS TURNED ON IN PREPARATION FOR WARMING UP THE RHR LINES PRIOR TO ESTABLISHING SHUTDOWN C0OLING, THE VALVE IMMEDIATELY WENT TO THE FULL OPEN POSITION. THE FIRST TIME THIS WAS ATTEMPTED THE WATER HAMMER OCCURRED; THE SECOND ATTEMPT RESULTED IN A LOW LEVEL SCRAM AND WATER HAMMER. VESSEL LEVEL WAS RESTORED TO THE NORMAL OPERATING BAND, THE OUTPUT JACK WAS CONNECTED TO ITS CORRECT POSITION, AND THE DAMAGED SUPPORTS WERE REPAIRED. THE CAUSE FOR THE SUBJECT OUTPUT JACK BEING IN THE WRONG POSITION COULD NOT BE DETERMINED.

1

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l EVENT DATE 05/09/83 DOCKET:298 COOPER TYPE:BWR REGION: 4 NSSS:GE .

ARCHITECTURAL ENGINEER: BNR0 FACILITY OPERATOR: NEBRASKA PUBLIC POWER DISTRICT REFERENCE LERS:

1 298/81-009 ABSTRACT:

DURING PERFORMANCE OF SP 6.3.10.9.1, FOUR RHR MECHANICAL (PSA-10)

SNUBBERS WERE FOUND TO BE MECHANICALLY FR0 ZEN IN PLACE. THIS IS IN '

VIOLATION OF TECH SPEC 3.6.H.1. THE REACTOR WAS IN COLD SHUTDOWN WHEN THIS EVENT WAS DISCOVERED. A SIMILAR FAILURE WAS REPORTED ON LER 81-09. REDUNDANT SYSTEMS WERE OPERABLE. THE FAILED SNUBBERS WERE PSA-10 MODEL MECHANICAL UNITS MANUFACTURED BY PA".IFIC SCIENTIFIC CORP.

INSTALLED IN SUPPORT RH-SS, 6, 16, & 71. A WATER HAMMER CAUSED BY RHR SHUTDOWN COOLING CAUSED AN OVERLOAD FAILURE OF THE SNUBBERS.

LARGER MODEL SNUBBERS WILL BE INSTALLED. THE RHR OPERATIONAL PROCEDURE WILL BE REVISED TO REDUCE WATER HAMMER.

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EVENT DATE 12/07/84 00CKET:325 BRUNSWICK 1 TYPE:BWR REGION: 2 NSSS:GE ARCHITECTURAL ENGINEER: UECX FACILITY OPERATOR: CAROLINA POWER & LIGHT CO.

ABSTRACT:

POWER LEVEL - 001%. ON 12-7-84, AT 1913 AND APPR0X 1 MIN LATER, A UNIT 1 AUTOMATIC REACTOR SCRAM AND RPS AUTOMATIC ACTUATION TRIP, RESPECTIVELY, OCCURRED DUE TO SPURIOUS INSTRUMENT UPSCALE SPIKES OF THE UNIT REACTOR POWER INTERMEDIATE RANGE MONITORS (IRMS) E AND H. ON 12-8-84, AT 0409, A UNIT 1 AUTOMATIC REACTOR SCRAM OCCURRED DUE TO SPURIOUS INSTRUMENT UPSCALE SPIKES ON IRMS E AND H. AT THE TIME OF THESE EVENTS, UNIT 1 WAS IN REACTOR STARTUP. DURING EACH UNIT SCRAM RECOVERY, NORMAL PLANT OPERATING PARAMETERS WERE MAINTAINED. THE ROOT '

CAUSE OF BOTH EVENTS IS ATTRIBUTED TO A DEFECTIVE INPUT SIGNAL CABLE TO IRM E AND INSUFFICIENT CONTACT TENSION ON THE SIGNAL LEAD CONNECTORS OF IRM H. THE FIRST EVENT WAS INITIATED BY WATER HAMMER IN THE UNIT MAIN STEAM LINES WHICH RESULTED IN INSTRUMENTATION OSCILLATIONS AND SIGNAL NOISE BEING INDUCED INTO THE REACTOR NUCLEAR INSTRUMENTATION. THE SECOND EVENT WAS INITIATED BY NOISE SIGNALS CAUSED BY ENERGIZING THE REACTOR NANUAL CONTROL SYSTEM. THE RESULT OF EACH INITIATING EVENT WAS INSTRUMENT UPSCALE SPIKES IN IRMS E AND H DUE TO PROBLEMS AFFECTING BOTH MONITORS. FOLLOWING THE SECOND EVENT WHEN THE ROOT CAUSE OF BOTH EVENTS WAS IDENTIFIED, APPROPRIATE REPAIRS WERE MADE TO THE SUBJECT IRMS AND THEY WERE RETURNED TO SERVICE.

l

EVENT DATE 03/31/85 00CKET:293 PILGRIM 1 TYPE:BWR REGION: 1 NSSS:GE ARCHITECTURAL ENGINEER: BECH FACILITY OPERATOR: BOSTON EDIS0N CO.

REPORTABILITY CODES FOR THIS LER ARE:

14 10 CFR 50.73(a)(2)(v): Event that could have prevented fulfillment of a safety function.

ABSTRACT:

POWER LEVEL - 100%. ON 3-31-85, WHILE PERFORMING A ROUTINE HPCI OPERABILITY FLOW TEST, THE HPCI TURBINE TRIPPED ON OVERSPEED.

SUBSEQUENTLY, A BLOWN RUPTURE DISC, BROKEN SNUBBER, AND 2 DEGRADED BASEPLATES WERE IDENTIFIED ON THE HPCI TURBINE EXHAUST LINE. .CAUSE OF THE TRIP WAS l die RESULT OF A FAULTY CONNECTOR IN THE HPCI TURBINE .

CONTROL SYSTEM. PROBABLE CAUSE OF THE BLOWN RUPTURE DISC, BROKEN SNUBBER AND DEGRADED BASEPLATES IS BELIEVED TO BE THE RESULT OF AN ANOMALOUS EVENT (I.E., WATERHAMMER). CORRECTIVE ACTION WAS TO REPLACE THE CONNECTOR, REPLACE THE RUPTURE DISC, AND REBUILD THE SNUBBER AND BASEPLATES. THE FAULTY CONNECTOR IS CONSIDERED AN ISOLATED EVENT. TO PRECLUDE RECURRENCE OF THE RUPTURE DISC, SNUBBER, AND BASEPLATE PROBLEM, THE DURATION AND FREQUENCY OF THE HPCI TURBINE EXHAUST LINE BLOWDOWN HAS BEEN INCREASED. FINAL CORRECTIVE ACTION IS PENDING ENGINEERING ANALYSIS OF ROOT CAUSE.

EVENT DATE 04/27/85 00CKET:388 SUSQUEHANNA 2 TYPE:BWR REGION: 1 NSSS:GE ARCHITECTURAL ENGINEER: BECH FACILITY OPERATOR: PENNSYLVANIA POWER & LIGHT C0.

ABSTRACT:

POWER LEVEL - 000%. ON 4-27-85 AT 0835 WHILE PREPARING TO PLACE THE

'B' LOOP 0F THE RHR SYSTEM IN SERVICE, A WATERHAMMER OCCURRED CAUSING A REACTOR WATER LEVEL TRANSIENT. THE REACTOR WATER LEVEL DROPPED APPROX 35 INCHES CAUSING A RPS ACTUATION. NO CONTROL R0D MOTION OCCURRED SINCE ALL RODS WERE FULLY INSERTED AT THE TIME OF THE EVENT.

NO EMERGENCY CORE COOLING SYSTEMS INITIATED. THE WATERHAMMER AND LEVEL TRANSIENT WERE CAUSED BY RAPIDLY FILLING PARTIALLY DRAINED DOWN RHR PIPING FROM THE REACTOR VESSEL. THE PIPING WAS PARTIALLY DRAINED WHILE WARMING.THE INJECTION LINE IN ACCORDANCE WITH THE OPERATING -

PROCEDURE. AFTER RESTORING THE REACTOR VESSEL LEVEL AND RESETTING THE RPS SIGNAL, THE 'D' RHR PUMP WAS STARTED IN SHUTDOWN COOLING. A SECOND WATERHAMMER OCCURRED AND THE SHUTDOWN COOLING SUCTION VALVES CLOSED DUE TO HIGH FLOW (AN ESF). THE SECOND WATERHAMMER RESULTED FROM RAPIDLY COLLAPSING STEAM P0CKETS IN THE RHR PIPING WHEN THE PUMP WAS STARTED. A WALKDOWN OF THE SYSTEM OUTSIDE OF THE CONTAINMENT WAS PERFORMED AND NO DAMAGE WAS FOUND. THE OPERATING PROCEDURE FOR RHR IS BEING REVISED TO DELETE THE SECTION FOR INJECTION LINE WARM UP TO PREVENT FUTURE WATERHAMMERS.

EVENT DATE l 05/18/85 DOCKET:293 PILGRIM 1 TYPE:BWR REGION: 1 NSSS:GE ARCHITECTURAL ENGINEER: BECH FACILITY OPERATOR: BOSTON EDISON C0.

REPORTABILITY CODES FOR THIS LER ARE:

14 10 CFR 50.73(a)(2)(v): Event that could have prevented fulfillment of a safety function.

REFERENCE LERS:

1 293/85-008 ABSTRACT:

POWER LEVEL - 100%. ON 5/18/85, WHILE PERFORMING A ROUTINE HPC.I OPERABILITY TEST) (REF.: PROCEDURE 8.5.4.1), THE HPCI TURBINE TRIPPED . .

FOLLOWED BY A* RAPID RESTART. SUBSEQUENTLY, A BROKEN UPPER SNUBBER, DISPLACED CONCRETE EXPANSION ANCHORS, AND A BENT PISTON R0D ON THE LOWER SNUBBER WERE IDENTIFIED ON THE HPCI TURBINE EXHAUST LINE. CAUSE OF THE TURBINE TRIP WAS MOST PROBABLY DUE TO TRANSIENTS DURING COLD QUICK STARTS. PROBABLE CAUSE OF THE BROKEN SNUBBER, THE CONCRETE ANCHORS BEING DISPLACED, AND BENT PISTON R0D IS BELIEVED TO BE THE RESULT OF WATER HAMMER. ON 5/23/85, WHILE PERFORMING THE HPCI OPERABILITY TEST (PROCEDURE 8.5.4.1), A HIGH FLOW ISOLATION WAS RECEIVED. PROBABLE CAUSE OF THE ISOLATION WAS DUE TO AIR BEING INDUCED INTO THE HIGH FLOW SWITCH DURING EXTENSIVE CALIBRATION FOR EQUIPMENT QUALIFICATION WORK PERFORMED ON 5/20/85, COMBINED WITH TRANSIENTS DURING A COLD QUICK START OF THE HPCI SYSTEM. SUBSEQUENTLY, THE ISOLATION WAS RESET AND THE HPCI SYSTEM WAS SUCCESSFULLY TESTED THREE TIMES ON 4/23/85 AT APPR0XIMATELY 1122 HRS. ON 6/6/85, THE HPCI TURBINE TRIPPED AND ISOLATED ON HIGH FLOW DURING SURVEILLANCE TESTING.

CAUSE WAS DUE TO IMPROPER INSTALLATION OF A TRANSDUCER ON THE HPCI CQNTROL VALVE OPERATING MECHANISM (RELAY PISTON). CORRECTIVE ACTION WAS TO REMOVE THE TRANSDUCER.

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ - - _ _ _ J

O 9 Y g

a POST 1981 WATER HAMMER OCCURRENCES IN PWRS en-I

EVENT DATE 12/09/82 00CKET:361 SAN ON0FRE 2 TYPE:PWR REGION: 5 NSSS:CE ARCHITECTURAL ENGINEER: BECH FACILITY OPERATOR: SOUTHERN CALIFORNIA EDISON CO.

ABSTRACT:

WHILE IN MODE 1, IT WAS OBSERVED THAT TWO SNUBBERS WHICH ARE COMPONENTS OF PIPE SUPPORTS S2-FW-189-H013 WERE DAMAGED AND DECLARED INOPERABLE IN ACCORDANCE WITH LIMITING CONDITION FOR OPERATION (LCO) 3.7.6. SUBSEQUENTLY, ON 12/22/82 WHILE THE UNIT WAS IN MODE 5, THREE ADDITIONAL SNUBBERS WERE OBSERVED TO BE DAMAGED AND DECLARED INOPERABLE. IT IS NOW BELIEVED THAT THE SNUBBERS MAY HAVE BEEN INITIALLY DAMAGED DURING STARTUP TESTING, WHICH INCLUDED A WATER HAMMER TRANSIENT IN MARC:;, 1981. THE ENGINEERING ANALYSIS REQUIRED BY TECH SPEC 4.7.6.G WAS PERFORMED WITH SATISFACTORY RESULTS. ALL .

DAMAGED SNUBB'ERS HAVE BEEN REPLACED.

EVENT DATE 01/25/83 DOCKET:309 MAINE YANKEE TYPE:PWR REGION: 1 NSSS:CE ARCHITECTURAL ENGINEER: SW FACILITY OPERATOR: MAINE YANKEE ATOMIC POWER CO. ,

COMMENTS STEP 2: MEI ARE STRESS RISERS AND ELB0WS. STEP 3: EFFECT DX- DAMAGE TO SG INTERNALS. PERSONNEL EXPOSED DURING REPAIR.

ABSTRACT:

~

AT 1432 HOURS JANUARY 25, 1983, THE MAINE YANKEE NUCLEAR PLANT TRIPPED FROM FULL LOAD WHILE ISOLATING AN ELECTRICAL GROUND. MAIN FEEDWATER FLOW WAS NOT AVAILABLE FOLLOWING THE TRIP SO STEAM GENERATOR LEVEL RESTORATION WAS ACCOMPLISHED, AS DESIGNED, BY THE AUTO START OPERATION OF THE AUXILIARY FEED PUMPS. APPROXIMATELY 15 MINUTES AFTER THE TRIP: A LOUD NOISE WAS HEARD IN THE PLANT MACHINE SHOP WHICH IS LOCATCD JUST BELOW THE MAIN FEED LINES; A CONTAINMENT FIRE DETECTOR l

ALARMED; AND CONTAINMENT HUMIDITY BEGAN TO RISE. THE CONTAINMENT WAS ENTERED FOR INSPECTION. THE FEED LINE WAS FOUND TO BE LEAKING SEVERELY NEAR THE #2 STEAM GENERATOR INLET N0ZZLE. STATION C00LDOWN WAS INITIATED TO PERMIT CLOSE ACCESS FOR INSPECTION AND TO EFFECT REPAIRS. VISUAL INSPECTIONS REVEALED A FEED LINE SIDE CRACK HAD

OCCURRED ADJACENT TO THE PIPE TO STEAM GENERATOR N0ZZLE WELD ON THE UPSTREAM EXHIBITED DEFORMATION OR OTHER DISTRESS. THE LEAK OCCURRED AS A RESULT OF A WATER HAMMER EVENT CAUSING THE ULTIMATE FAILURE OF WHAT WAS MOST PROBABLY AN EXISTING CRACK IN THE FEED PIPE. THE EVENT AND DAMAGE WAS TYPICAL OF THAT EXPERIENCED AT OTHER PWR'S AND AS EXTENSIVELY DESCRIBED IN NUREG'S AND OTHER INDUSTRY LITERATURE.

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' EVENT DATE 05/08/85 DOCKET: 382 WATERFORD 3 TYPE: PWR REGION: 4 NSSS: CE ARCHITECTURAL ENGINEER: EBASCO FACILITY OPERATOR: LOUISIANNA POWER & LIGHT CO.

ABSTRACT ON MAY 8, 1985, DURING A SURVEILLANCE TEST OF THE EMERGENCY FEEDWATER SYSTEM AT WATERFORD 3, AN EVENT OCCURRED THAT RESULTED IN DAMAGE TO STRUCTURES 7 STRUTS, 1 SNUBBER) SUPPORTING THE STEAM SUPPLY LINE (LATER REFERRED TO AS DRY PIPE) TO THE TURBINE DRIVEN EMERGENCY FEEDWATER PUMP. AT THE TIME OF THE TEST, AN EMPLOYEE OF THE LICENSEE, LOUISIANA POWER & LIGHT (LPL), AT THE TURBINE DRIVEN PUMP 90 FEET BELOW THIS PIPE REPORTED HAVING HEARD A LOUD NOISE FROM AB0VE. THE SYSTEM PASSED THE SURVEILLANCE TEST SATISFACTORILY AND LPL DID NOT INSPECT THE SYSTEM UNTIL THE NEXT DAY, DISCOVERING THE PIPE SUPPORT DAMAGE AT THAT TIME. AFTER FINDING THE DAMAGE, THE LICENSEE FOUND THATTWOHEATTRACINGCIRCUITSHADNOTBEENOPEgATINGPROPERLYTOKEEP THE DRY PIPE TEMPERATURE ABOVE THE REQUIRED 280 F. THE LICENSEE ALSO WAS -

ABLE TO DRAIN ABOUT A GALLON OF CONDENSATE FROM ONE OF THE STEAM TRAPS CONNECTED TO THE LINE, INDICATING THE POSSIBILITY OF WEEPAGE PAST THE MOTOR OPERATED VALVE ISOLATING THE DRY PART OF THE LINE FROM THE MAIN STEAM LINE.

THIS INFORMATION LEADS TO THE MOST LIKELY EXPLANATION OF THE EVENT AS WATER HAMMER FROM A STEAM DRIVEN WATER SLUG. THE LICENSEE HAS COMMISSIONED EBASCO TO DETERMINE THE CAUSE AND ANALYZE ALTERNATIVE CORRECTIVE ACTIONS TO PREVENT SIMILAR EVENTS.

EVENT DATE 08/13/85 00CKET:368 ARKANSAS NUCLEAR 2 TYPE:PWR REGION: 4 NSSS:CE ARCHITECTURAL ENGINEER: BECH FACILITY OPERATOR: ARKANSAS POWER AND LIGHT C0.

REFERENCE LERS:

1 368/85-015 2 368/85-016 ABSTRACT:

POWER LEVEL - 100%. ON 8/13/85 AT 1727 HOURS, A REACTOR TRIP OCCURRED WHILE OPERATING IN MODE 1 AT 100% FULL POWER (FP). THE CASING OF STEAM GENERATOR (SG) BLOWDOWN TANK DRAIN PUMP 2P-139A BROKE APART BECAUSE OF A WATER HAMMER. BLOWDOWN ISOLATION TO SECURE THE RESULTANT STEAM LEAK ALLOWED AIR TO ENTER THE CONDENSER FROM THE BLOWDOWN TANK VENT AND BROKEN PUMP CASING RESULTING IN A TURBINE TRIP ON LOW CONDENSER VACUUM. THE REACTOR TRIPPED ON HIGH PRESSURIZER PRESSURE BECAUSE OF THE TURBINE TRIP. POST TRIP PLANT RESPONSE WAS NORMAL.

EMERGENCY FEEDWATER (EFW) ACTUATED ON LOW SG LEVEL. WHEN THE "A" EFW TRAIN WAS BEING SECURED, CONTROL VALVE 2CV-1039 FAILED TO CLOSE.

REDUNDANT EQUIPMENT WAS OPERABLE. THE CLOSING COIL FOR 2CV-1C39 RAD SHORTED. INVESTIGATION REVEALED THAT THE CLOSING COIL WAS C0hNECTED TO A ' BREAK BEFORE MAKE' AUXILIARY SWITCH INSTEAD OF A 'MAKE BEFORE BREAK' AUXILIARY SWITCH. THIS CONFIGURATION ULTIMATELY LED TO C0Il FAILURE. THE CLOSING COIL FOR 2CV-1039 WAS REPLACED. THE WIRING DISCREPANCY WAS CORRECTED ON 2CV-1039 AND A SIMILAR EFW VALVE (2CV-1037) IN THE SAME EFW TRAIN. THE VALVES WERE TESTED, FOUND TO BE ACCEPTABLE, AND RETURNED TO SERVICE. PUMP 2P-139A WAS REPLACED.

INVESTIGATION REVEALED THAT SG BLOWDOWN TANK BYPASS VALVE 2SGS-22 WAS LEAKING BY THE SEAT ALLOWING HIGH ENERGY WATER TO BYPASS THE BLOWDOWN TANK PRESSURE REDUCTION CONTROL VALVE.

. EVENT DATE 10/04/85 DOCKET:361 SAN ON0FRE 2 TYPE:PWR REGION: 5 NSSS:CE ARCHITECTURAL ENGINEER: BECH FACILITY OPERATOR: SOUTHERN CALIFORNIA EDIS0N CO.

REFERENCE LERS:

1 362/85-029 2 361/84-079 ABSTRACT:

POWER LEVEL - 100%. ON 10-4-85 WITH UNIT 2 AT 100% POWER, 7 PACIFIC SCIENTIFIC SNUBBERS WERE FOUND FR0 ZEN ON THE SHUTDOWN COOLING SYSTEM.

THE UNIT 2 SNUBBERS WERE BEING SURVEILLED AS THE RESULT OF FR0 ZEN SNUBBERS BEING FOUND ON THE UNIT 3 SDCS, DURING THE 18 MONTH SNUBBER SURVEILLANCE TESTING CURRENTLY BEING PERFORMED. THE SNUBBER FAILURES ' ~

ARE ATTRIBUTED TO A WATER HAMMER WHICH OCCURRED WHEN, AFTER DRAINING AND REFILLING THE SYSTEM, AIR P0CKETS FORMED IN PORTIONS OF THE SDCS BECAUSE OF INADEQUATE VENTING CAPABILITY. AS CORRECTIVE ACTIONS, THE SNUBBERS WERE REPLACED AND THE PORTIONS OF THE SDCS, WHICH ARE POSTULATED TO HAVE CAUSED THE TRANSIENTS, WILL HAVE ADDITIONAL VENTING CAPABILITY ADDED. AN ENGINEERING ANALYSIS WAS PERFORMED WITHIN THE 72 HR ACTION STATEMENT REQUIREMENT OF TECH SPEC 3.7.6 WHICH DEMONSTRATED THE SYSTEM REMAINED FUNCTIONALLY OPERABLE AND CAPABLE OF PERFORMING ITS SAFETY FUNCTION.

EVENT DATE 05/04/81 DOCKET:206 SAN ONOFRE 1 TYPE:PWR REGION: 5 NSSS:WE ARCHITECTURAL ENGINEER: BECH FACILITY OPERATOR: SOUTHERN CALIFORNIA EDIS0N C0.

ABSTRACT:

IN ACCORDANCE WITH WESTINGHOUSE RADAR RESPONSE SCE-80-52, AN INSPECTION WAS PERFORMED ON THE THERMAL SLEEVE AND FEEDRING ASSEMBLY IN THE B STEAM GENERATOR AND THE FOLLOWING CONDITIONS WERE NOTED:

THERMAL SLEEVE DEFORMATION IN AREA 0F FEED N0ZZLE; RADIAL INDICATIONS AT SEVERAL FEEDRING DISTRIBUTION H0LES. LINEAR INDICATIONS ON SLEEVE TEE TO FEEDRING WELD; FEEDRING SUPPORT DEFORMATION AND LINEAR INDICATIONS AT "U" SUPPORT TO BRACKET WELD; DEBRIS IN FEEDRING.

SLEEVE AND SUPPORT DEFORMATION WERE CAUSED BY WATER HAMMER.

INDICATIONS AT DISTRIBUTION HOLES WERE ATTRIBUTED TO STRIPPING PHENOMENON. INDICATIONS AT SLEEVE TEE WERE ATTRIBUTED TO FLOW INDUCED VIBRATION. DEBRIS WAS FROM A DETERIORATED FLOW STRAIGHTENER.

CONDITIONS WERE CORRECTED ON 5/25/81 EXCEPT SLEEVE AND DISTRIBUTION HOLES. ENGINEERING ANALYSIS WAS PERFORMED TO JUSTIFY CONTINUED USE OF SLEEVE AND FEEDRING ASSEMBLY.

1 EVENT DATE 10/14/81 00CKET:247 INDIAN POINT 2 TYPE:PWR REGION: 1 NSSS:WE ARCHITECTURAL ENGINEER: UECX FACILITY OPERATOR: CONSOLIDATED EDISON C0.

REFERENCE LERS:

1 247/80-005 ABSTRACT:

WHILE IN HOT SHUTDOWN, A PERTURBATION IN THE SECONDARY BLOWDOWN PIPING ASSOCIATED WITH NO. 21 STEAM GENERATOR RESULTED IN FAILURE OF THE BOLTS BETWEEN THE STRUT ASSEMBLY AND THE BASE OF SNUBBER N0. 46-SR-3, A BERGEN PATTERSON HYDRAULIC SNUBBER HSSA-3. SIMILAR EVENT LER 80-005. THE BOLTS BETWEEN THE STRUT ASSEMBLY AND THE SNUBBER FAILED APPARENTLY DUE TO WATER HAMMER. THE BOLTS AND THE SNUBBER WERE '

REPLACED. VISUAL INSPECTION OF BLOWDOWN PIPING HAS BEEN ADDED T0 REACTOR CONTAINMENT BUILDING INSPECTION FOR LEAKAGE & ANOMALOUS CONDITIONS.

EVENT DATE 05/29/82 DOCKET:250 TURKEY POINT 3 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: BECH FACILITY OPERATOR: FLORIDA POWER & LIGHT C0.

hBSTRACT:

WHILE PERFORMING A LEAK INSPECTION INSIDE UNIT 3 CONTAINMENT, SEVERAL BROKEN HANGERS AND A BROKEN SNUBBER (#37) WERE DISCOVERED ON 3C STEAM GENERATOR BLOWDOWN LINES. CLOSER INSPECTION FOUND EVIDENCE OF APPR0XIMATELY 12 INCHES OF LINE MOVEMENT. THIS IS REPORTABLE IN ACCORDANCE WITH TECH SPEC 3.13.1. THE CAUSE WAS DETERMINED TO BE HYDRAULIC SH0CK OF THE BLOWDOWN LINES WHEN THE BLOWDOWN ISOLATION VALVE IS OPENED. THIS NEW BLOWDOWN SYSTEM WAS INSTALLED DURING THE RECENT STEAM GENERATOR REPAIR OUTAGE. SPECIAL INSTRUCTIONS FOR BLOWDOWN SYSTEM OPERATIOi NAVE BEEN ISSUE 3. B) FOLLOWING THE SEQUENCE FOR VALVE LINEUP AND OPERATION, THE HYDRAULIC SH0CKING IS AVOIDED.  :

EVENT DATE 09/06/82 DOCKET:261 ROBINSON 2 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: EBAS FACILITY OPERATOR: CAROLINA POWER & LIGHT C0.

ABSTRACT:

VALVE CVC-203 (LETDOWN RELIEF VALVE) LIFTED AND WOULD NOT RESEAT. THE UNIT WAS PLACED IN HOT SHUTDOWN CONDITIONS AT 0413 HOURS ON SEPTEMBER 7, 1982 TO ISOLATE CVC-203 FOR REPAIRS. THE RESULTING LEAKAGE WAS DETERMINED TO BE APPROXIMATELY 9 GPM WHICH IS IN EXCESS OF THE ALLOWABLE LEAKAGE DEFINED BY TECH SPEC 3.1.5.1 AND IS REPORTED PURSUANT TO 6.9.2.B.2. THE LEAKAGE WAS CONFINED TO THE PRESSURIZER RELIEF TANK AND CONTAINMENT. THE FAILURE OF CVC-203 WAS THE RESULT OF WATER HAMMER WHICH WAS CAUSED BY OPERATOR ERROR IN THE REALIGNMENT OF A CHARGING PUMP FROM RECIRCULATION TO NORMAL OPERATION AND RESULTED .

IN LOSS OF COOLING TO THE LETDOWN FLUID. CVC-203 WAS REPAIRED AND THE PERSONNEL INVOLVED IN THE EVENT WERE COUNSELLED. THE RELIABILITY OF CVC-203 WAS UNDER INVESTIGATION PRIOR TO THIS EVENT, AND EFFORTS TOWARD RESOLUTION ARE CONTINUING.

EVENT DATE 10/07/82 00CKET:275 DIABLO CANYON 1 TYPE:PWR REGION: 5 NSSS:WE ARCHITECTURAL ENGINEER: PGEC FACILITY OPERATOR: PACIFIC GAS & ELECTRIC C0.

ABSTRACT:

TESTING PERFORMED ON THE AUXILIARY SALTWATER (ASW) SYSTEM HAS REVEALED THAT THE SYSTEM IS SUSCEPTIBLE TO WATER HAMMER EFFECTS DURING ANTICIPATED OPERATIONAL TRANSIENTS. THESE TRANSIENTS INCLUDE PUMP TRIP AND RESTART SEQUENCES SUCH AS WOULD OCCUR FOLLOWING A LOSS OF i 0FFSITE POWER. THE PEAK PRESSURE OBSERVED DURING THIS TESTING EXCEEDED THE 100 PSIG SYSTEM DESIGN PRESSURE SPECIFIED IN THE FSAR.

( THE CAUSE OF THE SYSTEM WATERHAMMER IS BELIEVED TO BE WATER COLUMN l SEFARATION AND SUBSEQUENT COLUMN RECOMBINATION AT A POINT OF

! SIGNIFICANT PIPING SLOPE CHANGE. FURTHER EVALUATION OF THE EVENT AND l ASW SYSTEM DESIGN IS BEING CONDUCTED. RESULTS OF THE EVALUATION WILL ,

BE REPORTED IN A REVISION TO THIS LER.

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EVENT DATE 03/27/83 DOCKET:369 MCGUIRE 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: DUKE FACILITY OPERATOR: DUKE POWER CO.

ABSTRACT:

DURING UNRELATED MAINTENANCE ACTIVITY, HANGER IMCA-ND-H260 WAS DISCOVERED TO HAVE PULLED LOOSE FROM ITS WALL ANCHORS. ALL RESIDUAL HEAT REMOVAL (ND) SYSTEM MECHANICAL SNUBBERS WERE SUBSEQUENTLY FULL-STR0KED (ON 3/27/83) TO ENSURE THEIR OPERABILITY, REVEALING 2 SNUBBERS (1MCA-ND-H177 AND H317) TO BE LOCKED UP. THIS VIOLATES TECH SPEC 3.7.8 WHICH IS REPORTABLE PER TECH SPEC 6.9.1.13(B). ALTHOUGH DISCOVERED AFTER CORE WAS UNLOADED, FAILURES PROBABLY OCCURRED IN MODE

4. THE SYSTEM WAS NOT SEISMICALLY CHALLENGED AND CONTINUED T0-FUNCTION. IT WAS CONCLUDED THAT A WATER HAMMER HAD OCCURRED (0VERLOADING THE HANGER / SNUBBERS) DURING MODE 4, WHEN REACTOR COOLANT IS ALIGNED TO THE ND SYSTEM. VOIDS FORMED BY FLASHING OF ISOLATED VOLUME OF WATER BETWEEN VALVES ND-1 AND ND-2 DURING LEAK TESTING CREATED WATER HAMMER. THE HANGER ANCHOR BOLTS WERE RET 0RQUED, A BRACE ADDED, AND THE SNUBBERS REPLACE

D. PROCEDURE

S WILL BE REVISED TO PREVENT FUTURE WATER HAMMERS.

EVENT DATE 04/07/83 DOCKET:369 MCGUIRE 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: DUKE FACILITY OPERATOR: DUKE POWER CO.

REFERENCE LERS:

-- 1 369/83-016 l

ABSTRACT:

! WHILE IN MODES 5 AND 6, FOLLOWING THE IDENTIFICATION OF IN0PERABLE l

RESIDUAL HEAT REMOVAL SYSTEM (ND) HANGERS AND SNUBBERS (REF, 50-369/83-16), A SYSTEM WALKDOWN REVEALED ADDITIONAL HANGER FAILURES. '

THESE FAILURES WCRE DISCOVERED FROM 4/7/83 THROUGH 4/21/83, INVOLVING HANGERS IMCA-ND-H273, H287, H7, H21, H274, H282, H306 AND H308. THIS VIOLATES TECH SPEC 3.7.8 WHICH IS REPORTABLE PER TECH SPEC 6.9.1.13(B). THE ND SYSTEM WAS NOT SEISMICALLY CHALLENGED DURIN3 THIS PERIOD AND CONTINUED TO PROVIDE NORMAL COOLING. THE HANGER FAILURES l ARE ATTRIBUTED TO ND SYSTEM WATER HAMMERS WHICH RESULTED FROM VARIOUS CAUSES. THE VULNERABILITY OF THE ND SYSTEM TO EXPERIENCE WATER HAMMERS IS DUE TO DESIGN DEFICIENCY. THE HANGERS AND ND SYSTEM TRAIN

'A' WERE DECLARED IN0PERABLE AS NECESSARY. THE HANGERS WERE REPAIRED, INSPECTIONS PERFORMED, AND PROCEDURES CHANGED TO PREVENT FUTURE WATER HAMMERS.

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EVENT DATE 10/06/83 DOCKET:311 SALEM 2 TYPE:PWR REGION: 1 NSSS:WE ARCHITECTURAL ENGINEER: PSEG FACILITY OPERATOR: PUBLIC SERVICE ELECTRIC & GAS C0.

REFERENCE LERS:

1 311/83-056 ABSTRACT:

ON OCTOBER 6, 1983, DURING ROUTINE POWER OPERATION, A RUN TO PROVE OPERABILITY WAS PERFORMED ON N0. 23 AUXILIARY FEED WATER PUMP. UPON STARTING A NOISE WAS HEARD AND THE TURBINE TRIPPED. THE PUMP WAS DECLARED IN0PERABLE AND TECH SPEC ACTION STATEMENT 3.7.1.2A WAS ENTERED. BOTH ELECTRIC DRIVEN REDUNDANT AUXILIARY FEEDWATER PUMPS WERE OPERABLE DURING THE OCCURRENCE; THE IN0PERABLE PUMP WAS RESTORED TO OPERABILITY WITHIN 72 HOURS OF THE LAST SATISFACTORY SURVEILLANCE.

THE EVENT CONSTITUTED OPERATION IN A DEGRADED MODE PERMITTED BY A LIMITING CONDITION FOR OPERATION AND IS THEREFORE REPORTABLE IN ACCORDANCE WITH TECH SPEC 6.9.1.9B. THE TURBINE STEAM SUPPLY LINE WARMUP DRAIN VALVES HAD BEEN ALMOST FULLY CLOSED DURING THE SURVEILLANCE RUN ON OCTOBER 5, 1983. UPON STARTING 0F 23 AFW PUMP, CONDENSATE ENTERED THE TURBINE RESULTING IN WATER HAMMER AND A TURBINE TRIP. THE VALVES WERE OPENED AND A SUCCESSFUL TEST RUN WAS PERFORMED. A PROCEDURE CHANGE WILL BE ISSUED TO ENSURE THAT THESE VALVES ARE CHECKED OPEN FOLLOWING OPERATION.

EVENT DATE 12/05/83 DOCKET:395 SUMMER 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: GLBT FACILITY OPERATOR: SOUTH CAROLINA ELECTRIC & GAS C0.

ABSTRACT:

ON DEC. 5 AND 9, 1983, V:TH THE PLANT IN MODE 5, QUALITY CONTROL PERSONNEL WERE CONDUCTItG A HANGER / SNUBBER INSPECTION OF THE FEEDWATER AND THE STEAM GENERATOR ALOWDOWN SYSTEMS. CUT OF THE TOTAL

INSPECTED, ONE FR0ZEH SN9BSER AND SEVERAL ROTATED PIPE CLAMPS WERE

IDENTIFIED. AN ENGINEERINC EVALUATION WAS PERFORMED PER TECH SPEC 3.7.7, WHICH REVEALED NO ADVERSE IMPACT ON THE SYSTEMS' OPERABILITY.

THE FEELWATER SYSTEM PREWARMING PIPING HAS BEEN EXPERIENCING SOME WATER HAMMER PROBLEMS DURING STARTUP PRIOR TO TRANSFERRING TO THE MAIN FEEDWATER PUMPS. SYSTEM WALKDOWNS REVEALED THE DISCREPANCIES AS NOTED AB0VE. THE DISCREPANCIES WERE EVALUATED BY ENGINEERING AND RETURNED l

TO OPERABLE STATUS UPON SUCCESSFUL COMPLETION OF APPROPRIATE MAINTENANCE AND SURVEILLANCE TESTING. A PROCEDURE CHANGE WAS GENERATED TO ALTER THE VALVE LINE-UP SEQUENCE TO ELIMINATE THE POTENTIAL FOR WATER HAMMER IN THE PREWARMING PIPING DURING STARTUP.

EVENT DATE 04/22/84 DOCKET: 318 CALVERT CLIFFS 2 TYPE: PWR REGION: 1 NSSS: CE ARCHITECTURAL ENGINEER: BECHTEL FACILITY OPERATOR: BALTIM0RE GAS & ELECTRIC ABSTRACT:

ONAPRIL22,19g4,THECALVERTCLIFFSUNIT2PLANTWASOPERATINGINMODE3 (1400 PSIG, 406 F). THE OPERATORS WERE IN THE PROCESS OF REALIGNING THE MAIN FEEDWATER SYSTEM FOLLOWING THE TESTING 0F A NEWLY INSTALLED MOTOR DRIVEN AUXILIARY FEEDWATER PUMP. DURING THE AUXILIARY FEEDWATER PUMP TEST, SG LEVEL HAD BEEN LOWERED BELOW THE FEEDRING, MAIN FEEDWATER FLOW HAD BEEN SECURED, AND THE MAIN FEEDWATER ISOLATION VALVE WAS SHUT DUE TO LEAKAGE PAST THE FEEDWATER REGULATING VALVE. APPR0XIMATELY THIRTY MINUTES LATER, THE MAIN FEEDWATER ISOLATION VALVE WAS RE0PENED TO FEED THE SG USING THE MAIN FEEDWATER SYSTEM.

UPON OPENING THE MAIN FEEDWATER ISOLATION VALVE A SEVERE WATER HAMMER OCCURRED.

THE WATER HAMMER DAMAGED A MAIN FEEDWATER ISOLATION VALVE, A MAIN FEEDWATER REGULATING VALVE AND A MAIN FEEDWATER REGULATING BYPASS VALVE'. -

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9

EVENT DATE 04/28/84 1

00CKET:311 SALEM 2 TYPE:PWR  ;

REGION: 1 NSSS:WE ARCHITECTURAL ENGINEER: PSEG FACILITY OPERATOR: PUBLIC SERVICE ELECTRIC & GAS CO.

REFERENCE LERS:

1 311/84-010 ABSTRACT:

POWER LEVEL - 006%. ON APRIL 28, 1984, WITH REACTOR POWER LEVEL AT SIX PERCENT AND THE TURBINE NOT LATCHED, TESTING WAS BEING PERFORMED ON NO. 23 STEAM GENERATOR WATER LEVEL CONTROL SYSTEM. THIS TESTING WAS THE RESULT OF TWO REACTOR TRIPS WHICH OCCURRED DUE TO HIGH-HIGH LEVEL IN N0. 23 STEAM GENERATOR. THE EVENTS SURROUNDING THOSE REACT'OR TRIPS '

ARE DOCUMENTED IN LER 84-010-00. TEST RESULTS REVEALED THAT N0. 23 STEAM GENERATOR FEEDWATER FLOW INDICATION CHANNELS WERE NOT RESPONDING. BOTH CHANNELS WERE DECLARED IN0PERABLE AND TECH SPEC LIMITING CONDITION FOR OPERATION 3.0.3 WAS ENTERED. IN ACCORDANCE WITH THE ACTION REQUIREMENTS, A UNIT SHUTDOWN WAS PERFORMED WITHIN ONE HOUR. RADIOGRAPHY OF NO. 23 STEAM GENERATOR FEEDWATER FLOW N0ZZLE REVEALED THAT THE N0ZZLE HAD MOVED APPR0XIMATELY TWENTY-FOUR INCHES FROM ITS DESIGNED LOCATION; APPARENTLY AS A RESULT OF A PREVIOUS FEEDWATER HAMMER EVENT. N0. 23 FEEDWATER FLOW N0ZZLES WAS REPLACED, AND THE FEED FLOW TRANSMITTERS WERE CALIBRATED. THE STEAM GENERATOR FEEDWATER LEVEL CONTROL SYSTEM FUNCTIONED AS DESIGNED DURING THE SUBSEQUENT STARTUP ON MAY 5, 1984. DUE TO A UNIT SHUTDOWN, WHICH IS REQUIRED BY THE TECH SPECS, THE EVENT IS REPORTABLE IN ACCORDANCE WITH 10CFR 50.73(A)(2)(I)(A).

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EVENT DATE 08/05/84 DOCKET:370 MCGUIRE 2 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: DUKE FACILITY OPERATOR: DUKE POWER C0.

REPORTABILITY CODES FOR THIS LER ARE:

14 10 CFR 50.73(a)(2)(v): Event that could have prevented fulfillment of a safety function.

ABSTRACT:

POWER LEVEL - 000%. ON 8-5-84, MCGUIRE UNIT 2 OPERATORS DISCOVERED A BROKEN WELD ON THE RESIDUAL HEAT REMOVAL (ND) SYSTEM LETDOWN LINE TO THE CHEMICAL AND VOLUME CONTROL SYSTEM (NV). THE ND SYSTEM WAS IN SERVICE AT THE TIME, AND WATER WAS SPRAYING FROM THE BROKEN PIPE AND '

FROM THE STEM OF A VALVE (2NV-121) IN THE NV SYSTEM. AN ESTIMATED 3000 T0 7000 GALLONS OF CONTAMINATED WATER WAS CONTAINED IN THE HEAT EXCHANGER ROOM, THE NC AND CONTAINMENT SPRAY SUMP, AND THE B FLOOR DRAIN SUMP AND TANK. UPON DISCOVERY THE LEAKING LINE WAS ISOLATED. A SUBSEQUENT INSPECTION REVEALED A NUMBER OF SUPPORT / RESTRAINTS (S/R'S)

DAMAGED, AND THE BROKEN SOCKET WELD COMPLETELY SEPARATED. CAUSES OF THIS EVENT ARE ATTRIBUTED TO COMPONENT MALFUNCTION / FAILURE, DUE TO LOOSE PACKING ON 2NV-121, AND AN UNUSUAL SERVICE CONDITION, BECAUSE A GAS VOID IN THE LINE RESULTED IN A WATER HAMMER. THE UNIT WAS IN MODE 5, COLD SHUTDOWN, AT THE TIME. DAMAGED PIPING AND SUPPORT / RESTRAINTS HAVE BEEN REPLACED. OTHER CORRECTIVE ACTIONS PLANNED AND/0R COMPLETED INCLUDE TESTING 0F WELDS FOR CRACKS, MEASUREMENT OF VIBRATION TO STUDY FATIGUE DAMAGE, AND OTHER EVALUATIONS AND ACTIONS.

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EVENT DATE 02/17/85 00CKET:275 DIABLO CANYON 1 TYPE:PWR REGION: 5 NSSS:WE ARCHITECTURAL ENGINEER: PGEC FACILITY OPERATOR: PACIFIC GAS & ELECTRIC C0.

ABSTRACT:

POWER LEVEL - 048%. AT 0415 PST, 2-17-85, WITH UNIT 1 IN MODE 1, PLANT OPERATORS INITIATED A MANUAL REACTOR TRIP IN RESPONSE TO THE AUTOMATIC SHUTDOWN OF MAIN FEEDWATER PUMPS 1-1 AND 1-2 FROM EXCESSIVE THRUST BEARING WEAR INDICATION. IN ACCORDANCE WITH PROCEDURES, THE PRIMARY PLANT WAS STABILIZED IN MODE 3. THE THRUST BEARING WEAR INDICATORS WERE DETERMINED TO BE CONSERVATIVELY SET TO ALLOW FOR READJUSTMENT AS OPERATIONAL AND TRANSIENT DATA BECOME AVAILABLE. NO DAMAGE WAS OBSERVED. TO PREVENT RECURRENCE, THE THRUST BEARING WEAR DETECTION SYS. TEM HAS BEEN TEMPORARILY RESET TO NEW VALUES RECOMMENDED BY THE VENDOR. SUBSEQUENT TO THIS REACTOR TRIP, SEVERAL EVENTS TOOK PLACE IN THE BALANCE OF PLANT. THE MOST IMPORTANT WAS A WATER HAMMER EVENT IN ONE OF THE MAIN FEEDWATER BYPASS LINES WHICH IS NOT COVERED BY THE ASME CODE SECTION XI. A DETAILED ACCOUNT OF ALL THE EVENTS WILL BE REPORTED IN A SUPPLEMENT TO THIS LER.

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EVENT DATE 04/17/85 DOCKET:311 SALEM 2 TYPE:PWR REGION: 1 NSSS:WE ARCHITECTURAL ENGINEER: PSEG FACILITY OPERATOR: PUBLIC SERVICE ELECTRIC & GAS C0.

ABSTRACT:

POWER LEVEL - 018%. ON APRIL 17, 1985, DURING UNIT STARTUP OPERATIONS, A " WATER HAMMER" NOISE WAS HEARD IN THE REHEAT STEAM LINE ASSOCIATED WITH NO. 21 STEAM GENERATOR FEED PUMP. AN EQUIPMENT OPERATOR OPENED A STEAM TRAP DRAIN VALVE, A SOLID STREAM 0F WATER ISSUED FROM THE TRAP.

WHILE IN THE PROCESS OF DRAINING WATER FROM THE REHEAT STEAM LINE, NO. 21 STEAM GENERATOR FEED PUMP SPEED AND DISCHARGE PRESSURE DECREASED SHARPLY. THIS WAS FOLLOWED BY DECREASING STEAM GENERATOR WATER LEVELS. THE AUXILIARY FEEDWATER PUMPS WERE STARTED, AND.A UNIT LOAD REDUCTION WAS INITIATED. THE WATER LEVEL IN NO. 24 STEAM , ,

GENERATOR REA'CHED LOW-LOW LEVEL SETP0 INT (8%), WHICH RESULTED IN A REACTOR / TURBINE TRIP. THE CAUSE OF THE REACTOR TRIP WAS THE INABILITY OF THE STEAM TRAPS TO ADEQUATELY REMOVE THE CONDENSATE WHICH HAD COLLECTED IN NO. 21 STEAM GENERATOR FEED PUMP REHEAT STEAM SUPPLY LINE. THE ROOT CAUSE WAS ATTRIBUTED TO CRUD AND CORROSION PRODUCTS (NORMALLY FOUND IN STEAM DRAIN PIPING), WHICH COLLECTED IN THE REHEAT STEAM DRAIN LINE PIPING FOLLOWING THE EXTENDED SHUTDOWN FOR REFUELING.

BECAUSE OF THE AUTOMATIC ACTUATION OF THE REACTOR PROTECTION SYSTEM, THIS EVENT IS REPORTABLE IN ACCORDANCE WITH 10CFR 50.73(A)(2)(IV).

EVENT DATE 04/29/85 DOCKET:370 MCGUIRE 2 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: DUKE FACILITY OPERATOR: DUKE POWER C0.

( ABSTRACT:

POWER LEVEL - 000%. ON 4-29-85, A WATER HAMMER OCCURRED ON MAIN STEAM LINE 2C. WATER HAD ACCUMULATED IN THE LINE BECAUSE THE MAIN STEAM l

LINE DRAIN VALVES WERE CLOSED. THE WATER HAMMER OCCURRED AFTER THE MAIN STEAM ISOLATION BfPASS VALVES WERE OPENED, CAUSING THE VOLUME OF XATER TO BE RELEASED D0WN THE STEAM LINE. BEFORE THE WATER HAMMER, AN ATTEMPT HAD BEEN MADE TO OPEN THE MAIN STEAM LINE DRAIN VALVES, BUT r

THEY WOULD NOT OPEN. GAGS (MECHANISM USED TO PHYSICALLY HOLD VALVES f

IN POSITION) HAD BEEN INSTALLED ON THE VALVES, PREVENTING THEM FROM OPENING. THE GAGS WERE INADVERTENTLY LEFT ON THE VALVES BECAUSE 2 DIFFERENT TYPES OF GAGS WERE USED ON THE VALVES, AND ONLY 1 TYPE WAS l REMOVED. THE UNIT WAS IN MODE 4 AT THE TIME OF THE WATER HAMMER.

THIS EVENT IS ATTRIBUTED TO AN ADMINISTRATIVE DEFICIENCY BECAUSE N0 GUIDANCE OR PROCEDURE IS AVAILABLE TO PREVENT USING 2 TYPES OF GAGS ON THE SAME VALVE.

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