ML20203N491
| ML20203N491 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 09/29/1986 |
| From: | George Thomas PUBLIC SERVICE CO. OF NEW HAMPSHIRE |
| To: | Noonan V Office of Nuclear Reactor Regulation |
| References | |
| IEIN-86-060, IEIN-86-60, SBN-1207, NUDOCS 8610090151 | |
| Download: ML20203N491 (200) | |
Text
,
U George S. Thomas Vice President-Nuclear Production Pulse Service of New Hampshire N:w Hampshire Yankee Division September 29, 1986 SBN-1207 T.F. B7.1.2 United States Nuclear Regulatory Commission Washington, DC 20555 i
Attention:
Mr. Vincent S. Noonan, Project Director PWR Project Directorate No. 5
References:
(a) Construction Permits CPPR-135 and CPPR-136, Docket Nos. 50-443 and 50-444 (b) PSNH Letter (SBN-ll67), dated July 21, 1986, "Seabrook Station Probabilistic Safety Assessment Update" J. DeVincentis to V. S. Noonan Subj ect: Seabrook Station Probabilistic Safety Assessment Update; Additional Information
Dear Sir:
During recent discussions with the NRC Staf f, NHY comt9 tted to provide 1
additional information concerning the SSPSA material submitted in Reference (b). Pursuant to this_ commitment, enclosed please find the additional information that was requested.
Should you have any questions concerning this information please contact Mr. David A. Maidrand at (603) 474-9574, extension 2354.
Very truly yours, or S.
homas Enclosures cc: Atomic Safety and Licensing Board Service List hk p
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l P.O. Box 300. Seabrook, NH 03874. Telephone (603) 474-9574
Dicas Curren, Esquire Patsr J. Mathawa, Mayor Harmon & Welco City Hell 2001 S. Street, N.W.
Newburyport, MA 01950 Suite 430 Washington, D.C.
20009 Judith H. Mizner Silvergate, Gertner, Baker, Sherwin E. Turk, Esq.-
Fine, Good & Mizner Office of the Executive Legal Director 88 Broad Street U.S. Nuclear Regulatory Commission Boston, MA 02110 Tenth Floor
^
Calvin A. Canney Washington, DC 20555 City thnager Robert A. Backus, Esquire City Hall t
116 Lowell Street 126 Daniel Street P.O. Box 516 Portsmouth, NH 03801 Manchester, NH 03105 Stephen E. Merrill, Esquire Philip Ahrens, Esquire Attorney General Assistant Attorney General George Dana Bisbee, Esquire Department of The Attorney General Assistant Attorney General Statehouse Station #6 Office of the Attorney General Augusta, ME 04333 25 Capitol Street i
Concord, NH 03301-6397 i
Mrs. Sandra Gavutis Chairman, Board of Selectmen Mr. J. P. Nadeau RFD 1 - Box 1154 Selectmen's Office Kennsington, NH 03827 10 Central Road Rye, NH 03870 Carol S. Sneider, Esquire Assistant Attorney General Mr. Angie Machiros Department of the Attorney General Chairman of the Board of Selectmen l
One Ashburton Place,19th Floor Town of Newbury Boston, MA 02108 Newbury, MA 01950 Senator Gordon J. Humphrey Mr. William S. Lord U.S. Senate Board of Selectmen l
Washington, DC 20510 Town Hall - Friend Street g
(ATTN: Tom Burack)
Amesbury, MA 01913 4
Richa,rd A. Hampe, Eng.
Senator Gordon J. Humphrey Hampe and McNicholas 1 Pillsbury Street 35 Pleasant Street Concord, NH 03301 J
Concord, NH 03301 (ATIN: Herb Boynton)
Thomas F. Powers, III H. Joseph Flynn, Esquire Town Manager Office of General Counsel Federal Emergency Management Agency Town of Exeter 10 Front Street 500 C Street, SW Exeter, NH 03833 Washington, DC 20472 Brentwood Board of Selectmen Paul McEachern, Esquire RFD Dalton Road Matthew T. Brock, Esquire Brentwood, NH 03833 Shaines & McEachern 25 Maplewood Avenue j
Cary W. Holmes, Esq.
P.O. Box 360 Holmes & Ells Portsmouth, NH 03801 47 Winnacunnet Road Hampton, NH 03842 Robert Carrigg Town Office Mr. Ed Thomas Atlantic Avenue FEMA Region I North Hampton, NH 03862 442 John W. McCormack PO & Courthouse Boston, MA 02109
ENCLOSURE 1 TO SBN-1207
SUMMARY
LISTING OF ADDITIONAL INFORMATION Enclosure No. 2 Check Valve Failure Rates Used in the Seabrook EPZ Study Enclosure No. 3 Frequency of Unfilled RHR Piping Enclosure No. 4 Seabrook Station Risk Management and Emergency Planning Study, PLG-0432, December 1983 Enclosure No. 5 RHR Piping Failure Probability Enclosure No. 6 RRR Pressure Boundary Failure Modes Enclosure No. 7 RHR Rx Degradation Enclosure No. 8 CAP / COP System Containment Isolation Valves Enclosure No. 9 IE Information Notice No. 86-60
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l ENCLOSURE NO. 2 TO SBN-1207 CHECK VALVE FAILURE RATES USED IN THE SEABROOK EPZ STUDY i
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a CHECK VALVE FAILURE RATES USED IN THE SE ABROOK EPZ ' STUDY 1
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1 Enclosed is a more detailed description of the development of check valve l
. f ailure rates used in the Seabrook EPZ study.
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P CHECK VALVE FAILURE RATES USED IN THE SEABROOK EPZ STUDY The following provides further documentation for the estimated failure rates of the check valves in the pressure boundary of the ECCS and RCS, 4
as modeled in.the Seabrook EPZ study (Reference 1). The failure modes of
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concern are disc rupture or gross leakage of a check valve that is initially seated and tested to verify its p"osition and failure of check valves to reseat on demand.
1.
DISC RUPTURE / GROSS LEAKAGE 1.1 DATA COLLECTION The source of event descriptions for this analysis was Nuclear Power Experience (Reference 2).
In this search, which was done manually based l
on the NPE key word index, a total of 610 check valve failure events were identified (Attachment A is an expanded list of these failure events, based on the NPE-automated retrieval system--a total of 692 events). The initial list was then reviewed to identify leakage events (external and internal) in all systems of both PWRs and BWRs. A total of 163 events were identified. These events are marked "L" in Attachment B.
This review also provided further evidence that a large' number of check valve leakage events should not be considered for the failure mode of interest 1
in the V-sequence analysis, either in terms of mode or cause of fai. lure.
It was then decided to limit the data base to those events involving check valve leakage in the ECCS and RCS/ECCS system boundary of PWRs.
These were judged to be the closest category to the initially seated and tested check valves modeled in the analysis. No disc rupture events were identified in these events, and the maximum observed leak rate was 200 gpm. The majority of events involved very small leaks. Those considered to be more significant are listed in Table 1.
Even for the cases listed in Table 1, the exact leak rates were not always provided in the event reports. Consequently, leak rates were estimated, based on
-other available evidence:
the rate of boron concentration change, pressure reduction, and similarity to other occurrences for which the leak rates were known.
Also considered was an event that occurred in San Onofre Unit 1 in November 1985. The event involved failure of several main feedwater pump discharge check valves to reseat on demand resulting in overpressurization of the main feedwater system. A sumery of the event, as presented in Nuclear Power Experience (Reference 2), is provided as Attachment C.
As can be seen from the event description, four of the five failed check valves failed to reseat when the main feedwater pumps tripped. These failures obviously do not apply to the disc rupture / gross leakage mode of failure considered for the ECCS/RCS check valves. However, as it is i
described later, they are included in the estimated frequency of failure to reseat on demand. The fifth valve (feedwater regulating valve bypass line check valve) failed because of water hammer resulting from 1
1423P091686
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e TABLE 1.
CHECK VALVE LEAKAGE EVENT DATA BASE Sheet 1 of 2 eak Rate NPE Plant escripdon Range ven Reference (date)
(gpe)-
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V11. A.126 Zion 2 A leak rate'of ~0.25 gpm was detected from y 0.25 (October 1975) the "A" accumulator check ' valve - wrong size gasket installed.
~0.33 V11.A.32 Turkey Point 4 One of the three check valves in the high-head y --
(May 1973) safety injection lines to the RCS cold legs developed 1/3 gpm leakage with 180 psi of water pressure applied. Two other check valves showed only slight leakage - failure of sof t seats.
Vll.A.175 San Onofre 1 A tilting disc check valve located in the LPI y<5 1
(May 1978) system as the first valve inside containment, failed to close with gravity - valve installed in a vertical rather than a horizontal pipeline.
V11. A.114 Surry 1 Check Valves 1-SI-128; 130 leaked causing boron y < 10 t
(July 1976) dilution in the "B" accumulator.
y < 10 V11. A.182 Calvert Cliffs 2 The outlet check valves associated with the y < 10 (September 1978) safety injection tanks 218 and 228 leaked y < 10 reducing the boron concentration from 1,724 l
and 1,731 ppm to 1,652 and 1,594 ppm in-1-month period, respectively.
Yll.A.306 McGuire 1 Discharge check valves associat'ed with the cold y < 10
( April 1981) leg injection accumulator A leaked - cause y < 10 unspecified.
V11.A.343 Point Beach Check valve 1-852C, serving as the first-off y < 10 (October 1981) check valve from the RCS for the low head safety injection.
Yll. A.291 Surry 2 Check valve associated with the safety y < 20 (January 1981) injection accumulator "C" leaked, resulting in accumulator boron dilution - cause unknown.
V11.A.63 Ginna Accumulator "A" check valve leaked leading to y < 20 (September 1974) boron dilution (from about 2,550 down to 1,617 ppm) - cause unknown.
V11.A.85 Surry 1 Check valve associated with the IC accumulator y < 20
( August 1975) failed to seat, resulting in increase in accumulator level - cause unspecified.
developed 6 gpm leakage.
V11. A.105 Robinson 2 "B" safety injection accumulator check y < 20 January 1976 valve developed leakage - cause unspecified.
V.A.122 Zion 1 Discharge check valve on the accumulator ID y < 20 (June 1976) developed back leakage - cause unspecified.
V.A.407 McGuire 1 Cold leg injection accumulator check valve 20 < y < 50 (May 1983) leaked, resulting in low accumulator boron concentration - cause unspecified.
V.A.452 St. Lucie 2 The SIT outlet check valve developed excessive 20 < y < 50 (December 1984) leakage - foreign material caused ball galling leading to joint binding.
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TABLE 1 (continued)
Sheet 2 of 2 NPE Plant
'"9' Reference (date)
(gps) -
V.A.456' Calvert Cliffs 2 SIT check valve developed excessive leakage -
20 < y < 50 (January 1985).
ethylene propylene 0-ring material degradation.
V.A.437 Farley 2 Loop 3 cold leg safety injection check valve 50 < y < 100 (September 1983) developed excessive leakage - incomplete contact between disc and seat.
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V. A.273 Davis Besse 1 Gross back leakage through core flood check 20 < y < 50 (October 1980) valve - cause unspecified.
i (July 1982) -
200 gpm ring deteriorated.
-~200 V11. A.384 Calvert Cliffs 1 jlT outlet check valve leaked at the rate of y
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the failure of other valves. This failure also does not apply to the failure model of interest here.
In the process of reviewing the available data, a.recent review of eight BWR events (Reference 3) was also con.sidered. These events, listed in Table 2, involved testable isolation check valves in the pressure boundary and could be considered as precursors to an interfacing LOCA r These events were judged to be inapplicable for this study because the valves involved are'different from those considered here both in terms of design and operation. The reasons for inapplicability of each of the events are listed in Table 2.
In summary, the BWR check valves have air operators, whereas the PWR ECCS/RCS check valves are enclosed and cannot-be operated from outside. The latter group is verified seated,: either continuously (for the upstream valve) or during startup (for the downstream valve). Thus, the same mechanisms that cause the eight BWR check valves to be open and undetected do not apply to the PWR ECCS/RCS check valves.
1.2 SUCCESS (EXPOSURE) DATA To estimate the total check valve hours, the information provided in NUREG/CR-1363 on the number of valves in the ECCS and RCS in various PWRs was used. The details are prgvided in Table 3.
The total number of check valve hours is 1.0 x 10o.
1.3 FAILU8E RATE ESTIMATE
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Tne various leakage events were grouped into five leak ranges, as shown in Table 4.
For each group, a frequency per hour was estimated using the exposure time discussed above. Table 4 also provides the corresponding cumulative frequency points that are also shown in Figure 1.
The curve fit on a log-log scale was done using an IMSL code, which uses 'the least square method. The parameters of the line obtained from.this method correspond to the Bayesian most probable values based on a uniform prior distribution.
The equation of the line is
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-y = ax + b where x is the logarithm of the leak rate (gpm) and y is the logarithm of the frequency of exceedance per hour.
1 Using the data of Table 4, the following values for a and b were derived:
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Parameter a Mean
= 0.0976 i
95th Percentile = 1.0127 Sth Percentile = 0.6915 1
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TABLE 2.
SUMMARY
OF OPERATING EVENTS l
Event Percent System Date Power Involved Status Cause Reason for Inapplicability Vermont Yankee 12/12/75 99 LPCI/RLR Open Unknown PWR ECCS/RCS check valves are tested LER 75-24 and verified seated initially. They can not be lef t open undetected.
I Cooper 01/21/77 97 HPCI Open Loose Part PWR ECCS/RCS check valves are tested l
LER 77-04 Obstruction and verified seated. Any initial leakage or failure to be in the seated position will be discovered before the plant goes to power.
LaSalle-1 10/05/82 20 HPCS Open Dried Lubricant and PWR ECCS/RCS check valves do not have LER 82-115 Insufficient Preload air operators. They can not, therefore, in Air Operator; be opened externally.
Opened Bypass Line LaSalle-1 06/17/83 48 HPCS Open Thermal Binding; Check valve failed to close due to disk LER 83-066/03L Opened Bypass thermal binding. The PWR ECCS/RCS check m
Line valves are required to hold against RCS-pressure after being verified seated initially. These valves are closed and
.g stay closed. They are not cycled; i
therefore, the failure modes are different.
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LaSalle-1 C9/14/83 0
LPCI Open Maintenance Errors PWR ECCS/RCS check valves are tested LER 83-105/01T and verified seated before the plant
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goes to operation.
Pilgrim C9/29/83 98 HPCI Open Rusted Linkage on LER 83-48 Air Operator Hatch-2 10/28/83 90 LPCI Open Maintenance Errors PWR ECCS/RCS check valves do not have LER 83-112/03L on Air Operator air operators and will not open due to a similar maintenance error.
Browns Ferry-1 C9/14/84 100 LPCS Open Maintenance Errors
~ PWR ECCS/RCS check valves do not have LER 83-032 on Air Operator air operators and will not open due to a similar maintenance error.
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TABLE'3. CHECK VALVE EXPOSURE DATA N6mber of Start of Number of Check Valves Total Number of Plant Name Code Commercial Operation Years in ECCS Check Valve Ho_urs Arkansas Nuclear One 1 AR1 December 1974
'10 20 1.75+6 Crystal River 3 CR3 March 1977 7.7$N 23 1.56+6 Davis-Besse 1 DB1 November 1977 7.08 s.
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- 1. 61 + 6 Oconee 1 OE1 July 1973 11.42
. 20 2.00+6 Oconee 2 OE2 March 1974 10.25 21 1.89+6 Oconee 3 OE3 December 1974 10 21 1.84+6 Rancho Seco RS1 April 1975 9.67 30
- 2. 54+ 6 Three Mile Island 1 TIl September 1974 10.25 19 1.71+6 Three Mile Island 2 TI2 December 1978 6
19
- 9. 99+ 5 Arkansas Nuclear One 2 AR2 March 1980 4.75 30 1.25+6 Calvert Cliffs 1 CC1 May 1975 9.58 45 3.78+6 Calvert Citffs 2 CC2 April 1977 7.67 45 3.02+6 Fort Calhoun FC1 September 1973 10.25 45 4.04+6 Millstone 2 M12 December 1975 9
47 3.71+6 i
Maine Yankee MY1 December 1972 12 49 5.15+6 Palisades PA1 December 1971 13 21 2.39+6 St. Lucie 1 SL1 December 1976 8
30 2.10+ 6 I
Beaver Valley 1 BV1 April 1977 7.67 36 2.42+6 D. C. Cook 1 DC1 August 1975 9.33 34 2.78+6 i
D. C. Cook 2 DC2 July 1978 6.42 34 1.91+6 1968 14 27 3.31+ 6 Januarh74 10.42 36 3.29+6 Haddam Neck HN1 Indian Point 2 IP2 July 1 Indian Point 3 IP3 August 1976 8.33 45 3.28+6 Joseph M. Farley 1 JF1 December 1977 7
33 2.02+6 Kewaunee KE1 June 1974 10.5 19 1.75+6 North Anna 1 NA1 June 1978 6.5 3
2.05+6 Prairie Island 1 PRI December 1973 11 23 2.22+ 6 Prairie Island 2 PR2 December 1974 10 23 2.01+6 Point Beach 1 PT1 December 1970 14 21 2.58+6 Point Beach 2 PT2 October 1972 12.17 21 2.24+6 R. E. Ginna 1 RG1 March 1970 14 21 2.58+6 H. B. Robinson 2 R02 March 1971 13.75 25 3.01+6 Salem 1 SA1 June 1977 7.5 32 2.10+ 6 San Onofre 1 S01 January 1968 14 18 2.21+6 Surry 1 SUI December 1972 12 25 2.63+6 Surry 2 SU2 May 1973 11.58 25 2.54+6 Trojan TRI May 1976 8.58 22 1.65+6 Turkey Point 3 TU3 December 1972
.12 34 3.57+6 i
Turkey Point 4 TU4 September 1973 11.25 34 3.35+6 2
Yankee Rowe YR1 June 1961 14 17 2.08+6 Zion 1 ZIl December 1973 11 50 4.82+6 Zion 2 ZI2 September 1974 10.25 50 4.49+6 Total 1.08+8 6
NOTE: Exponential notation is indicated in abbreviated form; i.e., 1.75+6 = 1.75 x 10,
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TABLE 4.
STATISTICAL DATA ON CHECK VALVE LEAKAGE EVENTS IN PWR, ECCS, AND RCS SYSTEMS F equenc Frequency of Leak Rate Number of gf 9 e ce (gpm)
Events Exceedance (per hour) 5 3
2.94-8 2.06-7 10 7
6.86-8 1.77-7 20 5
4.90-8 1.08-7 50 4
3.92-8 5.90-8 100 1
9.80-9 1.96-8 200 1
9.80-9 9.80-9 NOTE: Exponential notation is indicated in abbreviated form; i.e., 2.94-8 = 2.94 x 10-8, 7
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, 3 ggggg g
g g g g g gg g i
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s 6
5 g
g 4
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N 3
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g N
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N 2
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7 LEGEND:
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-- - - STATISTICAL BOUNDS 4
AT 00% CONFIDENCE sN 3
BEST FIT
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g 2
- - ASSUMED 95TH AND STH PERCENTILES
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1 x 10'I0
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8 7
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3 2
1 10*II
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10 100 g,000 10 CHECK VALVE LEAK RATE (GPM)
FIGURE 1.
FREQUENCY OF CHECK VALVE LEAKAGE EVENTS 8
P e
Parameter b Mean
= 13.6943 95th Percentile = 14.2862 5th Percentile = 13.1024 Based on the above values, the "best-fit" line is
-y = 13.6943 + 0.0976 x N
with the following bounds:
b 95th Percentile = -y = 14.2862 + 1.0127 x Sth Percentile = -y = 13.1024 + 0.6915 x t
These lines are shown in Figure 1 as the "best-fit line" and " statistical bounds at 90% confidence." To account for uncertainty in the assessment of the leak rates, classification of data, estimation of exposure data, and the applicability of the data to the check valve and failure mode of concern in this analysis, the statistical bounds were further stretched by increasing the range factor of the frequency at 150 gpm from 3.7 to 10 and increasing the range factor of the frequency at other points proportionally (to RF = 14 at 1,800 gpm). The resulting new bounds are also shown in Figure 1.
2.
FAILURE TO RESEAT ON DEMAND To estimate the frequency of check valve failure to reseat on demand, two types of data were used:
(1) estimates from several generic sources of failure data, and (2) experiential data from eight U.S. nuclear power plants based on plant-specific PRAs performed by PLG.
Since the majority of data sources provided information on check' valve failure on demand without specifying failure to open and failure to close modes separately, the distribution developed here is based on failure on demand data. Review of check valve failure events from several plants indicate that the distribution is a good (and perhaps even conservative) estimate of the failure to reseat frequency.
An additional piece of information provided by the San Onofre event of November 1985 (Attachment C) was also incorporated into the estimate of check valve failure on demand frequency. Four of the five check valve failures (failures involving pump discharge check valves) apply to this mode of failure. NPE was reviewed for the period J9nuary 1,1971, through June 30, 1986, to see if there have been other check valve failures in the San Onofre main feedwater system. None were found.
The corresponding success data (number of demands) were developed by assuming an average of 10 system-wide demands per year, a population of eight check valves, and 15.5 years of operation from January 1,1971, through June 30, 1986. This resulted in an estimated 1,240 check valve 1423P091686 9
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demands. The corresponding failure frequency estimate (counting four of the five check valve failures) is
-3 Aso " IE Tg = 3.2 x 10 per demand.
This value was used together with the generic estimates as well as
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j plant-specific data from other plants in a Bayesian updating process described in Reference 4 to develop the fa^1 lure on demand frequency distribution.
N -
The following summarizes the data used.
e Generic Estimates Source Estimate Assigned Range Factor
- WASH-1400 1.00 x 10-4 5
NUREG-1363 1.10 x 10-4 3
i EPRI-81 7.00 x 10-5 10 e
Data from Nuclear Power Plants Number of Plant Events Number of Demands Oconee 3
6,855 Zion 0
6,970 Indian Point 2 0
1,440 Indian Point 3 0
1,550 Beznau (2 Units) 7 28,978 Pilgrim 0
2,394 t
TMI-1 12 8,716 e
San Onofre Unit 1 Main Feedwater System Check Valves Estimate
= 3.23 x 10-5
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Assigned Range Factor = 5 (A moderate range factor is used to represent higher degrees of uncertainty than indicated by the estimated four events in 1,240 demands.)
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- The assigned range factor (ratio of the 95th to the 50th percentile of lognormal) represents our uncertainty of the accuracy of the estimate.
See Reference 4 for the details of the methodology.
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The resulting distribution is shown in Figure 2.
Some key characteristics are:
i 95th 5th 1
Mean Median Percentile Percentile i
5.46-4 1.18-5 1.58:4 1.63-3 NOTE: Exponential notation is indicated in abbreviated form; i.e., 5.46-4 = 5.46 x 10-4 r
j 3.
REFERENCES l
1.
Flem'ing, K. N., A. Torri, K. Woodard, and R. K. Deremer, "Seabrook l
Station Risk Management and Emergency Planning Study," prepared for Public Service Company of New Hampshire and Yankee Atomic Electric j
Company, confidential, PLG-0432, December 1985.
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2.
Nuclear Power Experience, S. M. Stoller Corporation, updated monthly.
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3.
U.S. Nuclear Regulatory Commission, " Preliminary Case Study Report, Overpressurization of Emergency Cooling System in Boiling Water Reactors," February 1985.
4.
Mosleh, A., "PRA Data Base," Pickard, Lowe and Garrick, Inc.,
PLG-0500, 1986.
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es he A
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1 WASH-1400
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NU R EG-1363 EPRI81 SAN ONOFRE 1 f
h LUMPED DATA FROM 8 PLANTS (22 FAILURES IN 56,903 DEMANDS) 3 O
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10-6 10-5 10 10-3 10-2 4
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FAILURES PER DEMAND l
FIGURE 2. -CHECK VALVE FAILURE ON DEMAND FREQUENCY DISTRIBUTION
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ATTACHMENT A 1
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PAGE 1
-NPE CdECK VALVE LIST!44
- 693 ARTICLES TOTAL
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Tz[RE ARE 73 4RTICLE(5) TO BE DISPLAYED:
BWR - CNECE VALVE FAILURES - 19 b-1975 3.04.8.0010 : VARIDus CONTROL SLADE, DRIVE PROBLEMS
--- 3RESDEM 1 - 1960-72 s.04.5.3036 : DISE AND SEAT SUEFACES DIRTT - CRD CNECK WALVE LEACAGE
--- DRESDE4 2 - M47 F5 - REFUELING SNUTD0d4 5.05.E.0002 : CNECE VALVE LE4KAGE
--- DRESDEN 2 - SPRING 72 o
s.05.E.0003 : FW CNECE VALVE LE4K4GE
--- DRESDEM 2 - SPRING 71 9.05.E.0014 :
FW CNECK VALWE LEAK 43E - CHAMEED T3 $!LIC3M SEAL RING
--- DRESDE4 3 - SPRING 1974 S.06.E.0018 : 0-RINGS ADDED TO FW CNECE VALVES
--- DRESDEN 3 - IST 1/2 0F 1773 J
S.06.E.0022 : CHECC VALVE
- LUG LEAC
[
--- OTSTER CREEK - JULT 73 t
I
..)
l 0.05.E.0023 : $1LICONE 0-RINGS UN4CCEPTABLE - V4LVES LE4KED
--- DRESDE4 2 - M4R S M47 74 3.05.E.0024 : FW CHECE DISC DID NOT MATE WITN SE47
.g ()
--- MONTICELLO - SPRING 74
,/
5.05.E.0029 : LEAKT SASKET 2N FW CHECC W4LVE
--- ARNOLD
- JULY 74 (POWER ESCALAT!3N TESTING) i
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l 9.08.E.0034 : LEAK 4GE BETWEEN CNECK VALVE SEAT RIN3 AND V4LVE 3097
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--- au49-CITIES 1 - APR 74 (REFUELING $NUTD3WN) 3.05.E.0040 : FL4W TN FW CAECK VALVE J
l au49-CITIES 2 - FEs 75 - aEFUELING SMUTDOWN l
8.06.E.0042 : CNECE VALVE SEAL RING LE4EAGE
.)
--- QU4D-CITIES 2 - DEC F4 8 75 - REFUELI46 SNUTD0dm
.)
9.06.E.0043 : SEAT LE4KAGE 449 CRACEED SELLOWS IN FW CNECES
--- PEACM 30TTon 2 - M AY B JUNE 75 - SNUTDOWN AND 45S P3WER l
S.08.E.0048 : ERODES CONTR3L VALVES
--- TARAPUR 1 & 2 - 1970 9.05.E.0058 : FW C1Ett VALWE 0* RINGS C04*RESSED - EXCES$1VE LE4(4GE
--- DRESDEN 3 - APR S NAT T5 - REFUELING $NUTDOW1 b
3.05.E.0059 : FW CNECE VALVE 0-41465 DISASSOCIATED
--- DRE5DEM 2 - N3V 74 - REFJELI4S 5 NUT 90d4 II l
C.05.E.0075 714 NCLE LEACS IN FW CMECC VALVE SEAL PLATE, CRACKED CD4W3 LUTE
--- PEACE 50ffem 3 - AUG 75 - STARTI46 UP
- -~m C)l
PAGE 2
o 8.06.E.0051 : CRACcED FW VALVE SEAL PLATE 3ELLOWS
--- PEACM BOTTOM 2
- JUNE 75 - 740 MWE J
5.05.G.0002 : MISC. TURBINE CYCLE STSTE95 PROBLE15
--- MUMLESER3 - 1972-74 J
3.07.4.0015 : CHECC VALVE 2!SC-MINSE JOINT FRACTURED
--- 3UAD-CITIES 1 - APR 74 (SHUTD0dN FOR REFUELING)
()
3.07.4.0020 : RCIC TUR8!NE EXHAUIT CHECC ASSEM9 LED INC0tRECTLY
--- NONTICELL3 - JAN 75 - SHUTDOWN
()
S.07.A.0031 : CHECC VALVE SLOWI1G STEAM, R3D CONTR3LS NOT CHECKED FITIPATRICC - JULY 75 - (POWER ESCAL4 TION TESTINS)
J SMUTD0d4 S.07.4.0032 : EXH4UST LINE CHECC VALVE CLAPPER DISCONNECTED 3RUN$d!Cf 2 - APRIL 75 - (POST CRITIC 4L TESTING) SE P3WER 5.07.A.0034 : EXH4UST CHECC VALVE FLAPPER WAS J41MED l
--- SRUNSWICC 2 - OCT 75 - 63% POWER S.07.4.0040 : RCIC TUR8INF DISCHARGE CHECC VALVE FLAPPER-BISC STUD JROCEN
--- 3AUN$dICs 2 - FEB 75 - NOT ST4ND3Y
~
a.07.A.0043 : TORUS O!$CH4t3E VALVES NAD DIRTT SEATS AND MISSI4G FASTENERS
--- COOPER - 3CT 75 - SHUTD0dN 8.0T.C.0014 : SCALE ON VALf! SEATS SJ e
--- 90NTICELL3 - SPRING 74
> ')
5.07.C.0030 : SCALE ON CHECC VALVE SEATS
--.904TICELL3 - JAN 75 - SNUTBOWN s.07.C.0042 : IMPROPERLY A3 JUSTED CHECC VALVES
--- 3RUNSWICC 2 - MARCH 75 (POST CRITICAL TESTING)J 5t POJER 2
~
a.07.C.0043 : CNECC VALVE $3AL LEACED
--- TARAPUR 1 8 2 - 1969 04 FO l
3.07.C.0044 : STUCC CHECC d4LVES l
TARAPUR 1 & 2 - PRIOR TO 1973 1
3.0F.C.0047 : SCALE ON SEATING SURFACES OF CORE SPRAY CNECC VALVES l
--- 90NTICELi3 - SEPT 75 0.07.C.0061 : CORE SPRAY CNECC VALVES SEATED IMPROPERLY - DISCNARGE NE40ER I
DEPRESSURIZED
--- DRESDEN 2 - JAN T5 - REFUELING SNUT80d4 --- DRESDEN 3 - 4UG
{
T4 3.0F.D.0016 : WIRE AND G1!1CER ARBOR FOUND IN RNR SYSTE1
--- 3UAD-CITIES 2 - JAN 73 g
S.07.0.0035 : SCALE 04 VALdE TEAT
--- 90NTICELLO
- SPRING 74 C'
3.07.D.0061 : SCAL! 04 CHECC VALVE SEAT Mc4 tit!LLO - JAN 75 - SMUT 9044 5>
-~~
~.
.-s.,u-...-
_.._____._.m m__
e PAGE 3
e.
3.07.9.3076 : RNR AND HPCI VALVE PACKINE LEAKS
--- 3RUNSWICC 2 - SEPT 75 - 52% P0 DER 3.07.9.0082 : VALVE LEAKAGE, LPCI LOOP OVERPRES$URIIED - NEAT EXCHANGER GA$tET FAILED gj WERMONT TANKEE - der 75 - 993 P0 DER 3.07.0.0106~ : T04US SPRAY WALVE OPERATOR MOUNTINS PLATE WELD FAILED, CRACKED Y3CE ARM - IMSUFFICIENT diLD PENETRATION, UNDER$1IED MOUNTING 80LTS
[g
--- 3ROWNS FERRY 1 - MAY l SEPT T4 - COLD $NUTD0dN 3.07.E.0007 : STEAM VALVE DISC PIN FAILURE - RUPTURE DISCS RUPTURE - TE1P SdITCM DAMAGE
--* MONTICELLO - JULY 72 3.07.E.000B : STEAM VALVE LEAKAGE
--- MONTICELLO - JULY 72 3.07.E.0011 : LOOSE RUST - CNECC VALVE LEACED TORUS WATER BACK TOWARDS 1PCI l
TUR31NE
--- PILERIM - JULY 72 (PodER ESCALATION TESTING)
\\
J l
5.07.E.0018 : VACUJM IN TURBINE EXNAUST - DATER DAMMER
--- VERMONT TANKEE - 1971 (PREOPERATIONAL TESTING) 3.0T.E.0025 : WATER NAMMER OF TURBINE EENAUST - INSTALLE9 COMBENSING SPARGER
--- DROWNS FERRY 1 - OCT T2 (PRESP TESTING) 5.07.E.0048 : GLAND CONDENSER GASKET FAILED - POS$!DLE INJECTION VALVE LEAEAGE l
--- 3ROWNS FERRY 2 - NOV T4 - 421 P0 DER
/
g l
B.07.E.0053 : SURFACE FLAWS IN VALVE DISC AND SEAT
/
I,
--- PEACM DOTTOM 2 - FEB F5 - SNUTDodM p
8.0T.E.0054 : BENT WINGE PIN AND SAD DISC ON TURSI4E EXNAUST CHECK VALVE
--- MONTICELLO - JAN 75 - SNUTDOWN 9.07.E.0058 : MISSING WELD IN SWING CHECK, INSUFFICIENT TACK WELDS IN STOP CHEC(, VALVE DISSASSEMBLED WITH COOLANT, GREATER TNA4 212 DECREES F
--- ARNOLD - APR 75 l
~
S.07.E.0061 : STEA9 VALVE LEAKED
--- SUAD-CITIES 2 - DEC 76 - REFUEL!16 SNUTB0WN l
(*
8.07.E.0077 : SCRATCHED SEATING SURFACES - TURBINE EINAUST VALVE LEAEASE
--- MONTICELLO - SEPT F5 j
k J
9.07.E.0130 : HPCI TURRINE EXHAUST CNECE VALVES LEAKED - WORN WALVE SEnts t
--- 3UAD-CITIES 2 - DEC T& G SEP 75 - REFUELING=== 33AD CITIES l,
1 - JAN 76 - REFUELING 9.07.7.0007 : VARIJUS PROBLEMS WITN VALVE LIMITOROUE OPERATORS
--- 3RESDEN 2 & 3 - DEC 70 - JAN F1
)
8.09.C.0055 : DIRT IN SERVICE WATER CNECK VALVE
--- OYSTER CREEK 1 - NAT 75 - 390 MWE.
O v
m
a PAGE 6
)
s.03.C.3064 : Cross THREADED PIPE $ LOOSE VALVE 50LTS - LEAKS BRUNSWICC 2 - JUL 75 - FI POWER S.09.F.0006 : CHCCC VALVE MAD MI551NG DISK AND P3P*ET
--- SUAD-CITIES 1 - JUL 76 - SHUTD3W4 41 s.09.8.0007 : INSTRUMENT CHECK VALVE 5 HAD dEAK $PRINGS, PITTED PDPPETS AND BAD SEAT
--- 20AD-Cf7IES 2 - APR 75 - COLD SH3TDOW4
()
S.09.G.3147 : DAMAGED SEAT ON INSTRUMENT N2 VALVE
--- PEACH BOTTOM 2 - MAY 75 - 100E P3WER
()
3.11.A.0057 : STUCC DIESEL FUEL BALL CHECK VALVE
--- VERMONT TANKEE - JUL 74 J
B.11.A.3103 : STUCK DIESEL AIR CHECK VALVE
--- 3RUNSWICC 2 - DEC 75 - 360 MWE J
8.14.8.0017 : SEC040ARY C04TAIN1ENT VIOLATIONS
--- DRE5 DEN 2 8 3 - OCT 71 - MAR T2 5.14.8.0037 : CHECK VALVES LEAKEP
--- DRESDEN 3 - JAN 73 s.14.8.0056 : CONTAINMENT CHECK VALVE MISALIGNME4T
--.4CNTICELLO - NOV 73 s
8.16.8.0059 : LEAKING VACUJM BREAKER PENETRAT!3NS - INADVERTENT RELEASE b
--- PEACH 8OTTOM 2 - DEC 73 (POWER ESCALATION TESTIN3) ej 1
5.14.8.3072 : VACUUM SREAKER LEAKA3E
/
U
--- 3YSTER CREEK 1 - APR 74 - 642 MWE 3.14.8.3081 : DEFORMATION 3F CHECK VALVE RUBBER SEAL
/
--- PEACM BOTTOM 3 - MAR 75 - 100% P3WER
~
EI y
S.16.B.3088 : REVERSE FLOW CHECKS REMOVED FROM S;TS - M3 TOR CVERLCAD
--- COOPER - MAY 75 - COLD SNUTDod4 3.16.s.3115 : CHECC VALVE PISTON FOUND MUNS UP
--- PEACM 80TT3M 3 - OCT 75 - 55% PodER 5.15.A.0007 : THIN WALLED VALVES - 8WRS IN GENERAL'- 1970 ((P.15.31) s 3.15.A.0037 : NITR3 GEN ACCJ1ULATOR VALVE LEAKED 3RUNSWICC 2 - DEC 75 - SSI POWER
.)
3.16.C.0001 : GUESTION ON SUITABILITY OF CERTAIN PIPING AND VALVES
--- DYSTER CREEK 1 - JUL $9
{
-J S.16.C.0036 : PROCEDURAL PRCOLE1 - LPCI VALVE MOTOR BUR 4ED OUT DURI4G T!$T.
--- VERMONT TANKEE - MAY 73 B.15.C.3099 : PROCEDURAL PR3SLEM - EXCESS FLOW CHECK VALVE'S BYPASS VALdES LEFT CPEN (I
--- PEACM 30TTOM 2 8 3 - NOV 74 - 103E 8 SOE POWER
............. ALL Deut, *REss <RETuRu> CEv To so Ts MEwu
.................?
j g) I
~
- - - - - ~ _ - - - - - - - - -. -
~. -.. -. -. _ - -
-... ~ -.
PAGE TutRE ARE 123 ARTICLE (5) TO SE DISPLATED:
19-4UG-1986 13:37:37 BWR CHECK VALVES, 1976 - 1980 5.06.5.0047 : CRD RETURN LINE ISCLATION VALVE N43 WORN SEAT, RJST ON PISTON
--- 3RESDEN 3 - OCT 76 - REFUELINS SNUTDOWN B.04.3.0116 : CONTRCL ROD DID NOT LATCH - DIRECTIONAL VALVE FAILED BROWNS FERRY Z - JUN 80 - 401 POWER 3.04.e.0117 : WEST SDV NEA3ERS DID NOT DRAIN AFTER MANUAL SCRAN
--- 3RESDEN 3 - JUL 80 - SMUTDOWN 3.04.8.0145 : CONTROL RCD 3 RIVE SYSTEM 1ALFUNCTIONS C
3RUNSWICK 1 - AUG 80 - SJBCRITIC4L (STARTUP) --- 3RUNSWICK 2
- FES 81 - 1.5% P3WER --- OYSTER CREEK -N3V 80 - 63% PodER' 3.05.C.0257 : THREADED LOCKING DEVICES 3N VALVES, PUMPS, VALVE OPERAT3R$
FAILED
--- SWR'S IN GENERAL - MAR 83 ((P.083.293))
5.05.C.0261 : ADS LIA SUPP.T ACCUMULAT3R CNECK WALWES LEAKED NATCM 2 APR 53 - COLD SHUTDOWN l
3.05.C.3266 : ADSA AIR SUP*LY ACCU 1ULAT3R CNECK WALVES LEAKED
--- COOPER - APR 40 - REFUELING J
3.06.E.0054 : RUST ON FW C4ECK VALVE SEATS
--- PEACH 50rTOM 2 - MAR F6 SMUTD0dM J
l 3.05.E.0055 : FW CHECK VALtE LEAKAGE - EXCESSIVE DISC-T3-SEAT CLEAR 4NCE g,
--- DRESDEN 2 - 1AR 76 - REFUELING SNUTD0dN ty
~
3.06.E.0060 : FW CNECK VALiES LE AKED - WITON 0-RINGS DETERIOR ATED
/
I
--- SUAD-CITIES 1 - JAN TS - REFUELI15 SMUTDOWN
~
3.05.E.0064 : DIRT IN FW CmECK VALVES, IMPROPER DISC-TO-SEAT CLEARANCE,'
/'
NI EXCESSIVE LEAKAGE - TESTING METHODS WERE INADEGUATE
,s l
--- DRESDEN 3 - SEPT & OCT 76 - REFUELING SHUTDOWN 3.05.E.0065 : FW SAMPLE PR35ES LODGED IN HPCI & FW CNECK VALVES, FL4NSE J
GESKET BLOWN l
--- COOPER - JAN 77 - APPROXIMATELY 76I POWERI BRUNSWICK 2 - FEB l
76 - SHUTDOW4 4.05.E.3076 : WORN FW VALVE SEAT / DISC ASSE10LY PINS, DEFORMED. SEAT RI163 DRESDEN 2 - OCT 77 - REFUELINS $NUTD0dN t.06.E.0086 : FW CHECK VALWES SEATS MACHINED, SE4T SEAL 0-RINGS CHAMGED TO
~
KALREZ J
I 3UAD-CITIES 2 - SEPT F6 - REFUELING i
8.06.E.'0093 : W3RN SEAT / DISC ASSEM0LY PINS IN Fd CNECES.
l DRESDEN 3 - MAR 78 - REFUELINS
~
-S.06.E.0097 : WDRN SEATS A43 RINGS IN Fd CNECK VALVES 05
--- MILLSTONE 1 - MAR 78 - REFUELING S.06.E.0107 : DIRT ON FW CdECK VALVE SEAT f)
^
--- auAD-CITIES 1 - JAN 77 - REFUELING a.05.E.0109 : WORM BUSNINGS - FW CNECK WALVES LEAKED
.._._._..___m__. _ _
PAGE 6
~
r 3RJNSWICC 1
- APR 79 - REFUELING J
S.01.E.0110 : WORN FW CHECT VALVE SUSNINGS 3RUNSWICC 2 - MAT 79
--- MILLSTONE 1 - JUNE 79 - REFUELINS 8.05.E.0116 : FW CHECK VALVE KING PIN C3VER LEAK - MIGN CONTAINMENT TEMP
--- 3RUNSWICK 2 - AUG 79 - 96T POWER
!~
5.05.E.0127 : FW CNECK VALVE LEAK - WOR 4 SE4T/ DISC ASSE18LY PINS l
--- DRESDEN 3 - FES 80 - REFUELINS J
S.05.E.0131 : FW LEACED IMT3 YORUS UNDETECTED - CHECK V4LVE 804 NET SE4L FAILED VERMONT TANCEE - JUNE 80 - 35E P3WER S.0$.E.3134 : FW CHECK V4LWE SEAT / DISC ASSEMBLT P!45 W3RN
--- DRESDEN 3 - M4R SD - REFUELING 3.05.E.0141 : WORN SEAT, DIRT IN FW CHECK VALVES
--- 3U AD-CITIE S 1 - S E PT 80 - REFUELING SHUTDQvN l '
3.Or.A.0058 : BENT DISC WASHER, IMPROPER SEATING I4 RCIC EXNAUST VALVE
--- 90NTICELLO - 3CT 77 - REFUELINS SMUTD3WN 3.07.A.0059 : 200GN SEATINS SURFACES IN RCIC STE4M VALVES
--- COOPER - OCT T7 - REFUELINE SHUTDOWN Il
~
5.07.A.3062 : RCIC TURSINE STE41 EXHAUST CNECK VALWE LE4KED - FLAPPER B40KE 9
--- 3U43-CITIES 2 - SEPT T6 - REFUELING
/
I) l S.07.A.0071 : R30GH AND CR4CKED EXNAUST VALVE SE4TS
--- COOPER - APR 78 - REFUELING
/
bI t
3.07.A.0075 : DIRT ON TU481NE EENAUST VALVE SEATS l
s I
--- MONTICELLO - OCT 78 - REFUELI45 J
3.07.4.3076 : RCIC VALVE PIN $ NEARED 3ROWNS FERRY 1 - DEC 78 - REFUELING
(
S.07.A.0080 : DIRTY VALVE INTERNALS I
--- 3UAD-CITIES 1 - J AN 77 - REFUELING B.07.A.0083 : LOOSE VALVE 315C BLOCKED RCIC TU4 SINE EXNAUST LINE - BLOW 4' i
RUPTURE DISC l,
--- MATCH 2 - JUNE TD - DE P3WER 3.0F.4.0083 : DEFECTIVE TOROUE SWITCN, ROUSN SEATS - STEAM SUPPLY AND EEN4UST l
VALVES LEACED COOPER - APR 79 - REFUELING k
5.07.4.0094 : RCIC TURRINE FAILED TO ST4RT - BOLT LODGED IN STEAN EEH4UST gg CNECK VALVE
--- 3RUNSWICC 1 - NOV 79 - SNUTDod1 C
3.07.4.3095 : RCIC TRIP - CHECK VALVE L3CK 840KE1, VISR4TED CLOSED NATCH 1 - DEC 79 - 834 P3WER N
0.07.A.0096 : Rc!C DnAIN P3T LEVEL VALVE P4CKING RJPTURED - NORMAL dE4R
~
e PAGE F
l i
l l
--- ARMOLD - OCT 79 - 93E PodER 9.07.A.0103 : RCIC SYPASS VALVE TOR 30E SWITCN C3174 CTS DIRTY 3ROWNS FIRRY 3 - JULY 80 - 69I P3WER 8.07.A.0121 : RCIC TUR8INE TRIPPED - STEAM EXHAUST CHECC VALVE F4ILED 3RUNSWICC 2 - NOV 80 - 57I POWER
~
5.07.8,0018 : SSLC PUMP CHECC VALVE LEAES
--- QUAD-CITIES 2 - MAR 76 - SHUTDOWN 5.07.8.0028 : CHEMICAL DEP3 SITS ON $8LC VALVE SE4T
--- DRESDEN 3 - APR 78 - REFUELING SHUTDOWN 9.07.C.0070 : SCALE ON VALdE SEATS
--- 10NTICELLO ' SEPT 77 - REFUELING SHUTDOWN '
8.07.C.0083 : CORE SPRAY CHECK VALVE REMOVED FOR M4INTE1ANCE - CONT 4IMMENT VIOLATION
--- PILGRIM 1 - JUNE 79 - 103E POWER J
8.07.C.0090 d4TER HAM *ER 34M4GED P!PE SUPPORTS, VALVE 80LTS LO3SE 11LLSTONE 1 - FE3 80 - 100E P3WER j
S.07.D.0103 : LPCI CHECK VALVE PACCING LEAC
--- MILLSTONE 1 - SEPT ?S - 72% P3 DER 1
3.07.D.0107 : NUT N3LDING 4HR CHECC VALVE CLAPPER SHAFT VAS MISSINE
--- FITIPATRICC - OCT 76 - 253 POWER,
g; 7
8.07.0.3105 : DIRT AND SCR4TCHES 04 LPCI CHECC VALVE SEAT i
--- DRESDEN 3 - OCT 76 - REFUELINE SHUTDOWN
/
g 3.07.D.0115 : DIRT ON REACT 3R HEAD COOLING ISOL4T!3N VALVE SEAT AND DISC P!VOT PIN
/
()
~
--- DRESDEN 3 - NOV 76 - REFUELING $4UTD0dN
~
5.07.0.0116 : SMEARED WOODRUFF EEY IN RNR VALVE OPERATOR
--- COOPER - FEB FT - 96% POWER 3.07.D.0119 : WORN LPCI VALVE C3MPONENTS, SENT M373R $NAFT - VALVE SIMDING
--- BR0dNS FERRY 1 - MAR 77 - 851 POWER
~
5.07.0.3153 : STRIPPED LPCI VALWE LOCKNUT, DAMAGED YOKE AND OE44!NSS - WALVE J
BINDING
--- 3R0dNS FERRY 3 - FEs 78 - 100X P3WER J
l 3.07.0.0169 : LOV DP, RHR VALVES FAILED TO SEAT FULLY - HIGH T3RUS LEWE'.
f
- -- SRUNSWICC 2 - NOV 74 - H3T SHUTD3WN k
~#
8.07.0.0171 : LOOSE DISC RETAINING NUT, CHECE VALVES STUCK OPEN
--- FITIPATRI*K - MAR 5 DEC 78 - 35 POWER N
5.07.D.0180 : CRACCED LPCI CHECE VALVE DRAIN LINE
--- DRESDEN 2 - MAR 19 - REFUELIN3
()
l*
8.07.D.0185 : BROCEN RHR CdECK VALVE DISC STEM NUT PIN l
--- FITIPATRICC - JUNE 77 - COLD SHUTDOWN 6I l
8.07.0.0220 : TORUS SUCTIO4 VALVES FAILED DUE TO STEAM LEAK FR3M A CHECC V4LVE 1
l
1
[
I PAGE B
--- MATCH 2 - MAY $3 - 99E P3WER 9.07.D.0236 : RNA CHECE VALVE DID NOT SEAT
--- WERMONT TANKEE - SEPT 80 - 89E P3WER 3.07.0.0238 : INAD!Qb4TE $JPPORT - LPCI DRAIN LINE WELD LEAK
--- DRE5 DEN 2
- MAY $3 - SNUTDOWM 3.07.D.3259 : CHECC VALVE STUCC - PLANT MODIFICATI3N INSTALLED
--- 3RUNSWICC 2 - NOV 80 - 6SE POWER j
5.07.E.0076 : HPCI STEAM C4ECE VALVE GASKET FAILED
--- 3UAD-CITIES 1 - JAN 76 - REFUEL!1G SNJTDOWN
~
3.0F.E.0081 : IMPR3PER MOU4 TING - CORROSION AND PITTING ON NPCI VALVE SEAT
--- DRESDEN 2 - MAR 76 - SHUTDOWN s.07.E.309C : SIMDING IN M*CI TURSINE EXNAUST VALVES - RUPTURE DISC BLOdN
--- 3RDWNT FEaRT 3 - AUG 76 - 3E P3WER 5.07.E.0095 : LOD $E RETAINING NUTS, 8ROCEN RETAINI1G STUDS t!$ LODGED DISC IN HPCI TUR3INE EXN40$f CHECC VALVES j,
--- BRUNSWICC 1 8 2
- OCT 76 - COLD SHUTD3WN l
3.07.E.0113 : HPCI TURBINE EXMAUST VALVE DISC STUD FAILED, M!$ SING PARTS
--- C00PER - APR 77 - COLD $NUTD0d1 l
S.07.E.3114 : HPCI FLOW OSCILLATIONS AT LOJ SPEED - CNECK VALVE DISC NI1GE FRACTURED 3RUNSWICC 1 - MAT 77 - C3LD $NUTDOWN i
.j S.07.E.3121 : DIRTY EXHAUST VALVE SEATS
/
(
--- 90NTICELL3 - SEP 77 - REFUELINS SHUTD3WN
~
5.07.E.3122 : R305N AND SCRATCHED VALVE SEATS
/r
--- COOPER - 3CT 77 - REFUELING SMUTDOWN
~
SE y
3.0F,E.3136 : STICKING MPCI CHECE VALVES, WATER MAMMER - SNUsBER SNAFT BROCEN BRUNSWICC 2 - MAR 78 - $NUTD0d1 3.07.E.0139 : MECMANICAL IfTERFERENCE IN CHECK VALVE
--- 3ROWN5 FERRY 2 - MAR 78 - REFUELING 3.07.E.0152 : GLAND SEAL C4ECE VALVE LEAKED
--- PEACH OOTTOM 2 - OCT TB - 51E P0 DER s
~
3.07.E.3155 : DIRT ON TURaINE EXNAUST VALVE SEATS l
--- MONTICELL3 - OCT TS - REFUELINS J
B.07.E.0161 : CHECC VALVE 345KETED SEAT FAILED i
l
--- SUAD-CITIES 1 - JAN T7 - REFUELI1G
\\
5.07.E.0209 : TORUS SUCTIO4 VALVES FAILED DUE TO STEAM LEAK FROM A CHECC VALVE l
--- NATCH 2 - MAY 80 - 99E POWER 00 8.07.E.3213 : FOREIGN MATERIAL BETdEEN CHECK VALVE BODY & SEAT l
--- SROWNS FERRY 2 - SEP 50 - REFUELING SNUTDOWN E) 5.07.F.0057 : UNIO 4 SETWEE4 ISOLATION' FLOW CHECC VALVE AND STEAM LI4E F' 0W
$EN534 dAS CROS$-TNREADED Sh
9 PAGE 9
--- DRESDEN S - N3V 75 - SE POWER 8.07.F.0050 : TILTING DISC CHECK V4LVE INSTALLED I4 VRO1G ATTITUDE 4LL 8WR$ - JUL FS ((P.07.A.1753) s.07.F.3065 : PE8BLE IN FIRE PUMP CHECK VALVE FITZPATRICK - AUG 79 - C3LD SHUTDOWN 5.07.F.0066 : FIRE DRAIN C4ECK VALVE M4D OFF CENTER DISK
--- SRUNSWICC 2 - JUL 79 - SSI POWER l '
3.07.F.0070 : VALVE SLOW T3 CLOSE - SWITCH NEEDED 4DJUSTMENT PILGRIM - OCT 79 - 83E 3.07.F.0071 : VARI 3US VALVES LEAKED - DIRTY SEATS *REVE4TED CL3SURE
--- PEACH BOTTOM 3 - NOV 79 - REFUELING S.07.F.0072 : VARI 3US CONT 4INMENT ISOL4 TION VALVES' LEAC RATES UNACCEPT48LE MATCH 1
- APR, M4f, DEC 79 - REFJELIN3 AND 97% P3 DER ---
MATCM 2 - MAT 79 - SNUTD3dN j
B.07.F.0078 : VARI 3US PRIM 4RY CONTAINME47 ISOL4TIO4 VALVE LEAKS
--- QUAD-CITIES 2 - NOV-DEC 79 - REFJELING FE8 83 j
e.07.F.00S1 : VARI 3US C01741NMENT ISCLATIO1 VALVES FAILED LLRT
--- MONTICELLC - FEB & MAR 83 - REFUELING a.07.F.30S3 : VARI 3US VALVES LEAKED IN EXCESS OF TECH SPECS
--- NATCH 2 - MAR S APR 83 - COLD SHUTDOW1 g,
3.07.F.0089 : PRIMARY CONTLINMENT PENETRATION LEAKS EXCEEDED TECH $*ECS
--- COOPER - MAY 80 - REFUELING
/
pg 5.07.F.3131 : VARIJUS VALVE LEAKS
/
1ILLSTONE 1 - OCT 80 - REFUELI4G gg 5.05.C.0086 : EXCES$1VE PL4f BETWEEN SERVICE WATER CHECK VALVE DISC AND DISC ARM VERMONT YANKEE - AUG 76 - 95% POWER J
9.0S.C.3121 : LEAKING DRAI4 VALVE, C0CKED SPRI1G I4 RECIRC PUMP SEAL PRES $JRE VALVE s
--- COOPER - OCT 77 - REFUELING $NJTDOWN 3.0S.C.3171 : CORR 3DED SERVICE WATER VAULT DRAIN VALVE SPRINGS l
--- 20AD-CITIES 1 - FE8 77 - REFUELI1G f
S.OS.C.018S : CHECK VALVES D4MAGED BY REVERSE PRES $URE - VALVE SPRING I4 PUMP
~
J
--- SRUNSWICK 2 - AUS 79 - 333 P0 DER 5.05.C.3199 : SW PUMP LOCK COLLET DISENS4GED FR31 IMPELLER AND SHAFT
--- MATCH 1 - J AN 80 - 94E POWER 9.03.C.0201 : RWR SW SEDPL4TE DRAIN VALVE FAILED - OPER4 TOR SPRI4G CORR 3DED II
--- QUAD-CITIES 2 - DEC 79 - REFUELING l ~
8.05.C.3210 : RNR SW CHECK VALVE LEAKED CI
~
- - SR3WNS F!RRY 2 - AUG 80 - 91% P0JER 5.07.A.0045 : DIRT ON TIP PURGE CHECK VALVE BALL 04 SEAT IIl
PAGE 10
/
DRE5 DEN 3 - OCT 76 - REFUELING SHUTD0dN I
s.09.8.3055 : LOSSE LEVEL INSTRUMENT 3E1$I1G LINE CHECK' VALVE NUT
--* COOPER - DEC 77 - 75% POJER 3.07.E.0353 : DIRT IN JET PUMP RISER DP INSTRUMENT LINE VALVE RJ l
l
--- QUAD-CITIES 2 - MAR 75 - REFUELING SHUTDOWN 3.07.E.3616 : NO INTERNALS IN INSTRUMENT LINE CHECK VALVE
- SHUT 304N 3.09.F.0016 : CRUD ON INSTRUMENT LINE Cd!CK VALVE POPPET COOPER - APR 78 - REFUELING 5.09.F.0025 : RCP DP INSTRJ4ENT LINE CNECK VALVE LEAKED - POS$1BLE FOREIGN J
MATERIAL 04 SEAT
--- HATCH 1
- JUN'78 - 21 POWER J
l 8.07.F.0026 : DIRT IN INSTRUMENT LINE CdECC VALVE 3UAD-CITIES 1 - FES 77 - REFUELING 3.07.G.3*03 : SERVICE AND INSTRUMENT AIR STSTEM DEPRES$URIIATIONS, STUCC l
CHECK VALVE, REACTOR TRIP 3RUNSWICC 2 - M AY 76 - 15% P3 DER, JUL 76 3.07.G.0121 : 0-RINGS LEAKED IN CONTAIN1ENT OMTGEN SAMPLING VALVES
--- NINE MILE PT 1 - APR 77 - DMSHUTDOWN 5.09.G.3151 : STUCK 8ALL F'0AT CHECK VALVE - W4TER'IN REDIATION 104tT3R g
--- FITIPATRICK - JAN 78 - 971 POWER g
3.07.G.0171 : INADEauATE DRAINING, LEAKING FITT!4GS, CONDENSATION.M3!STURE
/
AND AIR LE4KAGE INTO CAC 10NITORS 3RUNSWICK 1 - MAR 78 - 553-100% POWER --- ORUNSWICK 2 - MAR 78 - APPROXI14TELY 973 P3 DER 7
gg 8.09.G.3178 : LEAKING FITTINGS, STUCK VALVES, CONDENSATION - R4DIAT!01 LND NYOR3 GEN - ORYGEN MONITORS FAILED
--- 3RUNSWICK 1 - MAR & APR 78 - 1303 POWER --- 3RUNSd!CE 2 =
APR 78 - 32% POWER s.09.G.3179 : CRACKED PUMP $ HAFT, WATER IN SAMPLE LINE, STUCK VALVES LOW l
FLOWS IN CAC MONITOR 5 t
3RUNSWICC 1 - APR & MAY 78 - 69% TO 92% POWER --- BRU15 WICK 2 - MAY 78 - 100% P0JER 8.07.G.3234 : INSTR. CASINET EXHAUST CHECK STUCK FROM C3RROSIO4 SAUNSWICC 1 - JUN 79 - 1303 PodER J
3.09.G.3265 : ADS HARD-$EATED CHECK VALVE REPLACED WITN SOFT-SEATED DESIGN,'
$UPP3RTS ADDE 3
--- 3kR$ IN 3ENERAL --- PEACM 80TT3M 2 - JAN 80 - STARTUP (41%
POWER)
PEACH 30TTOM 3 - JAN 30 - $HUTDOWN U
3.11.A.3134 : DG AIR RECEIVER CHECK VALVE RUSTED Closed 3RUN$d!CC 2 - NOV 76 - C3LD SH'JT30WN
)
9.11.A.3343 : DEFECTIVE FUEL CHECK VALVE
--- 3aQNSWICK 2 - OCT 80 - 97K PoJ!R gg j I'
PAGE 11 a
b 9.11.B.3189 : DIRTY LINEAGE, BINDING BREAKER INTERLOCK - RHR VALVE FAILED TO OPEN NATCH 1 - JUL 78 - 693 P3WER 3.11.8.3198 : 8 LOWN LPCI C3NTROL POWER FUSE gj
--* MONTICELLO - NOV 78 - 80% POWER 9.12.A.0060 : DIRT IN SERVICE AIR AND PROCESS VALVES - R/A IN AIR SYSTEM g)
--- PEACH S0TTOM 2 & 3 - SEP 78 - 39% POWER (UNIT 2), 1303 P3WER (UNIT 3)
(I 9.12.A.0066 : R/A IN BREATdING AIR SYSTEM
--- PEACH 90TTOM 2 - SEP F8 --- SWR $ IN GENERAL ((
5.14.8.3188 : SINDING VACUJ1 BREAKERS DRESDEN 2 - DEC 7? - 22% POWER, FEB 78 - SHUTDOW1 5.14.8.0190 : GROWING 0F BJSHINGS - VAC'JUM BREAKER 81NDING
~
--- MONTICELLO - FE8 78 - 103% POWER J
3.14.8.0226 : VACUJM BREAKER DESIGN DEFICIENCY 3RUNSWIC( 1 & 2 - NOV 78-APPR0XIMATELY 100% POWER 3.14.8.3342 : EXCESSIVE C01TAINMENT LEA (AGE THRU AIR COMPRESSOR DISCHAR3E CHECC VALVES
--- VERMONT fANKEE - AUG 78 - SHUTD0dN 8.14.8.0612 : VACUUM RELIEF SYSTEM POTENTIALLY IMOPERABLE IN SWRS JITN 14RC I E II CONTAIN1ENTS - BUTTERFLY VALVE-CONTR3L SEPARATI31, SAFETY gj CLAS$1FICATI3N g
SkRS IN SENER AL - APR 85 --- PE ACH 80TTOM 2 & 3 - 1783 I)
~
5.15.A.0051 : SNUB 3ERS FAILED T3 LOCK UP, 3-RING DETERIORATED
--- WERMONT YANKEE - JUL 76 - REFUELING SHUTDOWN f
~
8.15.A.0089 : DETERIORATED VALVE SEATS AND 0-RINGS, PACKING LEAKS, ETC.
,s 3 DAD-CITIES 2 - JAN 78. REFUELING 3.15.4.0244 : REACTOR SCRA9MED WHEN AIR C01 PRES $3R FAILED J
--- 3ROWNS FERRY 1, 2 & 3 - AUG 78 --- M01TICELLO - FE8 81 -
100% POWER 5.16.C.0646 : THREE VALVES CLOSED - RCIC AND HPCI IN0PERA8LE 3RUNSWICC 2 - SEP 80.SHUTD0d4 J
............. ALL DONE, PRESS < RETURN) (EY TO G3 T3 MEN'J................?
l THERE ARE 138 ARTICLE (5) TO BE DISPLAYED:
19-4UG-1986 13:48:17 8WR - CHECK VALWES - 1951 - 1986 3.04.8.0169 : CRD HCU ACCU 1ULATORS LOST PRES $URE - LEAKING VALVES i
--- SUS 3UEHA1NA 1 - AUG 82 - PRE 0PERATIONAL 1
)
4 8.04.8.0213 : MANUAL SHUTD3dN - CRD ACCJMULATOR INOPERABLE - CR0 PU1P C1ECK ID CPUD SUILDUP VALVE FAILED TO CLOSE
--- LASALLE 1 - JUN 85 - 72% POWER i'
(3 8.05.A.0047 : REACTOR SCRA1 ON EXCESSIVE RCS LEACAGE - RECIRC PUMP SEAL DETERIORATION, LOW WATER LEVEL - WEAR, SLOW RESPONSE OF-FLOW VALVE POSITI3NERS II
--- GRAND GULF 1 - FEB 86 - 62% POWER
I PAGE 12 s
8.05.C.3452 : MSIVS & FW CHECK VALVES FAILED LLRT - SEAT AND DISC WEAR
)
--- ARNOLD - FEB 85 - REFUELING 3.05.C.3455 : MSIV CHECK VALVES SEATED IMPROPERLY QN GRADUALLY DECREASING AIR gj PRESSURE - CHECK VALVES REPLACED SWR $ IN 3ENERAL - APR 85
((P.06.D.345))
3.05.C.0477 : ADS 3PERA81LITY PDTENTIALLY COMPR39ISED - EXCESSIVE AIR LEAKAGE SRVS DEFECTIVE 3WR$ IN 3ENERAL - JUN 86 --- GRA1D GULF 1 - MAR 56 - 80$
()
P3WER 3.05.D.0056 : EXCESS FLOW CHECK VALVE I40PER48LE - ELECTRICAL AND 1ECH44ICAL COMPONENT FAILURES
--- HATCH 1 - JAM S3 - REFUELING.
5.05.D.0065 : UNIDENTIFIED LEAKAGE INTO DRYWELL FL30R DRAIN PU1P - RECIRC PUMP SYPASS VALVE, VENT VALVEr FW CHECK VALVE LEAKAGE
--- ARNOLD - APR 84 - 103I P3WER
.)
3.06.0.0039 : MAIN STEAM LEAKAGE TO AUK BOILER SYSTEM - RELEASE 3ROWNS FERRY 1 - OCT S1 - 97% PodER j
B.06.0.0047 : REACTOR TRIPS - DEH SYSTEM PRESSURE TR ANSIENTS - MS STP ASS-VALVE SEATIN3 IMPROPER, FLUID PRESSURE LOW
--- dNP 14Y $ 4 - 18, 203 POWER 9.06.0.0052 : TUR9INE, REACTOR TRIPS - MIGH M0ISTURE SEPARATOR DRAIN TA1K gj LEVEL - TWO PHASE FLOW SURGE IN TURSINE CROSS-ARQUND PIPING g
SUSQUEHAMNA 1 - OCT 85 - 64% P3WER
=
I) 9.05.E.0149 : FW CHECK VALVES LEAKED
--- HATCH 2 - J AN 81 - REFUELING j-9.05.E.3163 : FW CHECK VALdES LEAKED - DOR 4 SEAL RINGS 03 s
--- 11LLSTONE 1 - OCT 82 - REFUELING 3.05.E.0165 t FW CHECK VALVE SEAT LEAKAGE - RESILIENT SEAL MFG DEFECT
- -- VERMONT TANKEE - MAR S3 - REF'JELING 3.05.E.0166 : FW CHECK VALVE LEAKED - FAILED 8041ET SEAL RING DRESDEN 2 - MAY 83 - 71% POWER s.03.E.0169 : FW CHECK VALVES LEAKED = GAPS ON PERIMETER OF DISC SEAL MATERIAL, ALIGNMENT PR08 LENS
--- LASALLE 1 - NOV 83 - SHUTDOWN 8.06.E.0170 : FW CHECK VALVE LEAKED - SEAL RING SEATING AREA CUT
--- DRESDEN 2 - N3V 53 - 55% POWER j
B.06.E.0172 : FAILURE OF A1CHOR DARLING SOFT SEAT VALVE SEALS - IMPROPE4 VULCAN!!!NG U
--- 3WR'S IN GENERAL - FE3 84; LASALLE 1 - AUGr NOV 53; INDIVIDUAL PLANTS AS LISTED
(
9.0$.E.0174 : F W, COMBUST!3LE GAS CONTR3L & DRYWELL EQUIPMENT DRAI1 VALdES EXCEEDED LLRT LIMITS - MISALIGNME1Ts PITTING, WEAR
--- LASALLE 1 - FEB S4 - SHUTDOWN gg
s PAGE 13 o
8.06.E.0177 : FW SPEED INCtEASE PU1P D41 AGED - DISCHARGE CHECK VALVE FAILED TO CLOSER SE4RINS OVERHEAT
'j
--- NINE MILE PT. 1 - NOV 83 - 76% P3WER 9.05.E.0118 : INCREASED GENERATOR LDAD 5 APRM READINGS - 5TH STAGE FW HEATER Il FL00DED
--- MATCH 1 - JAN 84 - 981 P3WER C) a.06.E.3185 : ADDITION TO VI.E.169 RE: FW CHECK VALVE SEAL LEACAGE
- C3RRECTIVE ACTIONS LASALLE 1 & 2 - NOV $3 - SHUTD3W1
)
S.06.E.3190 : FW CHECK 8 RCIC STEAM HIGH LEAKAGE R4TES - BUSHING WEDGED VALVE OPEN, DISC AND SEAT MISALIGN1ENT, DAMAGE
--- LASALLE 1 - OCT 84 - COL 3 SHUTD0dN 9.06.E.3197 : REACTOR, LOW TURSINE VACUUM TRIPS - HEATER SHELL RELIEF V4LVE J
DISCHARGE RUPTURE DISCS SLEW - RELIEF VALVES LIFTED AFTER LP FW HEATER HIGH DATER LEVEL TRIP - POWER FLUCTUATIONS
--- JNo NOV $4 - 95% POWER j
9.05.E.0199 : FW CdECK VALVE FAILED LLRT - SEAT DETERIORATED
--- LASALLE 1 - MAR 85 - dot SHUTD3W1 9
8.06.E.0200 : FW CHECK VALVE LEAKASE - SEAL RIN3 Wat1 DRESDEN 3 - APR 85 - SHUTDOWN 5.06.E.3213 : STOP CHECK V4LVES IN AUX FW PUMP STEAM SUPPLY FAILED 3N L3W FLod - DISC LSSEMSLY DEGRADED - VIBRLTION El
--- SWR $ IN SENERAL - FE8 86 ((P.06.E.630))
.,y 3.06.E.0218 : RFACTOR SCRAM, CONTAINMENT ISOLATI3NS - FW TRANSIENT DURI1G
/
I PUMP SWAP
- TRANSIENT-INDUCED WATER HAMMER CAUSED DRAIN LINE SREAK i
--- PEACH SOTTOM 2 - DEC $5 - 44% PodER
/.
g) 8.07.A.3139 : RCIC TURBINE EXHAUST CHECC VALVE INTERNALS DAMAGED - TURSINE EXHAUST BLOC (ED, EXHAUST DIAPHRAGM RUPTURED
--- NATCH 2 - NOV 81 - 961 POWER 9.07.A.0149 : RCIC TURBINE EXHAUST CHECC VALVES DAMAGED DURING TESTINS 3N LUX l
STEAM - LOW STEAM FLOW
--- SHOREHAM 1 - FEB S2 - STARTUP TESTING 8.07.A.0150
.R CI C & HPCI TUR8!NE EXHAUST CHECK VALVES DAMAGED
.L3W STEAM FLOW SUSQUEHANNA 1 & 2 - APR 81 - PRE-0P TESTING 9.07.A.0152 : SWING CHECK WALVE PROBLEMS
--- SWR'S IN GENERAL - JU1 82 j
a.07.A.0153 : RCIC TURBINE EXHAUST CHECC VALVE 04M4GE EXPERIENCE REVIEW!D HPCI TUR8INE EXHAUST CHECC VALVES SUSPECT II 3WR'S IN GENERAL - JUL 82 9.07.A.0162 : RCIC TESTA 9LE CHECK VALVE, EQUALIIING VALVE FAILURES TO CLOSE -
' ()
INDICATION PROBLEM $r SPRING TENSID4 LDJUSTMENTS NEEDE)'
--- LAS4LLE 1 - JUL $2 - 2X POWER (PRE-OP); AUG 82 - 3, 95 P3WER (PRE-OP)I SEPT 82 - 2X PodER (PRE-3P) gg
t PAGE 14 3
S.07.8.0048 : SBLC CHECK V4LVE LEAKED - SEAT REW3R(ED
--- MILLSTONE 1 - OCT 82 - REFUELING
)
9.07.C.3100 : CONT 4CTOR FAILED TO ACTUATE CSS VALVE OPE 4 CIRCUITRY 3ROWNS FERRY 2
.9AR $1 - 99% P0JER 41 3.07.C.0117 : CORE SPRAY AND RWCU WELD CRACKS FOUN3 - IGSCC
--- 3ROWNS FERRY 1 - AUG S3 - REFUELING
()
3.07.C.0118 : HPCS CHECK VALVE FAILED - BYPASS VALVE FAILED - ACTU4 TOR SPRING LODSE C
--- LASALLE - OCT 82 & JUM 83 - 23 & 42% POWER a.07.C.3132 : HPCS CHECK VALVES FAILED TO PREVENT REVERSE FLOW FROM P'J4* -
SCORED SEATING SURFACES - FM
'.ASALLE 1 - JJL S4 - 97% POWER B.07.D.3291 : SUPPRESSION CHAM 3ER LEVEL HI - RNR PUMP DISCHARGE CHECK V4LVE FAILED 70 SELT, OPERATOR FAILED T3 REALIIE ABNORMAL CDNDITION 3RUNSWICK 1 - DEC 81 - 05 POWER CSTARTUP)
)
8.07.D.0297 : RHR PUMP DISCHARGE CHECK VALVE DISC LODSE - DESIGN PR3BLE1 SUSPECTED J
--- FITIPATRICK - MAR 82 - 12% POJER (STARTUP) a.07.D.3300 : LPCI CHECK V4LVE DRAIN /EQUALIIING LINE WELD CRACK - INADE3UATE PIPE SUPPORT DRESDEN 3 - APR 82 - REFUELING 9.07.D.3315 : RELEASE TO INTAKE STRUCTURE - UNDETECTED LEAK PATH THROUG4 HP SW SYSTEM FR31 RHR HK
--- PEACH BOTTON 3 - NOV S2 - 89% POWER
,/
g 5.07.0.3321 : RHR PUMP DISCHARGE CHECK WALVE FAILED - DISC HANGER RET 4111NG CAP SCREWS MISSING, SHEARED - POSSIBLE VISRATION DAM 4SE
/
bN FEB 83 - 130% P3JER
--- FITIPATRICK
.s 9.07.0.3343 : RHR PUMP DISCHARGE CHECK VALVE WOULD N0! SEAT PROPERLY.- DISC J
MISALIGNED, DISC d!NGE PIN BENT SUSQUEHANNA 1 - AUG 83 B.07.D.0348 : RHR PUHP DISCHARGE CHECK VALVE FAILED TO SEAT PROPERLY - DISC MISALIGNED, DISC HINGE PIM BENT j
--- SUSQUEHANNA 1 - NOV S3 - 100% P0JER s
a.07.D.0372 : RHR CHECK VALVE LEAKED EXCESSIVELY - IMPR3PER SEATINS -
MISALIGNED HINGE ARM J
SROWNS FERRY 3 - OCT $4 - REFUELING 8.07.D.3375 : LPCI CHECK V4LVE FAILED T3 HDLD PRESSURE - TIGHT PACCINS, SH4FT CORR 35!ON k
--- LASALLE 1 - OCT 84 - COLD SHUTD04N U
3.07.0.3406 : RCS NOT ISOL4BLE FROM LP COOLANT SYSTEMS - PCIV LEAKASE
--- SWRS IN SENERAL
- JUN 86 -*- AR10LD - MAR 86 - SHUTD3WN ---
PILGRIM - APR 86 - 93% POWER
()
8.07.E.0235 : HPCI STEAM EKHAUST CHECK WALVE DISC PIN F4ILURE - DISC LO3GED IN PIPING 90NTICELL3 - MAR 81 REFUELINS
~
PAGE 15 1
8.37.E.3237 : HPCI TURBINE EXHAUST CHECK VALVE LEACED PLST SEATING RIMG THRE4DS 90NT! CELLO 1 - APR 81 - REFUELINS 8.07.E.0252 : ADDITION TO VII.E.245 RE: HPCI GLAND SEAL CONDENSER HEAD SASKET BLOWN 3ROWNS FERRY 2 - SEP 51 - SHUTD0dN (FDLLOWING SCRLM)
()
9.37.E.0275 : HP I4JECTION STOP CHECK V4LVES DISC STUCK SHUT - INCORRECT SEAT ANGLE
()
SWR $ IN GENERAL - AUG 82 ((P.07.A.3753) 8.07.E.3284 : HPCI TUR8INE EXHAdST FAN CHECK VALVE LEAKED - WEAR - 1EFECTIVE VALVE SEAT
--- 3RESDEN 3 - FE8 82 - REFUELING 3.07.E.3286 : HPCI PUMP TRIPPED DURING SCRAM REC 3VERY - CONTROL CIRCUITS WETTED - GLAND SEAL CONDE4SER GASKET FAILED - OVERPRESSURE
--- PILGRIM - AUG 82 - STARTUP 8.07.E.0311 : ALAR 95 DUAINS HPCI SURVEILLANCE - VALVE PROBLEMS, C049U1ICATION ERRORS
~
PILGRIM - SEP 83 - 96% P3WER 8.07.I.0335 : HPCI TURBINE EXHAUST RUPTJRE DISCS FAILED - OVERPRESSURE FR01 JATER IN EXH4UST LINE, SLUGGISH CHECC VALVES PEACH S3TTOM 3 - OCT $3, JAN 84-- 17, 100% POWER 3.07.E.3354 : HPCI IS0LATI3N - EXHAUST LINE DIAPHR4GM RUPTURED - F1 SLOCKED
~
Il e
DRAIN LINE - PLANT SHUTD0dN - CHECK VALVE FAILURE - DEFECTIVE 7
SEAL, DISC OJT OF ADJUSTMENT, LOOSE NUTS, NICKED SEAT f
C) :
--- HATCH 1 - JAN 85 - 90% P3WER 3.07.E.3360 : HPCI, RCIC TJRBINE VALVES NOT LOCAL LEAK RATE TESTED PR3PERLY -
/
E LEAKING BONNETS - PERSONNEL, PROCEDURAL ERR 3R$
--- 3ROWNS FERRY 1, 2, 3 - M4R 85 - 70% P3WER, REFUELINS, COLD
?
SHUTDOWN S.07.E.0377 : HPCI IN0PERA3LE - TESTABLE CHECK INJECTION VALVE DISC P3SITION INDICATING SJITCH 00T 0F A D J 'J S TM E N T J
COOPER - NOV 85 - 7X POWER (STARTUP)
S.07.E.0385 : HPCI CHECK V4LVE INTERNALS DAMAGED, MISSI4G - CAP SCREWS 10T LOCKED IN PLLCE SHOREHAM - NOV 85 - C3LD SHUTDOW1 s
8.07.F.0118 : PRIMLRY CONTLINMENT VALVE LEACS - MORMAL DEAR
--- COOPER - MAY 81 - REFUELING J
8.07.F.0121 : VALVE LEAKS - FOREIGN MATERIAL /0PERATOR ADJUSTMENT / FAILED SEAL /
{
~
CORR 3SION
--- NINE MILE PT 1
.APR S MAY 81 - REFUELING 5.07.F.0136 : DRAG 3N EXCESS FLOW CHECK VALVES FAILED TO FUNCTION IF LINE BREA< WAS NOT NEAR THE VALVE
()
LASALLE 1 & 2 - SEP 81 - CONSTRUCTION PHASE 5.07.F.0152 : VARI 3US VALVE LEAKS - WEARS FOREIGN 1ATERIAL gg FITIPATRICK - NOV 81 - REFUELING
.s
..o m..-
t PAGE 16 J'
3.07.F.3169 : VARI 3US ISOLATION VALVES FOU1D LEAK!1G DUt!NG LLRT - IMPR3PER j
SEATING DUE To DA9 AGED C39P01ENTSs CRU3 8UILDUP
--- PEACH 80ff0M 2 - JUN $2 - REFJELING 3.07.F.0193 : VALVE LEAKAGE FOUND DURING LLRT HATCH 2 - FE8 82 - REFUELING S.07.F.0196 : EXCESS FLOW CHECK VALVE FAILED TO SEAT - CRUD 3RUNSWICC 1 - OCT 82 - SHUTDod4
(
8.37.F.3210 : LLRT LEAKASE - NORMAL WEAR
--- FITIPATRICK - JUN 83 - REFUELING S.07.F.0211 : CIVS LEAKED EXCESSIVELY - VARIOUS CAUSES
~
--- LASALLE 1 - AUG $3 - 3% POWER (54UTD0dN) 3.07.F.0215 : DFP RAN BACKJARDS - ENGINE DAMAGED - CHUNC 0F WOOD L3DGED IN DISCHARGE CHECK VALVE
--- LASALLE 1 & 2 - DEC 53 - DE POWER (STARTUP PREDPERATI3NAL) s 3.07.F.0227 : VARI 3US VALVES LEAKED EXCESSIVELY
--- MONTICELLO - SEP 8 OCT 82 - REFUELING 8.07.F.0243 : LLRT LEAKAGE IN EXCESS OF TECH SPEC - NOR9AL WEAR - WALVE MODIFICATIONS J
--- QUAD-CITIES 1 - MAR 84 - COLD SHJTDOW4 3.07.F.3256 : RHR, MSL, TIP, HPC!s LPC!r CS & OTHER CIVS FAILED LEAC RATE g,
TESTING - NORMAL WEAR
,p
--- HATCH 1
- OCT-DEC 84 - COLD SHUTDOWN O
S.07.F.3257 : EXCESSIVE CIW LEAKAGE RATES
--- PEACH 80TTOM 2 - JAN SS - REFUELING f
b>
S.07.F.3266 : PCIVS FAILED LEAK RATE TESTS - VARIOUS CA35ES
--- PILGRIM - DEC 83 - REFUELING 4
J 8.07.F.3271 : CORRDSION FAILURES OF 410 SS VALVE STEMS - IMPROPER HEAT TREATMENT
--- SWR $ IN SENERAL - JUL 85 --- BRUNSWICK 2 - AUG 82'---
SR3dNS FERRY 3 - FE3 84
+
3.07.F.0291 : VARI 3US CIVS FAILED LLRT - NORMAL WE AR, INTERNALS MIS $!NG FROM
)
3YPASS VALVE i
--- dATCH 2 - APR 85 - REFUELING 8.07.F.0304 : ESF ACTUATIONS, CHARGING WATER SPILLAGE - DRYWELL PRESSURE SWITCHES SHORTED - M0!$TURE IN JUNCTION 83XES FR3M E ARLIER INADVERTENT FIRE PROTECTION ACTUATION k
/
3R0dNS FERRY 1 - APR-MAY 86 - C3LD SHUTDOWN 3.07.F.33058: VALVES EXCEEDED LLRT LIMITS'- NORMAL WEAR 3UTs TORQUE SWITCH N
ADJUSTMENTS 1 FAILURE, VALVE DESIG1
--- PEACH BOTTOM 3 - SEP SS - SHUTD0dN f)
B.07.F.0306t 58GT FILTER TRAIN IN3PERA3LE - CHARC3AL FILTER BED W!T -
LEAKING DELUSE VALVE 1 ASKED Bf LEAKING DRIP CHECK VALVE
--- HATCH 1 1 2 - NOV 85-94R 86 - REFUELING, 85% P3WER I
.. ~
- ~ ~ -
_n...
n
-. -. ~ _ _ - - - _ _. - - - -.
- _ _ -. ~.,
a PAGE 17 l
1, f.'
5.08.C.3234 : FAILED SHAFT PAC (ING AND S' tic WATER SUPPLY CHECK VALVE FOR RMR SW P 'J M P
--- MONTICELLO - MAR 81 - 948 POWER 5.DS.C.3245 : RNR SW PUMP 3!$CH4RGE CHECK VALVE FAILED - WEAR M,---
HATCH 1 - JUL 81 - 935 P3WEA
- > l 5.08.C.3281 : RHR SW PUMP 3!$ CHARGE CHECK WALVE ST*JCK OPEN - DISC NJT L30$E, COTTER PIN MISSING
()
i
--- FITZPATRICE - MAY 82 - 130% P3WER l
8.05.C.3285 RHR SW PUMP 4EAD LOW, DISCHARGE CHECC VALVE LEAKAGE - WEnt, SET SCREJ ER0DED 4RNOLD - MAY 82 - 90% P0JER, AUG 82 - 51% P0JER 5.05.C.0288 : RHR SW PUMP 1EAD AND FLOW LOW - CHECK VALVE STUCC OPEN IN DISCHARGE FL3W PATH
--- HATCH 2 - AUG 82 - 965 P3WER 3.05.C.3308 : ESV CHECK VALVE INTERNALS WORN, MISSING - ACCELERATED R3N T!1E
--- SUSQUEHANN4 1 - APR 83 - COLD SH3TDOWN (STARTUP TEST!1G) 4 8.05.C.3311 : POTENTIAL DA1 AGE TO REDUNDANT SAFETY EQUIPMENT FROM 34C(FLOW THROJGH EQUIPMENT AND FLO3R DRAIN SYSTEMS J
i j
3WR$ IN SENERAL - JUL 83 ((P.38.C.135))
5.0$.C.0331 : RHR$J PUMP DISCHARGE CHECC V4LVE F4ILED T3 SEAT PROPERLf.- DISC STUD THREAD JEAR MONTICELLO - DEC $3 - 79% POWER tj 8.05.C.3334 : SW PJMP VALVE FAILURES
--- HATCH 2 1 1 - OCT 82, OCT 83 - 55, 70E POWER
/
()
5.05.C.3335 : ADDITION To VIII.C.308 RE: LOST P4CIFIC CHECK VALVE MUT
--- SUSQUEHANNA 1 - APR 83 - COLD SH9fD0d1 (STARTUP TEST!1G)
,r 8.05.C.0338 : RHR SW VAULT SUMP DISCHARGE CHECK VALVE LEAKAGE - FM BUILSUP 3U40-CITIES 2 - DEC $3 - REFUELING j
l 1
5.05.C.0339 : HPSWP DISCHARGE CHECK VALVE STUCK DP!N - VALVE DISC PIN, 4RM INTERNAL WEAR 1
--- PEACH 80TTOM 2 - DEC 83 - 97% P0JER 5.05.C.3346 : POTE1TIAL LOSS OF ESW - F4ILURE OF BfPASS OR CHECK V4LVES -
FLOJ DETECTOR INSTALLED SUSQUEHANNA 1 - OCT 82 - SHUTD3W1 (PRE 0P TESTING) 1 5.08.C.0358 : PLANT SHUTD0JN - DRYWELL FLO3R DRAIN SUMP PUMPS FAILED, SJMPS J
6 DVERFLOWED - BROCEN DRIVE COUPLINES, FAULTY FW CHECK WALVE HINGE PIN
--- HATCH 2 - DEC 84 - 99E P3WER k
J 8.09.C.0134 : RCS INSTRUMENTATION LINE EFCWS FAILED'TO CHECK FLOW IMITI4LLY j
--- HATCH 2 MAY 83 REFUELING
$)
8.09.E.0634 : WORN INTERNALS ON INSTRUMENT LINE EXCESS FLOW CHECK WALVES l
--- BROWNS FERRY 3 - JAN 81 - REFUELING SHUTDOWN f) 3
)
' INSTRUMENT LINE EXCESS FL3W CHECK VALVES FAILED C
5.09.E.0681 :
3ROWNS FERRY 1.- SEP 81 - REFUELING
{
PAGE 18 a.09.E.3866 : EXCESS FLOW CHECC VALVE 0*ERABILITY NOT VERIFIABLE WITH3UT REM 0/ING VESSEL HEAD - DESIGN DEFICIENCY
--- SUSQUEHANNA 1 AUG S3 - 0% P3dEt (PREOP TESTING)
B.09.E.3900 : MSL INSTRUMENT LINE VALVES INOPERABLE - INSUFFICIENT FL3W, FAULTY POSITION Sd!TCHES ARNCLD - FEB S3 - SHUTD0dN (3
3.09.E.1035 : RPS, ATWS TRIP SIGNALS - INCORRECT EXCESS ' FLOW CHECK VALVE 4
OPENED DURIN3 TEST - PERSDNNEL ERR 3R
--- 2UAD-CITIES 1 - MAR SS - REFUELING
()
5.09.r 3376 : RCLDASS CHECC VALVES STUCC SNUT - FOREIGN MATERIAL
- u
--- PILGRIM - JUN 81 - 103% POWER 8.37.G.0391 : INSTRUMENT LINE EXCESS FL3W CHECK VALVES FAILED - VALVES REPLACED
--- NINE MILE PT 1 - JUN S1 - REFUELING B.09.G.0551 s' ADS INSTRUME47 PRESSURE L3W - AIR DRYER UNIT $0LEN0!D* VAL /ES STUCC, NITR03EN SUPPLY CHECK VALVE LEAKIN3 - MODIFICAT!3NS REQUIRED
--- LASALLE 1 - DEC 82 - 19% POWER (STARTUP TESTING)
J S.11.A.3412 : DGS TRIPPED 3N HIGH COOLING JATER TEMP - CAUSE UNKN0J1, C30 LING WATER PUMP DISCHARGE CHECC VALVE SUBSEQUENTLY REPLACED
--- DRESDEN 3 - OCT & NOV 81 - 66 1 50% P3WER 9.11.A.0430 : ADDITION To XI.A.412 RE: DGS TRIPPED ON HIGH COOLING JATER TEMP
') b
--- DRESDEN 2 8 3 - DEC 81 --- BWR$ IN GENERAL ((
j 3 11.A.0478 : EMER3ENCY DG STARTED SLOJ - FAILED FUEL DIL DAY TANC RETU4N Ik LINE CHECK VALVE
/
--- VERMONT fANKEE - OCT 52 - 1005.P3WER gg S.11.A.0490 : CHECC VALVE FAILURES IN D3 RAW WATER C00 LING SYSTE15 = INTERNAL DISASSEMBLY ABRASION, C3RR3510N
--- SWRS IN 3ENERAL - MAR 83 --- DRESDEN 2 & 3 --- QUAD-CITIES 1 l
& 2 ((
S.11.A.0521 : DG MECHANICAL PR05LE15 REVIEJED SWR $ IN G E N E R AL - AUG 8 3 --- S RUN S WIC C 1 --- CLI1T01 1. 8 2 --
- DRESDEN 3 --- FERMI 2 --- 2UAD-CITIES 2 --- SH3REHA1 SUSQUEHANNA 1 83 --- 0THERS AS NOTED IN J
a.11.A.0566 : DG PROBLEMS START FAILURE, VOLTA 3E COULD NOT BE MANUALLY ADJUSTED - FJEL 0!L RETUR1 LINE CHECC VALVE LEAKINGr L O C A'.
M 3 D E SELECT SWITC1 CONTACTS DIRTT
--- HATCH 2 - OCT 83 - 103% POWER
}t J
B.11.A.3641 1 DG MALFUNCTI34 - PIN BROKE OFF AIR START CHECK VALVE 5 L30GED 5
IN A1 AIR INLET VALVE SRAND GU' F 1 - MAR 85 - SHUTD3JN O
=
9.11.B.0526 : LOSP - TRANSFORMER FAILURE - CABLE FAULT - INSULAT!04 DEGRADED -
AGE, ENVIRONMENTAL DAMA3E f3 k
--* 3WR$ IN 3ENERAL - JUN 86 (CP.11.3.861))
6)f VENT LINE SLOCCED t.12.A.0052 : RADWASTE SERVICE AIR PIPING CLOGGED
J PAGE 19 UNM04!TORED RELEASE 3YSTER CREEK - JUN 82 - 84% PodER 9.12.A.3108.
UNM0NITORED CST $JMP DISCHARGE - RADIOCHERISTRY NOT INF3R1ED BEFORE MAINTENANCE BEGAN (f
SHOREHAM - APR 86 - C3LD SHUTD3d1 a.14.A.0161 : EXCESSIVE PENETRATION AND VALVE LEAKAGE - VALVES WDRN, DA14GED -
()
VARIOUS CAUSES QUAD-CITIES 2 - MAR 85 - REFUELING C>
9.14.8.3347 : ISCLATION CHECK VALVE LEACED WORN 3-RING NINE MILE PT 1 - MAR 81 - REFUELING 1'~
5.14.8.3458 : REACTOR SUILDING PRESSURE L0d - VENT SYSTEM SUPPLY FA1 FAILED -
BLADES BR0(E'3FF
--- LASALLE 1 - OCT 82 - 32% POWER (STARTUP TESTING) 8.14.8.0572 : EXCESS FLOW CHECC VALVE FAILURES - VALVES INSTALLED I4 dR3NG LINES - INCORRECT MFR NAME TAGS FOLLOWING MODIFICATION
--- JNP FEB 84 - PREOP TESTIN3 3.15.A.0246 : CHECC VALVE FAILURES - NF3 DEFECTS
--- 3WR$ IN GENERAL - DEC 81 ((P.15.3733) 8.15.A.3258 : ADDITION TO XV.244 RE: AIR C3MPRES$0R FAILED, REACTOR SCRAMMED
--- MONTICELLO - FEB 81 - 103% POWER 5.15.A.3300 : POTENTIAL FOR PIPING SYSTEM S PIPE SUPPORT DAMAGE - PIPE I
BENDING - TE1P DIFFERENCES INDUCED BY STRATIFIED FLOW
,l
--- SWR $ IN 3ENERAL - DEC 84 --- dNP AUG 84 - 1% P0 DER (STARTUP) ((
/
g) 8.15.A.0302 : CONTAMINATION OF SREATHING AIR SYSTE4S - AIR COMPRESS 3RS FAILED, RADI0 ACTIVE GAS S FM IN AIR SURGE AT SYSTEM STARTUP
/'
kI 3ROWNS FERRY 1, 2 8 3 - SEP 84 - VARIOUS POWER LEVELS *---
s SWRS IN GENERAL - JAN 85 ((P.15.455))
l 9.15.A.0310 : LIMITORQUE ACTUATOR WORM SHAFT GEAR FAILURES - IMPACT L3431NG DF CLUTCH MECHANISM DURIN3 M3DE CHANGES AT HIGH SPEED 3WRS IN 3ENERAL - AUG 85 --- 3THER UNITS AS LISTED IN J
ARTICLE ((
8.16.C.3724 : VENT ACCUMULATOR CHECC VALVES LEAKED - INADEQUATE PM MONTICELLO - MAY 81 - REFUELIN3
/,
5.16.C.3748 : 10CFR50.44 AND TMI LESSONS LEARNED REQUIREMENTS CONFUSED /AEP3RTING REQUIREMENTS OVERLOOKED - MANAGEME1T PROBLEM
--- PILGRIM - JUN 81 - 100% POWER 8.16.C.3757 : RdCU SYSTEM RETURNED TO SERVICE WITH ISOLATION VALVE IN3P!RABLE l
J TECH SPEC REQUIREMENTS TO 3E CHANGED l
90NTICELLO - JUN 81 - 103% POWER 0
i 5.15.C.0834 : SURVEILLANCE MISSED - ADMIN AND PR3CEDURAL CONTROL BREACD3dN I
3YSTER CREEK - MAY 81 TO FEB 82 - VARIOUS c3 3.16.C.0935 : NONIDENTICAL REPLACEMENT PARTS - RECURRING PROBLEM
--- SWR $ IN 3ENER AL - FES 83 ((P.15.C.1845))
gg
s PAGE 23 L
e 3.16.C.1073 : EECW CHECK VALVE INSTALLED SACKWARDS SROWNS FERRY 1 - OCT S3 - REFUELING j j 9.15.C.1109 : REACTOR WATER LEVEL DECREASED - LPCI CHECK VALVE STUCK 3 PEN, PACKING GLAN) TIGHT - IMPROPER MAINTENANCE
. gj.
4
--- LASALLE 1 - SEP 83 - COLD SHUTD0JN S.15.C.1241 : EXCESS FLOW CNECK VALVES NO TESTED - ADMIM, PER$0NNELl ERR 3R
()
--- 3YSTER' CREEK - JUN 84 - REFUELING 3.15.C.1271 : C3RE SPRAY PIPING OVERPRES$URIZED, PAINT DA1 AGED - W3RKERS gg.
CONTAMINATED.- OPERATOR TESTING,' MAINTENANCE ERRORS i
--- 3ROWNS FERRY 1 - AUG $4 - 1035 P3WER i
5.16.C.1319 : INADVERTENT SAFETY SYSTEM STARTS DURING TESTING - INSTRJMENTS NOT ISOLAT!D
--- 3ROWNS FERRY 3 - SEP $4 - REFUELING 3.16.C.1437 : EXCES$1VE DRYWELL LEAKAGE - TEMPORARY NOSE SLIPPED OFF FITTI1G -
VALVE SOCKET WELD CRACK FOUND LATER - VIBRATION
--- SROWNS FERRY 1 - JAN 95 - 0% POWER (STARTUP) 8.15.C.1461 : POTENTIAL V!3LATION OF PRIMARY CONTAINMENT - OPEN FLOJ CHECK l
VALVE BYPASS VALVE - PERSONNEL ERROR
--.dNP FEB 85 - 100E PodER J
3.16.C.1506 : EXCESS FL0d CHECK VALVES INOPERABLE DURINS STARTUP - SYPASS VALVES LEFT 3 PEN LIMERICK 1 - APR 85 - HOT SHUTD0dN gj 8.16.C.1545 : REACTOR, TUR31NE TRIP ON' LOW CONDE15ER VACUUM - BREAEEi INADVERTENTLY TRIPPED, LUBE d4TER PU1P CHECK VALVE LEAKED
/
--- SRAND GULF 1 - MAY 85 - 741 P3dE4 9.16.C.1580 : REACTOR HP SCRAM SIGNAL - CRD FL3d INCREASED WITH R03T VA*.VE
/
CLOSED - CHECK VALVE LEAKAGE gg PEACH 80TTOM 2 - MAY 85 - COLD SHUTD0dN
............. ALL DONE, PRESS < RETURN > KEY TO GO T3 MENU.................?:
JTHERE ARE 81 ARTICLE (S) 70 BE DISPLAYED:
19-AUG-198 6 13: 5 5 : 23 PWR - CHECK VALVES - 1960-1978 l
P.05.F.0004 : PR05'LEMS WITH VALVE OPERATINS SYSTEM
~~~ SHIPPINGPORT - 1760'S CHECK VALVE INTERNALS REPLACED WITH NORE DURABLE MATERIALS P.05.F.0016 :
i YANKEE R3WE - AUG 74 - REFUELING SHUTD0dN P.06.A.0014 : CONTROL & ST3P VALVE PR08LEMS CONN YANCEE (HADDAM NECK) - 1967, 68, 69 i
1 i
P.06.D.3025 : VALVE STEM C3RROSION'& BINDING 1767, 68
--- CONN. YANKEE (HADDAM NECC) g P.05.E.0022 : CHEC(VALVE GASKET FAILURE - JATER HAMMER, 8 LOWDOWN, $4FEff.
INJECTICN, PIPE SUPP3RT DAMAGE g3 SURRY - 3CT 72 (P3WER ESCALATION TESTING)
~
P.06.E.0029 : CASTING DEFECT IN CHECK VALVE
' gg s!,
SuRRv 2 - MAY Ts
=
8 PAGE 21
)
P.05.E.0055 : HIGH ENERGY LINE MODIFICATIONS NEGATED AUX. FW REDUNDANCY SURRY 1 1 2 - AUG 74 P.06.E.0064 : RELEASE THROJGH FEED PUMP SEALS
--- PT. BEAC4 - MAR 75 - COLD SHUTD0dN P.05.E.0106 : Fd CHECE VALVE GASKET LEAXED
--- PALISADES - PRIOR TO 1974 P.06.E.0119 : EXCESSIVE FW FLOW CAUSED DATER HAMMER - $1UBBERS & PIPE SUPP3RTS DAMAGED
()
COOK 1 - JAN 76 & MAR 77 - SHUTD3WN P.06.E.3128 : CHECC VALVE FAILED - BURR ON HINGE - PUMP CASING GASKET REACT 3R TRIPPED RUPTURED ROSINSON 2 - AUG 77 - 103% POWER J
P.06.E.0142 : FAULTY VALVE TORQUE SWITCN REPLACED
--- DAVIS-BESSE 1 - OCT 77 (POWER ESCALATION TESTING) - H3T STANDBY P.06.E.0153 : FW REGULATINS VALVE FAILED OPEN - TU31NG FITTING BROKE
.RCP SEALS DAMAGED FT. CALH3UN 1 - APR 74 - 90% P3WER P.05.E.3155 : FW CHECC VALVE PIN BR3KE SAN ONOFRE 1 - APR 76 - STARTUP P.05.E.0162 : LOOSE B'ONNET BOLTS, SCRATCHED GASCET - AUX FW VALVES LEACED
,,II
~
--- DAVIS-3 ESSE - JUL 77 - SHUTDOWN = APR 78 - 70% P3 DER P.03.E.0171 : CONDENSATE STORAGE TANK LEVEL LOW - VALVING ERROR, CHECC #ALVE I)
STUCC
--- TROJAN - JUL 78 - HOT STANDBY
/
g P.05.E.0155 : AUX FW CHECC VALVE ORIENTED dRONG - PLATE HINGE LUGS MISSING -
PLATE RECOVERED FROM SG
--- FT. CALH3UN 1 - NOV 75 - REFUELING P.05.E.0353 : AUX FW PUMP PROBLEMS - VALVE LEACS DEFECTIVE FUSE HOLDER ROBINSON 2 - DEC 77s JUN S1 - 103% P0 DER, STARTING 'P J
P.06.F.0069 : WDRN. CORR 0DED PUMPS EXTENSIVE REPAIRS - VALVES INSTALLED IN LUBE WATER PIPING 1977 ST. LUCIE 1 P.07.A.0013 : CHECC VALVE ASSE1 SLED INC3RRECTLY s
PALISADES - MAY 72 P.07.4.0020 : INDICATIONS IN VALVES k
J
--- 1AINE YANCEE - SUMMER 1972 EAFTER HOT FUNCTI3NAL TESTING)
P.37.A.0025 : WELD SLAG UNDER VALVE SEAT - SI TANK CONCENTRATION L3W OI
--- MAINE YANKEE - DEC T2 P.07.A.0037 : CHECC VALVE SOFT SEAT' FAILURES C)
--- TuRcET PT 4 - MAT 73 P.07.A.0360 : CRACC INDICATIONS IN VELAN CHECC VALVES 8
T s
PAGE 22 l
--- CALVERT CLIFFS 1 - SEP 73 (FUNCTIONAL TESTING)
P.07.A.0063 : CHECC VALVE LEAKAGE - ACCJMULATOR DILUTIO4 31NNA - SEP 76 P.07.A.3055 : CHECK VALVES DID' NOT SEAT, PROCEDURAL PROSLEM - 80R04 CONCENTRATI04 LOW
--- SURRY 1 - AUG 75 - HOT STAND 8Y
()
P.07.4 3105 : LEAKING CHECK VALVE - ACCUMULATOR SDRON L3W 405INSON 2 - JAN 76 - 103% P3WER
()
P.07.A.0114 : LEAKING CHECC VALVES C AUSED 50RON DILUTION IN SI ACCUMULATOR SURRY 1 - J UL 76 - 103% POWER
)
P.07.A.0120 : B0404 DILUTI3N IN SI ACCU 1ULATOR CAUSED ST' LEAKING CHECK #ALVES SURRY 2 - AUG 76 - 103% POWER 3
P.07.A.0122 : BACK LEAKAGE IN ACCUMULAT3R DISCHARGE LINE CHECK VALVES IION 1 - JUN 76 - 50% POWER j
P.07.A.3126 : WRON3 $!!E GASKET INSTALLED IN ACCUM3LATOR CHECK VALVE IION 2 - OCT 75 - 30% P0 DER
)
P.07.A.0143 : IMPR3PER SEATING OF BIT DUTLET VALVES CAUSED DILUTION OF 5AT'S IION 2 - MAR 77 - 50% P0 DER, APR 77 - 48I P0JER
)
P.07.A.0172 : BOLTS & BRACKET FOR CHECK VALVE DISC MISSING INDIAN PT. 2 - MAY 78 (L3d POJER PHYSICS TESTS) gj P.07.A.3175 : TILTING DISC CHECK VALVE INSTALLED I4 WROMG ATTITUDE
--- SAN ON3FRE 1 - MAY 78 - APPRDXIMATELY 55X P0JER
/
()
P.07.A.0182 50 ROM CONCENTRATION LOW - SI TANK CHECK VALVES LEAKED
--- CALVERT CLIFFS 2 - SEP 78 - 985 POWER
,f G
s
~
P.07.A.0187 : STOP CHECK DISC SPRING JAMMED - SPRING RETAINER ADDED
--- ARKANSAS ONE 2 - JUL 78 - PRE 07 TCSTING J
P.07.A.0225 : $1 TANK 80R31 CONCENTRATI3N LOW - CHECK VALVE LEAKED
--- MILLSTONE 2 - APR 77 - 99% POWER P.07.8.0012 : CRAC(ED ROCK 3 HAFTS AND CRACKED 80DY:!N CHECK VALVES SURRY 1 S 2 - NOV 74 s
P.07.s.0053 : CTS 40!!LES LEAKED - VALVES NOT SHUT YIGHTLY - CHECK VALVES
-INSTALLE0 3ACKWARDS
--- C00K 2 - NOV 78 - SHUTDOJN J
P.07.E.0020 : NYDR3 GEN DILJTION BL0 DER CHECK VALVES. STUCK l
--- DAVIS-5 ESSE 1 - APR 75 - $8% POWER t
P.07.E.0026 : CONTAINMENT v4COUM RELIEF VALVE HAD DEFECTIVE WELDS L335E FLAN 3E BOLT - DIRTY VALVE SEAT O
--- DAVIS-BESSE 1 - MAY 75 - SHUTDOW1 P.07.E.3028 : CHECK VALVES LEAKED - DISCS S SEATS CLEANEDs LAPPED C
DAVIS-BESSE 1 - JUN 78 - REFUELING 6hl P.07.E.0038 : ADDITION TO ITEM VII.E.26 RE: LEAKIN3 CONTAINMENT VACUU1 RELIEF
PAGE 23 VALVE
--- DAVIS-8 ESSE'1 - MAY & JU1 78 - REFUELING P 07.E.0237 : CORa3SION FAILURES OF 410 $$ VALVE STEMS - IMPROPER HEAT TREATMENT gj PWAS IN SENERAL - JUL 85 --- DC01EE 1 - DEC 71 --- FARLEY 1 -
FEB 84 ((9.37.F.271))
P.03.A'3035 : KER0 TEST V4LVES INADEQUATE IN SAMPLI4G. SYSTEM
--- TURCEY PT. 3 & 4 - IST 1/2 0F 1973 P.03.4.3153 : SODIUM SULFATE ACID SYSTEM C3MPONENTS CORRODED, PUMP SEAL WORN
--- THREE MILE IS. 1 - 1975 P.05.A.0230 : CHECC VALVE LEAKED - REL!tF VALVE 3, FLOW TRANSMITTERS DAMAGED -
CHECC VALVE REWORCED DAVIS-3 ESSE 1 - JUL 77.(AFTER INITIAL FUEL LOADI45) - HOT STAND 8Y P.38.A.3262 : MAKEUP SYSTE1 VALVES LEACED R/4 GAS DURING VALVE INSTALLATION J
ON NITROGEN 1EADER - RELEASE CRYSTAL RIVER 3 - FEB 78 - HOT STAND 8f l
P.05.A.3304 : VCT VALVES LIAKED TO ATM3 SPHERE - RELEASE
--- FT. CALH3UN 1 - APR TT - 93X POWER, SEPT 78 - 98E P3WER P.35.4.3325 : CHECC VALVE SPRING PRESSURE EXCES$!VE - SPRING RETAINER MODIFIED - BEAVER VALLEY 1 - OCT TS - SHUTDOWN El P.05.8.3025 : SCALE HELD C4ECK VALVE OPEN "g
--- CONN YANCEE (HADDAM NECK) - APR 74 I)
P.05.8.0025 : VELAN STOP CNECKS STUCK CLOSED - SEATS REDESIGNED
--- RANCHO SECO - 1974 (N3T FUNCT!3NAL TESTING) 7 EI P.35.8.0043 : DEBRIS IN CHECK VALVE CAUSED PUMP TRIP, CRACKED PUMP SEAL
~
s
--- R081NSON 2 - JAN 75 - 103% POWER P.05.8.0079 : UNPLANNED RELEASE - RELIEF VALVE LIFTED DUE TO CHECK VALVE LEAK THREE MILE IS. 1 - SEPT 75 - SU8 CRITICAL l
P.08.8.0082 : DIRT ACCUMULATION - CHECC VALVE FAILED TO SEAT PROPERLY PT. BEAC4 2 - MAR 76 - REFUELING SHUTDOWN P.08.8.0090 : DECAY HEAT PJMP M370R SHAFT SHEARED
--- THREE MILE IS. 1 - M4f & JUN 75 - 20% POWER P.OS.8.0095 : SERVICE WATER B0OSTER PUMPS TRIPPED DURINS ROUTINE $NIFFI1G
--- ROBINSON 2 - DCT 75 - 103% POWER P.05.8.3118 : SERVICE WATER PUM* DISCHARGE CHECC VALVE STUCK - LUBRICATION
--- PALISADES - FES TT - 100! POWER Of P.03.8.3127 : PUMP DISCHARSE CHECC VALVE JAM 9ED - TRAVEL STOP READJUSTED -
REDUNDANT PU1P STARTED 04 WR3NG 4CV SUS
--- CALVERT CLIFFS 1 - APR 77 - 103% POWER g)
'P.05.8.3150 : PIPING DESIG1 ERROR, RHR PUMP NPSN MARGINAL - VALVES THRDITLED, i
ORIFICES INSTALLED 6D FARLEY 1 - JUL 77 (AFTER INITIAL FUEL LOADING)
-c.
DAGE 24 9
P.05.8.0151 : CHECC VALVE STUCK - FLAPPER NINGE *I1 CORRODED j
PALISA3ES - SEPT 77 - 103% POWER P.3S.B.3132 : SER'/ ICE WATER SYSTEM CHECC V4LVE RUSFED gj DAVIS-SESSE 1 - 14Y TS - REFUELING P.05.8.3183 : SERVICE WATER CHECK VALVE HIMGE PI4 CORRODED
()
PALISADES - APR 70 - 70% POWER P.33.8.3187 : DEC4Y HEAT FLOW RATE REDUCED BEL 04 TECH SPEC LIMIT T3 REPAIR
()
VALVE SONNET LEAK DAVIS-8 ESSE 1 - JUN 78 - REFUELING P.08.8.3195 : LEAKS FOUND 11 SPENT FUEL POOL COOLING SYSTEM.
--- ARKANSAS ONE 1 - MAY l N3V 77 - 70% P3WER & COLD S H'J T D O W N P.05.8.0293 : THRE4DED LOCCING DEVICES 3N VALVES, PUMPS, VALVE OPER473R$
FAILED TURKEY PT. 3 - OCT 79 --- DAVIS-9 ESSE 1 - OCT 78 TO DCT 79 --
SALEM 1 - 3CT 79 --- ARCANSAS ONE 2 - FEB 80 --- PWR$ 11 GENERAL P.05.8.0346 : CRANE CHECK VALVE DISC PIVOT PINS JORKED LODSE
--- NORTH AN1A 1 - OCT 77 (C3NSTRUCTION)
P.05.C.0004 : HOSE SLIPPED 3FF TEMPORART FITTING - RELEASE
--- SURRY 1 - AUG 74 - 783 MWE P.07.H.0044 : CHECC VALV! FAILED, JORN SEALS IN SETECTOR - RELEASE I.j
--- MAINE YANKEE - APR 75 - 70% P0 DER P.07.H.0117 : INSTRUMENT AIR COMPRESSOR DISCHARGE CHECK VALVE STUCK, #35313LE RCP SEAL DAM 4GE - VALVES REPLACED. AJTOMATIC BACKUP ADDED ST. LUCIE 1 - APR 77 - 1303 P3 wet
.f gg P.11.A.3122 : DIESEL GENERLTOR, CIL SUPPLY CHECK VALVES IMPROPERLY INST 4LLED IN DRAIN LINE
.)
--- INDIAN PT. 3 - OCT 75 P.11.A.3209 : COMPRESSOR HEAD GASKET 8LEW ' TANK CHECK VALVES LE4KE3 IN DG l
AIR START SYSTEM (EWAUNEE - OCT 77 - 100% POWER P.11.A.0236 : DG AIR STARTING CHECK VALVE 840KE - LINE GASKETS 8LEW OJT COOK 2 - MAR 78 - 0% POWER (AFTER INITIAL CRITICALITY).
J P.11.A.0268 : FUEL INJECTIJN PUMPS, INJECTOR FAILED - PISTON 80LT L30$E COOK 2 - SEPT 78 - SH3TD3JN & 97E POWER J
P.11.A.3269 : DG AIR START RESERVOIR PRES $URE L3J = AIR COMPRESSOR RELIEF, CHECC VALVES LEAKED g
FARLEY 1 - SEPT 78 - 100I POWER P.11.8.0213 : THER9AL OVERLDADS ADJUSTED TO PROPER SETTING - CHECK VALVE LEAKED
()
ROBINSON 2 - JUL 77 - 103% P3WER P.12.A.0064 : LEAKING CHICC VALVES. IMPROPERLY ROUTED LINES - UNREP3 RTE) b TRITIUM RELEASES
PACE 25 IION 1 5 2 - 3CT 76 P.16.A.3358 : CONTAINMENT !$3LATION VALVES & AIRLOCK LEAKED - VALVE $!ATS' REPLACED, FOREIGN MATERIAL FLUSHED, LOOSE STEM SEAL PLAIE TIGHTENED p
PRAIRIE !$. 1 - MAR TT - REFUELING P.16.8.0078 : VALVES LEACED EXCESSIVELY - SERVICE JATER CHECK VALVES DIRTY, O
PURGE ISCLATION VALVE OPERAT3RS ADJUSTED COOK 1 - FEB ?? - REFJELING P.15.A.0373 : CHECC VALVE FAILURES - NFS DEFECTS FEB 80 --- CRYSTAL RIVER 3
- APR 83 ---
THREE MILE 15.
SURRY 1 - J A1 & APR $3 --- FT. CAL 40JN - NOV 78 --- ARKA15AS ONE 2 - 3CT 81 --- DAVIS-SESSE 1 - OCT 83 --- PWR'S IN GENERAL P.15.C.3769 : ADDITION T3 EVI.C.639 RE: PE4ETRAT!01 R004 FLOOR DRAIN RE10 VAL DAVIS-SESSE 1 - JUL 75 - STARTUP (0% POWER)
............. ALL DONE, PRESS < RETURN) (EY TO G3 T3 MENJ................ 7:
j THERE ARE 183 ARTICLE ($) TO BE DISPLAYED:
19-AU3-1985 16:30:56-PWR CHECC VALVES - 1979 - 1982 P.36.8.3089 : CRD 3UIDE TU3E SUPPORT P!1 FAILED.- SGS DAMAGED 'JITHIM H3JRS JESTINGH3USE PWR'S IN GENERAL - JUL 82 P.05.8.0018 : RC MAKEUP LINE CRACKS, THERMAL SLEEVE DAMAGE - FATIGUE FAILURE SUSPECTED
--- CRYSTAL RIVER 3 - JA4 8 2 --- 3 C O N E E 1, 2 & 3 --- PW R' S 11 kj GENERAL P.05.8.0020 : RCS CHECK VALVES LEACED
(
I) 1CGUIRE 1 - FES 82 - HOT SHUTD3W1 P.05.C.0047 : PORV STUCK OPEN - SOLEN 0ID OPERATED CONTROL VALVE LOCCED IN
/'
ENER3IIED STATE BY VENT 1ESTRICTI34
)
s 3INNA - JAN 82 - RAPID C30LD0J1 J
7.05.C.0030 : PORVS IN0PERABLE - LEAKIN3 DIAPHRAGM, AIR SUPPLY CHECK VALVE
--- SURRY 1 - OCT 82 - SH'TD3JN J
P.05.8.0033 : GENERATOR EXCITER REQUIRED REPAIRS - REACTOR TRIPPED. FJ 4AMMER DAMASE
--- PT. BEACH 1 - MAR 81 - P3WER OPERATIO1 P.05.'D.0222 : MSCV DISC STJD LOCKING DEVICE FAILED, NUT FOUND IN TUR81NE CONTROL VALVI
--- TURCEY P3 INT 4 - MAY 80 - SHUTDOJN P.05.D.3223 : AUX FW PUMP STEAM SUPPLY CHECK VALVE LEAKED I
NORTH ANNA 2 - JUN 80 (PRIOR T3 INITIAL CRITICALITY) - H3T i
STANDBY b
P.06.D.3224 : MAIN STEAM NRV GUIDE TO DISC WELD FAILED 9AINE YA1CEE - JUN 83 - STARTUP (15% POWER)
{}
P.05.0.3236 : ADDITION TO VI.D.207 E 222 RE: MSCV DISC STUD NOT LOCCING DEVICE TURKEY PT. 3 & 4 - NOV 83 - SHUTDOWN kh P.05.D.0266 : MS CHECK VALVE ROCKSHAFT DISPLACED AKIALLY - SET SCREJS L30$ENED
.~.. _ -
~_
a PAG!
26 o
ROBINSON 2 - JUN 82 - REFUELIN3 P.05.D.0271 : ADDITION TO VI.D.264 RE: 15 CHECK VALVE ROCKSHATT DISPLACED Ax! ALLY SET SCREW LOOSE 1ED ROBINSON 2 - JUN 82 - REFUELING e1 P.05.D.0274 : CHECC VALVE 3ISC/$7E1 SEPARATION - CYCLIC FATIGUE FAILURE
--- SEQUOYAH 1 - NOV 82 - REFUELIN3
()
l P.05.D.3294 : ADDITION TO VI.D.274 RE: CHECC VALVE STUD BROKEN - FATIGUE.
SEQUDYAH 1 - NOV 82 - REFUELIN3
()
P.06.E.0199 : Loose VALVE 1UT & WASHER - STEAM $UPPLY VRLVC IN0PERA3LE BEAVER VALLEY 1 - MAR 79 - STRRTUP (OE POWER).
)
P.05.E.02A'4 : VALVE STEAM LEAKS SISA8 LED PRVS
--- OCO. NEE SEP 79 - ST% POWER J'
-P.05.E.3257 : ADDITION TO VI.E.227 RE: FW PIPIN3 CRACKS
--- IION 1 --- PT. 8EACH 2 --- KEWAUMEE - JUN TO SEP 79 j
P.05.E.3259 : ADDITION TO VI.E.250 RE: FW PIPE CRACKS
--- SALEM 1 - JUN 79 - REFUELING P.05.E.0272 : VALVE UNQUALIFIED SEISMICALLY - PIPE BRACE ADDED
--- SALEM 1 - JAN 80 - 984 P3WER j
.P.0$.E.0278 : CHECC VALVE 41NGE PIN SET SCREW N3T FULLY INSERTED CRYSTAL RIVER 3 - APR 80 - COL 3 SHUTD3WN g;
P.05.E.3230 : VALVE LEAKS RAISED PENETRATI3N R001 HUMIDITY ABOVE DESISN LIMIT
--- 3CONEE 2 - JUN 80 - 45% POWER
,/
()
P.06.E.3301 : PENETRATION 400M HUMIDITY HIGH - FW VALVES LEAKED 3CONEE 2 - OCT 83 - 100% POWER
/.
gg
- .06.E.3315 : $HUTDOWN REQJIRED TO DEBURR CHECK VALVE PIVST PIN BUSHI1G NORTH AN1A 2 - AUG 80 (LOW POWER PHYSICS TESTING)
DI
'I POWE4 P.06.E.3319 : LEACING CHECC VALVE OVERPRESSURIZED AUX Fd PUMP sdCTION STRAINER - CASE CRACKED
--- C00K 2 - J AN 81 - 72% P0 DER P.06.E.3325 : DIRTY CHECC' VALVE STUCK DPEN COOK 1 - FE8 81 - 100E POWER I
P.05.E.0331 : FW CHECC VALVE SEAT / DISC REQUIRED LAPPING SEQUOYAH 1 - MAR 81 - COLD SHUTDOWN 1
P.05.E.0340 t MAIN FW CHECC VALVE DISC STUD NUT MISSING k
l TURKEY P3 INT 3 - APR $1 - REFUELING U
P.05.E.3362 : AUX FW PUNP CHECC VALVES LEACED
--- COOK 2 - JUL 81 - 90E P0JER P.05.E.3368 : TDAFJP SUCTI3N PIPING OVERPRE550RIZE3 - RELIEF VALVES 115FALLED
()
1CGUIRE 1 - AU6 81 (PRIOR TO INITIAL CRITICALITY) - STARTUP
()}
P.06.E.0370 : Fw P'JMP DISC 4ARGE CHECK VALVE FAILED - UNIT TRIPPE3'
~
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4._. _ - - -
PAGE 27 PT. SEAC4 1 - NOV 80 - GREATER THAN 7PX POWER P.05.E.0373 : EFW PUMP TUR31NE STEAM SUPPLY CHECC VALVES IMPROPERLY ASSEMBLED ARKANSAS ONE 2 - OCT S1 - SHUID0dN P.05.E.0378 : FW CHECK VALVE DISC RETAI11N3 NUT COTTER PINS FAILED SURRY 1 - JAN & APR 83 - SHUTDOW1 P.06.E.3385 : STRIPPER TEED HEATER TUSES LEAKED - RELEASE
--- SURRY 2 - NOV 81 - REFUELING P.05.E.0386 : STEA1 SUPPLY CHECK VALVE PARTS LODGED IN AUX FW PUMP TRIP VALVE
- DISC NUT RETAINING PIN INSTALLATIO1 SUSPECT
--- 10RTH ANNA 1 & 2 - DEC 81 - 99 & 100% POWER P.05.E.0388 : ADDITION TO VI.E.373 RE: FW PUMP STEAM CHECK VALVE
-- ARKANSAS ONE 2 - OCT 51 COLD SHUTD0JN P.05.E.0398 : AUX FW PUMP LEAK 0FF LINE CHECK VALVE FAILED
--- COOK 1 - APR $2 - 103E P3 DER
)
P.05.E.3400 : 48FP EQUALIIING LINE CHECC VALVE STUCK DAMAGING PUMP THRUST BEARING - VALVE INTERNALS RE10VED INDIAN PT. 2 - MAY 82 - NOT SHUTDOWN P.05.E.3402 : MAIN FW CHECC VALVE STUD 1UT FASTENER FAILED, MODIFIED TURKEY PT. 4 - JUN 82 - SHUTDOJN P.05.E.0412 : EFW PUMP TURRINE STEAM SUPPLY CHECK VALVE FAILED - DESIGN ERROR S#,j
--- ARKANSAS ONE 2 - OCT 82 - REFUELING P.05.E.3413 : STEAM SUPPLY VALVE INOPERABLE - NUT 5 WASHER LODGED 11 VA6VE
/
C' 3EAVER VALLEY 1 - SEP 82 - SHUTDOWN P.05.E.0415 : SG FJ !$0LATION VALVE LOW NITROGEN ACTUAT3R PRESSURE - LEAKY
/'
E PLUG ON NITR3 GEN ACCUMULATOR s
--- SAN ONOFRE 2 - SEP 82 (PRE 0PERATIONAL)
J P.05.E.0416 : AUK STEAM TUR5INE PRES $URE CONTROL VALVE FAILED To CL3SE FOREIGN MATERIAL UNDER SEAT - UPSTREAM CHECK VALVES SJSPECTED SOURCE TURKEY P3!NT 3 - SEP $2 - 100E POWER P.05.E.3417 : AUX FW PUMP STEAM SUPPLY CHECK VALVE DISC STUD BR0CENa 1UT AND i
COTTER PIN MISSIN3 l
SALEM 2 - OCT 82 - 82E POWER t
P.05.E.3463 : STEAM SINDIN3 IN AFW PUMPS - LEAKA3E FROM MAIN FJ SYSTE1 PWR$ IN SENERAL - JAN 84 --- C00< 2 - JUL & 3CT S1 - 70X POWER & HOT STANDSY --- CRTSTAL RIVER 3 - OCT 8 3EC $2, N3V S3 -
HDT STAND 8T k
.)
P.05.E.3542 : AUK FW PUMP INOPERABLE - STEAM LINE CHECK VALVE DISC NOT IN i
PLACE - DISC-TO-HANGER STUD SHEARED OFF Of SALEM 2 - AUG 82 - 824 P3JER P,0%.E.0623 : ADDITION TO VI.E.417 8 542 RE: AUX F4 PUMP STEAM SUPPLY CNECC f)
VALVES DAMAGED - PRES $URE OSCILLATIONS - VALVES REPLACED
--- SALEM 2 - AUG, OCT 82 - 821 P3WER g
DAGE 28 P.07.A.3200 : CHECK VALVE LEAK!D - VALVE B3DY OVERSIIED ARKANSAS ONE 2 - JAN 79 - 19% POJER (POJER ESCALATI3N TESTING)
P.07.A.0204 : INCORRECT VALV! WEIGHTS USED IN PIPE STRESS CALCULATI3NS -
gj HANGERS MODIFIED
--- NORTH AN14 1 - MAR 79 - 92% P3WER P.07.A.3229 : BAT 30RON C04 CENTRATION L3W - VALVE LEAKED - LIMIT $d!TCHES READJUSTED
--- IION 2 - OCT 79 - 94% POWER
()
P.07.A.0230 : $ WINS CHECK VALVE SEARING CAP 0-RING FAILED 1AINE YA1KEE - OCT 79 - 97% P3JER P.07.A.3246 : HPI CHECK VALVE SEAT HOLD-DOJNS FDUN3 LODSE a
--- THREE MILE !$ 1 - FEB 80 - SHUTD3WN P.07.A.3256 : CONTAINMENT PENETRATION CHECK VALVE SEATS LEAKED CONN YANCEE (HADDAM NECK) - MAf 50 - REFUELING j
P.07.A.0260 : HPI VALVES LEAKED - PIPINS HEATED EXCES$1VELY 3CONEE 1 - JUN 80 - 73% POWER P.07.A.3262 : FAILED CHECK VALVE, 3 PEN ISCLATION VALVE - 20 GAL RELEASE CRYSTAL t!VER 3 - JUL 80 - SHUTD3WN j
P.07.A.3266 : SIT RECIRC LINE LEAKED - JELD CRAC<ED - VALVE CONTROLLER FAILED MILLSTONE 2 - JUL 80 - 100% P3JER gj P.07.A.3270 : CHECC VALVE STUCK OPEN - DISC WELD / VALVE 30DY INTERFERE 4CE
+
--- SEQUDYAH 1 - SEP 80 - HOT STANDBf
/
g P.07.A.3271 : SIT PRES $URE LOW - NITROGEN CCNTROL VALVE FAILED ARKANSAS ONE 2 - OCT SO - 103E P3WER
,7 gg P.07.A.0273 : CFT 3VERPRES$Ut!!ED-- CtECK VALVE IMPROPERLY ASSEMBLED l
--- DAVIS-BESSE 1 - DCT 80 - HOT STANDBY P.07.A.0275 : HIGH HEAD SI PUMP C H E r k. VALVE ANTI-R3 TAT!3N DEVICE BINDINS DUE To FLOW OR PRESSURE Su tt G E S SEAVER VALLEY 1 - OCT 80 - SHJTD3WN P.07.A.0278 : ADDITION TO VII.A.273 RE: CHECK VALVE
--- DAVIS-5 ESSE 1 - OCT 83 - HOT STAND 8Y i
P.07.A.0280 : SI TANK LEVEL LOW - 30LEN3ID VALVE LEAKED
--- CALVERT CLIFFS 2 - NOV 83 - 100% POWER' P.07.A.3285 : SI SYSTEM CHECK VALVE STUCK 3 PEN
{
- j
--- SALEM 1 - DEC 80 - SHUTDOWN t
P.07.A.3291 5! ACCUMULAT3R BORON DILUTED - CHECK VALVE LEAKED g,
--- $URRY 2 - J AN 81 - 103% POWER P.07.4.3294 : LPI CHECK VALVE DISC COCKED 3Y DEP3 SIT BUILDUP g3 3CONEE 1
- FE3 81 - O! P3WER (HEATUP)
P.07.A.0302 : ADDITION TO JII.A.294 RE: CHECK VALVE gg
--- 3CONEE 3... MAR 81 - 54UT30WN
~
. r., 4..
)
PAGE 29
)
P.07.A.0306 : ACCU 1ULATOR DISCHARGE CHECK VALVES LEAKED 1CGUIRE 1 - APR 81 (PRIOR 70 INITIAL CRITICALITV) - H3T STANDSY P.07.A.3307 : COLD LEG INJECTION ACCUMULAT3R DISCHARGE CHECK VALVES LEACED
--- MCGUIRE 1 - APR 51 (PRIOR TO INITIAL CRITICALITY) - H3T STAND 8f
()
P.07.A.0310 : LEAK TEST DA'iAGED UNI ACCUMULATOR VALVES - SEAT TYPE CHANSED 1CGUIRE 1 - MAY $1 (PtIOR TO INITIAL CRITICALITV) - C3LD SHUTDOWN g)
P.07.A.3311 : ACCU 1UL AT04 CHECK VALVE F AILED - 83TH RHR TR AINS IN0PER A8'.E IN C3LD SHUTDOW1 1CGUIRE 1 - MAY 81 (PRIOR TO 11ITIAL CRITICiLITY)
C3LD SHUTDOWN I
-,l P.07.A.0315 : SI CHECK VALVES STUCK OPEN
--- PT SEACH 1 - JUL 81 - SHUTD0VN
.e 1
P.07.A.0316 : ADDITION TO VII.A.240 RE: TILT!nG DISC CHECK VALVES THREE MILE IS 1 - JUN 61 - COL 3 SHUTD3WN
- t P.07.A.3339 : CHECK VALVE SISC MIS $1NG - C3NSTRUCTION ERROR C00K 2 - OCT 81
PROCEDURES REVISED k,p
--- NORTH ANTA 1 N3V 81 - 79E POWER P.07.A.0343 : LEAKING SI SYSTEM CHECK VALVES SEATED FOLLOWIN'G FLUSHING
/
Il
- T BEACH 1 - OCT 81 - REFUELIN3 P.37.A.3345 : SI ACCUMULAT3R BORON CONCENTRATION HIGH - VALVES LEAKED, FAULTY PROCEDURE
$I s
--- NORTH ANMA 2 - OCT 81 - 97E POWER
.P.07.A.3346 : ADDITION TO VII.A.331,335 8 337 RE: ACCUMULATOR BORON J
CONCENTRATIO15 NORTH ANNA 1 S 2 - SEP 81 - 133X POWER P.07.A.3352 : 81T INLET VALVE FAILED - 3 HAFT PACKING T03 TIGHT SURRY 1 - DEC 31 - C3LD SHUTD3JN P.07.A.3363 : LPSI SWING CdECK VALVE WEAR, DAMAGE - DESIGN
--- PALISADES - SEP 81 - REFUELING P.07.A.3370 : SWING CHECK VALVE PROBLEMS PWR$ IN 3ENERAL - JUN 82 (CB.7.A.1523)
J P.07.4,0372 : HP INJECTION STOP CHECK VALVES STUCK - UP 70 580 PSID REDJIRED T3 ESTABLISH FLOW
--- DAVIS-5 ESSE - JUN 82 - REFUELING U
P.07.4.0375 : ADDITION TO VII.A.372 RE: HP INJECTI3N ST3P CHECC VALVE STUCK -
INCORRECT SEAT ANGLE
()
--- DAVIS-SESSE - JUN 82 - REFUELING --- PWRS IN GENERAL.- AUG 82 6D l
)
DA0E 30 o
P.07.A.3384 : SIT DUTLET C1ECK VALVE LEAKED = 0-RINGS DETERIORATED CALVERT CLIFFS 1 - JUL 82 - STARTUP e
P.07.4.3386 : COLD LEG INJECTION CHECK VALVE LEA <ED - CARBON STEEL CL3SJRE STUDS CORRODED
--- NORTH ANNA 1 - SEP 82 - REFUELING gj P.07.A.0390 : ADDITION 70 VII.A.275 RE: VELAN CHECC VALVE BINDING P403 LENS -
()
DISCS REPLACED SEAVER VALLEY 1 - OCT 82 - SPECIAL TESTING P.07.A.3392 : SI CHECK VALVES STUCK - DISC STUD PR3TRUDED A83VE NUT, DISC MISALIGNED
--- ARKANSAS ONE 2 - DCT $2 - REFUELING P.07.A.3396 : MINOR LEAKAGE INTO 31T C01P00NDED SY SIT LEVEL' INDICATING SYSTEM FAILURE CAUSED TEC4 SPEC LIMITS TO BE EXCEEDED - VALVE LEAKAGE, TEMP EFFECTS ON REFERENCE LEG PALISADES - SEP-DEC 82 - 100E PodER P.07.A.3397 : SI TANK LEVEL LOW, BORON CONCENTRATION LOJ = VALVE LEAKAGE PALISADES - N3V S2 - 100t POWE4 7.07.A.3400 : ADDIFION T3 /II.A.399 RE: CHECK VALVE DISC MIS $1NG
--- C00K 2 - DEC S2 - REFUELING P.07.4.3403 : ACCUMULATOR 30RON CONCENTRATION LOW - CHECK VALVE LEACED SURRY Z - SEP 82 - 103% POWER kj P.07.A.3422 : ADDITION TO VII.A.384 RE: $1T GUTLET CHECK VALVE LEACED RINGS DETERI3 RATED - MATERIAL CHANSED
--- CALVERT CLIFFS 1 & 2 - JUL 82 - STARTUP
/
O P.07.8.0068 : SPRAf PUMP M3 TOR STATOR C31LS/ PHASE LEAD CHAFED, SHORTED
--- CALVERT CLIFFS 1 - SEP 79 - 103% POWER
/.
P.07.8.3080 : CHECC VALVE DISC AND RELATED PARTS N3T INSTALLED
--- SEQUDYAH 1 - APR 80 - PRCOPERAT!3NAL
.)
P.07.8.0054 : SWST LEVEL L3d - R85 SYSTEM VALVE FAILED TO RESEAT, ALARM FAILED
--- 3CONEE 3 - APR 80 - HEATUP P.DT.8.3123 : CONTAINMENT SPRAY ADDITIVE TANK CHECK VALVES LEACED
--- RANCHO SECO - APR 82 - SHUTDOJN
.s P.0F.8.0131 : C ot4T A INMENT SPRAY & NA0H ADDITIVE SYSTEM CHECK VALVE FAILED
- SEAT & DISC MACHINED GINNA 1 - SEP 82 - 103% POWER J
P.07.8.0134 : SPRAf PUMP CHECK VALVE PIN / SWING ARM WORN i
JUN 82 - 100% POJER 1
GINNA P.37.8.0142 : CONTAINMENT SPRAY PUMP DISCHARGE CHECK VALVE FAULTY - PIN / SWING O
ARM JORN
--- GINNA - JUL 82 - 1001 POWER P.OT.D.0034 : STUCC VALVES FREED BY CYCLIN 3
- CHECC VALVE INSTALLED BAC(WARDS (I
- PR3CEDURE ERROR SEP 79 TO NOW 83 - 133% POWER --- C00< 2 - JJN $3
- COOK 1
- 5) /
75% Podka i
T
- AGE 31 o
P.07.0.0047 : C3NTAINMENT STEAM HEAT SUPPLY CI CHECK VALVE FLANGES LEAK!D.-
j SOLTING RELAtED YANKEE R3dE - APR 82 - 130% P3 DER P.07.0.0053 : CONTAINMENT !$0LATION VALVES FAILED LLRT - COMP 0NENT F AIL-JRE, DESISN/FASRICATION ERRORS DAVIS-5 ESSE - JUN 82 - SHUTDOW1
()
P.07.D.0059 : C3NTAINMENT !$0LATION VALVE LEAK RATE GREATER THAN RANGE 3F TEST EQUIPME1T - NEW TESTER PURCHASED
()
==- SALEM 1 - APR 8 2 - RE FUELING (CO'.D SHJTDOWN)
P.07.E.0078 : HYDR 3 GEN PURSE FAN TRIPPE) - CORR 33ED CHECK VALVE ARKANSAS ONE 1 - MAR S0 - 1005 P3WER P.07.E.0034 : SW LEAKED BY IDLE VENT FA1 SEALS, 'IPE F0JLING - HYDR 3 GEM PURGE FLOW LOW ARKANSAS ONE 1 - JUN 5 JUL 80 - 100% POWER P.07.E.0090 : HYDR 3 GEN PURSE FAN TRIPPED - SEAL WATER LEAK CAUSED CHECK VALVE l
l STICKING AND FILTER RESTRICT!DN
--- ARKANSAS ONE 1 - SEP S0 - 100E P3WER P.07.E.0129 : FIRE PROTECT!DN SPRINKLER SYSTEM ALARM CHECK VALVE LEAKED -
DISC SEAL INVERTE3
--- CONN YANCEE (HADDAM NECK) - MAT 52 - 300% POWER P.07.E.0151 : HYDR 3 GEN DILJTION SLOWER CHECK VALVE STUCC - MISAPPLICAT!3N, l
I LIFT CHECK REPLACED WITH SWING CHECK i
--- DAVIS-SESSE - Nov 82 - 99E POWER 9
4
()
P.07.E.3174 : FIRE PROTECTION SPRINKLER SYSTEM CHECK VALVE GASCET LEACE3
--- YANKEE R3dE - SEP 82 - 82% POWER l'
()
l P.07.E.3175 : SPECIAL HAIARDS HEADER ALARM CHECK VALVES FAILED TO SEAT s
--- YANKEE R3WE - DEC 82 1 FE8 83 - REFUELING & 100% PodER P.05.A.3341 : LOOP FILL CHECK VALVE LEACED - SEATI4G $URFACES LAPPED l
CONN. YA4KEE CHADDAM NECK) - FEB 79 - REFUELING d
P.05.A.0352 : STORAGE TANK 80RON LOW = CHECK VALVE LEAKED
--- GINNA 1 - APR 79 92% POWER l
P.05.A.0401 : CHAPSING PUMS DISCHARGE CHECC VALVE SEAT LEAKED PT. SEAC4 2 - DEC 79 - 98% POWER l
P.05.A.3404 : LEAKING CHECC VALVE DILUTED SORIC ACID ST3 RAGE TANKS
--- GINNA 1 - DEC 79 - 3% P0JER P.05.A.0414 : ADDITION TO VIII.A.404 RE: BAST VALVE LEAK J-
--* SINNA 1
- DEC 79 - 35 PoJER EI P.05.A.3438 : MAKEUP SYSTE1 CHECK VALVES LEAKED - SUILDUP DN SEATS
--- DAVIS-8E3SE 1 - MAY 83 - REFUELING
()
P.03.A.0459 : WRONS INTERNALS USED IN VALVE
--- NORTH AN1A 2 - SEPT 83 (PRIOR 70 POWER OPERATION) - C3LD SHUTDOWN l
t
PAGE 32 P.03.A.0477 : HPI 8 UMP BEARING DVERHEATED - COOLING SYSTEM DEFICIENT, FILTER CLOG 3ED, CHECK VALVE STUC(, PER$31NEL. ERR 3R 3CONEE 2 - Nov 83 - 130% POWER P.08.4.0508 a!T, BAST 8040N DILUTED - CHECK VALVE LEACEO gj
--- 10RTH ANNA 2 - DEC 80 $ FEB 81 - 1001 POWER P.05.A.3527 : VALVE BONNET GASKET LEAKED - PO$f-1AINTEN4NCE TESTIN3 01tTTED
()
SEQUOYAH 1 - DEC SO - 103% POWER P.05.4.3553 : CHAR 3!NG PUMP FAILED TWICE
()
--- 4RKANS45 DNE 2 - JUL, AUG 81 - 72, 40% POWER P.05.A.0554 : CHAR 3ING LINE CHECK VALVE STUDS CORR 3DED COOK 2 - OCT 51 - SHUTD0JN P.05.A.0566 : VCT 545 $UPPLY VALVE LEAKED - RELIEF VALVE SETPOINT L3W -
RELE4SE CALVERT CLIFFS 2 - N3v 81 - HDT STAND 3Y P.05.A.3575 : CONT 4!NMENT !$0L4T!0N VALVES REQUIRED MAINTENANCE 3EAVER V4LLEY 1 - APR 80 - SHUTD3WN J
P.39.A.3580 : DI W ATER HE A3ER C3NTAMIN ATED - CHECK VALVE LE AKED JAN S2 - 85% PodER ARKANSAS ONE 1 J
P.05.A.3618 : CHAR 3ING PUMP OIL PUMP CHECK VALVE STUCK - GUIDE FINGER BENT COOK 2 - APR 52 - 100E POWER P.09.8.0228 : SERVICE WATER VALVE DISC 40R4 AT HINGE
--- COOK 2 - JAN 79 - 100% P3WER r
P.08.8.0233 : SALTWATER PU1P DISCHARGE CHECK VALVE FAILED TO OPEN = FDRIIG4 DEBRIS SUSPECTED
--- CALVERT CLIFFS 1 - FE3 79 - 96E POWER
/.
gg P.08.8.0236 : SERVICE WATER CHECK VALVE DISC HINGE POINTS WORN - REPLACED WITH SS DISCS COOK 1 - FEB 79 - 1005 POWER P.05.8.3249 : ADDITION TO VIII.5.234 RE: CHECK VALVE HINGE WEAR COOK 2 - APR 79 - 103% P,WER P.05.8.3255 : C3NT41NMENT !$0LATION VALVES LEAKED'- NEW TYPE CHECK WALVES INST 4LLED
--- COOK 2 - JUN 79 - REFUELING P.05.8.0265 : WELD CRACKS FOUND FOUND I4 83 RATED WATER PIPING SEPT 79 - 100% P0JER SAN ON3FRE 1 P.08.8.0290 : NESW VALVES LEAKED - $AND DEPOSITS, PITTED SEATS k
J
--- COOK 2 - DEC T9 - REFUELING U
f P.08.8.0310 : RIVER WATER CHECK VALVES DETERIORATED
--- SEAVER V4LLEY 1 - APR 80 - SHUTDOWN I
P.03.8.0319 : SERVICE WATER PUMP DISCHARGE CHECK V4LVE DETERIORATED
- ALISADES - AUG 79 - 30% POWER
- 5) /
P.0$.B.3364 : ADDITION T0 d!!!.3.317 RE: CdECK VALVES - SEAT RING C3RR35101 l
1 PACE 33 PALISADES - AUG 79 - 80% POWER P.08.5.3380 : CHARSING PUMP SW PUMP SPL4SH'O WITH JATER - MOTOR WINDINGS SHORTED SURRY 2 - APR 81 - 133% POWER gj P.03.8.3382 : INADVERTENT CONT 4INMENT SPRAT - DESISN/ VALVE ERR 3R
--- CRTSTAL RIVER 3 - JUL 80 --- SEQUDYAH 1 - FES 81 - CD'D SHUTDOWN --- PWR$ IN GENERAL - MAR 81 g)
P.09.8.3394 : SW PJMP DISC 1ARGE PIPING, VALVE SEAT RING THREADS CORR 03E3 INDIAN PT. 2 - M47 81 - COLD $1UTDOWN
()
P.03.8.3411 : LEAKING NESW CHECK VALVES TO SE REPL4CED d!TH DI4PHR4SM TTPE VALVE
--- COOK 1 - JUL 81 - REFUELING P.03.8.0423 : CHARSING PUM8 SW CHECK VALVE FAILED TO SEAT - DISC W04N/ BINDING
/
SURRY 2 - AUG 81 - 133% POWER P.05.8.3461 : CHARGING PUM8 SW PUMP CHECK VALVE WORN, STUCK SURRY Z
- JAN 82 - 85% P3WER P.09.8.3493 : SW STSTEM SWING CHECK VALVE DISC SE!!ED FULLY OPEN - CORR 3513N DEPOSITS, DISC ST'D DETEq!0 RATED J
--- DAVIS-SESSE - JUN 82 - C3LD SHUTDOWN P.03.8.3521 : CCW SUPPLY IS3LATION VALVE LEAKED - CLAPPER AND SEAT LAPPED
--- PT. BEACd 1 NOV 82 - REFUELING (j
'l P.09.8.3566 : RSCU SW CHECK VALVE STUCK CL3 SED SUMMER - DCT 82 - 0% POWER (ST4RTUP) 1
,/
)
P.39.a.3571 : $W PJMP STRAINER CLOGGED - $d 84Y FL300ED, VALVE G4SCET F4 FLED SALEM 2 - DEC $2 & JU1 83 - $5% POWER 8 SHUTDOWN
,/.
g P.03.8.0610 : ADDITION To VIII.S.3?B RE: MISSING FLANGED BEARI1G 04 V4LVE ACTUATOR FOU1D TO BE CAUSE OF FAILURE DAVIS-3 ESSE - APR 81, NOV 82 - 93% POWER
-8 P.09.D.0226 : AUX FW PUMP SPEED LOW - 0/ERSPEED CHECK VALVE LE4KED, GDVERNOR MISADJUSTED SALEM 2 - MAR 82 - 100% POWER P.09.F.3120 : VUCDT FLOW T37ALIIER ERRATIC - CHECK VALVE LEAKED
~/
1CGUIRE 1 - AUG 1 SEPT 82 - 50% P0 DER P.09.H.0581 : R/A SAS MIN 0 TOR CHECK VALVE LEAKED - POSSISLY DIRT FR3M PJMP 3
VANE WEAR
--- SINNA - 3EC 81 - 100% P0 DER J
P.09.H.0637 : CONT 4!NMENT !$0LATION CHECK VALVE LE4KED, VALVE INTER 1ALS DIRTY
- REPORTING PROCEDURES REVISED
--- PT. SEACd 2 - APR 80 $ JUN 82 - REFUELING E 103% P0 jet b
P.09.H.0640 : RADIATION N04! TOR SAMPLE RETJRN LINE CHECC VALVE LEAC - F3 REIGN MATERIAL, BOLTID 30N4E'T TTPE VALVE INSTALLED I)
--- GINNA - APR 82 - REFUELI4G kh P.09.H.3689 : R/A 10NITOR CHECK VALVE LEAKED - F3 REIGN MATERIAL 1
1
- AGE 34
)
SINNA - JUN & SEPT 82 - 1003 P3WER J
P.13.A.0060 : C3NTAMINAT101 0F PRIMARY MAKEUP T41C RESULTED IN ERR 31E3US 63RON ANALYSIS
--- SAN ONOF4E 1 MAY 80 - REFUELIN3 g)
P.11.A.3408 : CHECC VALVE LEAK DG SL3W T3 RE ACH RATED SPEED FARLEY 1 - MAY 80 - 100% POWER
()
P.11.A.3450 : DG START TIMES EXCESSIVE - FUEL SUPPLY CHECK VALVE LEAKED FARLET 1 - MLR 70 SEPT 83 - 133% POWER
()
P.11.A.0482 : ADDITION TO XI.A.466 RE: DG AI4 1 TART SYSTEM
--- (EWAUNEE - JAN 81 - 100X F0WER j
P.11.A.3525 : DG PROBLEMS
--- FARLEY 1 & 2 - APR E 1AY 81 - 103% POWER & SHUTD3dN J
P.11.A.3541 : DG C30 LING W4TER VALVE SH4FT/ DISC MISALIG1ED - V4LVE 3PER4 TION REVERSED
--- PALISADES - JUN 81 - $6% POWER J
P.11.A.3570 : DG FJEL DIL CONNECTION LO35E, CHECC VALVE SEATED IMPR3PER.Y FARLEY Z - SEPT 81 - 100! POWER P.11.A.3589 : DG F'JEL OIL LINES REQUIRED M4NUAL PRIMING
--- ROBINSON 2 - NOV 81 - 50% POWER P.11.A.0614 : DG C30 LING W4TER PUMP DISCHARGE CHECC VALVE DISC SEP4 RAT!3N -
ENGINE TRIPPED ON HIGH TEMP
--- 8WR$ IN 3ENERAL - DEC 81 ((8.11.4.430))
P.11.A.3656 : EMER3ENCY DG PROBLEMS
--- COOK 1 & 2 - JUL 82 - REFUELIN3 1 100% POWER j.
P.11.A.0717 : DG MECHANICAL PR3BLENS REWIEJED
~
s
--- PWR'S IN GENERAL - AU3 83 --- CALVERT CLIFFS 1 8 2 - LPR 33 -
-- RANCHO SECO - MAY 83 --- SAN ONOFRE 1,
,j P.11.A.3721 : DG STARTING FAILURES - BR3CEN INLET AIR CHECK VALVE DISC, MIS $1NG COUNTERWEIGHTING - FUEL SUPPLY LINE REQUIRED PRIMING
--- CALVERT CLIFFS 1 & 2 - JUL 82 $ LPR 83 - 100! PodlR P.11.8.0519 : CIRCUIT BREACER CONTACTS MIS 4LIGNED (EWAUNEE - APR 81 - 100% POWER P.13.A.0146 : WASTE GAS RELEASES - RUPTURED COMPRESSOR RUPTURE DISCS, F41 LED J
PRES $URE CONTROL VALVE - ALARM RESPONSE DELAYED SAN ONOFRE 2 L 3 - SEPT S2, MAY 34 - 1003 POWER, HDT STANDBY
\\
./
P.14.A.0323 : CONT 4!NMENT PENETRATION LEAKAGE - VARIOUS CAUSES
--- PALISADES - N3V 80, SEPT 81, AUG S OCT 83 - 03 P3 DER (COLD b
SHUTDOWN & REFUELING)
P.14.8.0172 : RUST SCALE L30GED IN H2 PJRGE SYSTEM CHECC VALVES II
--- ARKANSAS ONE 1 - JUN SO - 100% P3WER P.14.8.3228 : CONTAINMENT PURGE VALVE AIR SUPPLIES LEAKED NM f
- ALISADES - SEPT B1 - REFUELIN3 I
1 s
- AGE 35 o
P.16.8.0237 : SW PUMP ROOM DRA!45 UNIS0LABLE - DESIGN ERROR j
--- CALVERT CLIFFS 1 5 2 - NOV 81 - 100% POWER
- .15.A.3376 : CONTAINMENT ISOLATION SERVICE AIR CHECK VALVE LEAKED - DISC DIRTY
--- PT. BEAC4 1 - OCT 81 - REFUELING P.15.A.3379 : SERVICE AIR CHECK VALVE LEAKED - INTERNALS DIRTY
- T. BEAC4 1 - OCT 81 - REFUELING P.15.4.03$4 : AUX FW SYSTE1 CHECK VALVES L3CKED SEISMIC PROTECTION SALEM 1 5 2 - FES 82 - REFUELING S 103% POWER P.15.A.0391 : CHECK VALVES LEAKED SEAT C3RROSI3N
--- PT. 8E4C4 1 - MAR 82 - 93% POWER J
P.16.C.0823 : SURVEILLANCE PROCEDURE DEFICIENT - PLANT 3RDERED SHUT D3W1 4RKANSAS ONE 1 - JUN 19 * $TARTOP ((
J P.16.C.0854 : REQUIREMENTS FOR MODIFYINS TEST PR3CEDURES DEFICIENT
--- DAVIS-3 ESSE 1 - JUN 77 - STARTUP P.16.C.0722 : ADDITION TO XVI.C.823 RE: PROCEDURE VIOLATION (RKANSAS ONE 1 - J UN 79 - STARTUP P.16.C.1135 : RIVER WATER SYSTEM CHECK VALVES INSTALLED BACKWARDS SEPT 83 - REFUELING
--- 3EAVER V4LLEY 1 I#
P.16.C.1139 : MAINTENANCE FAILED TO DOCUMENT REM 3V4L 0F CHECK VALVE 11 TERN 4LS SURRY 2 - SEPT 83 -'130% POWER
'l I
()
P.16.C.1297 : CHECK VALVE CLOSURE WEIGHT MISSING 7
--- NORTH ANNA 1 - MAR 81 - SHUTD0dN P.16.C.1561 : VALVE NOTOR FOUND DISCONNECTED - M3DIFICATION WIRING ERROR,
/
()
--- SEQUDYAH 1 - NOV B1 - 103% POWER
?
J P.16.C.1776 : BAAT PUMP DISCHARGE VALVE FOUND CLOSED - PROCEDURE ERROR DAVIS-8 ESSE 1 - M4Y 82 - COLD SH3fDOW1 J
P.15.C.1848 : NON!3ENTICAL REPLACEMENT PARTS - REC 3RRINS PR08LEM 3EAVER V4LLEY 1 - OCT 80 --- BELLEFONTE 1 5 2 - DEC 82 ---
PWR'S
)
+
P.15.C.1909 : EDG STARTING AIR SANC PRES $URE Lod - OPERATOR ERROR SEPT 82 - 100% POWER SURRY 1 J
............. ALL DONE, PRESS < RETURN > (EY TO GD T3 MENU............. g..ft THERE ARE 95 ARTICLE (S) TD BE DISPLAYED:
19-4UG-1986 14:37:33 Pdt - CHECK VALVES - 1983 - 1986 P.05.C.0059 : RCS LEAKAGE - PORV 8 LOCK WALVE, SI C4ECK WALVE, SG TURE b
--.1ILLSTONE 2 - MAR 83 - 130% P3 DER.
P.05.E.0063 : PORV LEAKAGE PROBLEMS, CONTROL AIR ISOLATED, PCS VENT PAT 4 L3ST
()
- RA4 DOM MAI1TENANCE PRO 3LEM - PORW LEAKAGE PROBLE1 U1RES3LVED SALEM 2 - JAN 83 - C03LD3dN, SHUTDOWN P.05.F.0043 : PRIMARY C00L4NT SYSTEM LE4KASE - CHECK VALVES FLUSHED T3 G) j
)
PAGE 36 IMPR3VE SEATING PALISADES - JUL $4 - HOT STAND 9Y j
P.35.F.3052 : PLANT SHUTD0J4 - EXCES$1VE RCS LEACASE - VALVE PACKINS, RCP FLANSE LEAKED
()
--- CATAWBA 1 - OCT 85 - 103% POWER P.36.A.3096 : REACTOR TRIP DN LOW SG LEVEL, POWER TRANSFERRED FROM 1AIN 70
()
RESERVE - tut 91NE STOP VALVE FAILED TO CLOSE - DISC SEPARATEDs SEAT DAMAGED 1AINE YANCEE - N3V 84 - $0% P3 DER
().
P.05.A.3101 : TURBINE / REACTOR TRIP - HIGH TURS!4E VIaRATION DURING CONYROL VALVE TESTIN* - MAIN $ AUt FEED PUMP PROBLEMS - SPEED SETPOINT J
HIGH. OIL PU1P DISCHARGE CHECK VALVE STUCCr MOTOR FAILED DAVIS-8 ESSE - JUN 85 - 85% POWER P.06.A.3104 : REACTORr TUR3INE TRIP - F1 ACCUMULATED IN MAIN TUR8INE HP CONTROL CIL SYSTEM - DIRT ENTERED STRAINER $ AND ORIFICES DURING MAINTENANCE INDIAN PT. 3 - OCT US - 30% POWER P.05.0.3290 94IN STEAM N3NRETURN CHECC VALVES STUCK OPEN - PACKINS INDUCED FRICTION - C3MMON MODE FAILURE
--- P WR $ IN S EN E R AL- - AUG S3 --- TROJAN - MAR 83 P.05.D.0305 : ADDITION TO VI.D.290 RE: MAIN STCA1 NONRETURN CHECK VALVES STUCC UPON - PACCING INDUCED FRICTION TROJAN - MAY, SEP & QCT - COLD SHUTDOWN, POWER OPERATION gj P.05.D.0319 : REACTOR / TURBINE TRIPS - GROUND IN FW PUMP TRIP, CIRCUITRY - MSCV PROSLEMS, FRAYED INSULATION - VIGRATION
/
p)
TROJAN - MARS APR 84 - 130, 78% POWER
~
P.05.D.3328 : ADDITION.70 VI.D.319 RE: REACTOR /TURSINE TRIPS - MSCV P433LE15 -
/'
HIGd FRICT!3N F3RCES FOUND NN s
TROJAN - MARr APR 84 - 130, 78% POWER P.36.D.3333 : MSIV FAILURES - ACTUATOR ROD SEAL PISTON LEAKAGE, FAULTf MAND SWITCH CONTACTS, REDUCED ACTJATOR PRESSURE
--- CALVERT CLIFFS 1 - JUL-DEC 84 - VARIOUS POWER LEVELS P 05.D.0347 : MSIVS FAILED TO CLOSE - AIR SUPPLY CHECK VALVES SEATED IMPR3PERLY - INSUFFICIENT SOFT-SEATED POPPET GUIDING, R3USH SODY SEAT FINISH
--- SYRON 1 - MAR 85 - HOT STAND 8Y 9.05.D.3348 : ADDIT!0N 70 WI.D.347.RE: 1SIV CHECK VALVES SEATED IM*RD'EtLY ON GRADJALLY DECREASING AIR PRESSURE - CHECK VALVES REPLACED
--- PWR$ IN SENERAL - APR 85 --- BYR3N 1 - MAR 85 - H3T STAND 8Y -
j
-- OTHER UNITS AS LISTED IN ARTICLE P.05.0.3333 : REACTOR TRIP - MSIV CLOSED - PRES $URE TRAMSDUCER FAILED, CHECK g
VALVE LEAKED JATERFORD 3 - MAY 85 - 17% P0 DER P.35.D.3367 : MSIV FAILED TO CLOSE FULLY - GAS SUSBLE I4 HYDRAULIC FLJI)s HP HYDRAULIC SuePRES$3R 9 LADDER PR03LEM3s ACCU 1ULAT3R CAP EN) C4ECK VALV! FAILURE
+
]
I
--- CALVERT CLIFFS 2 - JUL-AJG 85
- HOT STAND 8Y w--.
PAGE 3T P.05.E.0444 : FW AIR ACCUMJLATOR CHECK VALVES LE4KED WATER IN INSTRUMENT AIR
--- CRYSTAL RIVER 3 - MAR 83 - REFUELING P.05.E.0454 : SD4FJ PUMP I1 OPERABLE - P37ENTIAL STEAM SINDING, 84 CELE 4K4GE g
THROJGH DISCHARGE VALVES SYSTEM REVIEW UNDERTAKEN 4091NSON 2 - JUL S3 - 79E POWER
(
P.05.E.0462 : AUX FW PUMPS IN0PERABLE, VAPOR B0UND - CHECK VALVEr LUBE 31L' COOLER LEAKASE SURRY 2 - NOV & DEC 83 - 100% PodER
().
P.06.E.3468 : FAILURE OF ANCHOR DARLING SOFT SEAL VALVE SEALS - IMPROPER VOLCANIIING
--- PWR$ 11 SENERAL - FE8 84 --- INDIVIDU4L PLANTS AS LISTED (CS.S.E.172))
P.05.E.3481 : MDAFWP DISCH4RGE CHECK VALVE LEAKING - HINGE PIN BUSHINGS MISSING, WORM FARLEY 1 - DEC 83 - 100% POWER
)
P.05.E.3483 : REACTOR TRIP - RCP TRIP - WATER IN M3 TOR LEAD PENETR4TIONS -
MAIN FW RESULATING V4LVE LEACAGE, FAULTY R00 POSITION 11DICATOR
~
l SURRY 2 - MAR 84 - 133% POWER P.05.E.0489 : REACTOR TRIPS - SG FW LEVEL CONTR3L SYSTE4 PROBLEMS -
COMP 3NENTS D4MAGED - WATER H41MER SALEM 2 - APR 84 - 22, 6, 5% P3WER Il P.35.E.3495 : REACTOR, TUR3INE, GENERAT3R TRIPS - 1AIN FW TRANSIENT - STUCC PILOT VALVE, CHECK VALVE FAILURE J
1CGUIRE 2 - MAY 54 - 100 & 23% P3WER 4
7 g)
P.05.E.3505 : REACTOR TRIP DN L3W SG LEVEL - CIV FLILED CLOSED - SOLENDID VALVE FAILURE
/
E)
--- 1CGUIRE 1 - JUL $4 - 100t POWER P.06.E.0521 : REVERSED TDAFP ROTATION CAUSED SUCTI3N PIPING OVERPRES$UtIIATION - SUCTION INSTRUMENTATION DAM 4GED - CR4CKED
~>
CHECC V4LVE dELD 1CGu!R! 2 - AUS S4 - 100% POWER f
P.06.E.0524 : MANU4L REACT 3R TRIP FW HEATER TU3E LEAK - SG. LEVELS DECREASED
- BINDING CHECK VALVE STE1 MILLSTONE 2 - Nov 84 - 62% POWER P.05.E.0555 : LP FJ HEATER CASCADE DRAIN CLOSED, HEATER SHELL SIDE UNDER J
VACUJM - CHECK VALVE INSTALLED SACKWARDS 3YRON 1 - MAR 85 - 20% P3WER 1
P.05.E.0574 : ADDITION TO VI.E.447 5 454 RE: AUK FJ PUMPS TRIPPED, IN3PER49LE
{
DISCHARGE WALVES LEAKED
--- 108INSON 2 - APR, JUL S3 - SHUTD3WN, 79% POWER P.06.E.3588 : REACTOR TRIP, ESF ACTUATI3NS LOW SG LEVELS - SG FW CHECC VALVE LEAKAGE, PR3CEDURAL ERROR
()
--- J0LF CREEK - JUN 85 - HOT STANDBY (PRE 0PERATIONAL)
P.05.E.0594 : REACTOR TRIP DN MAIN FW PUMP TRIP - FW PU1P RECIRC VALVE FAILED kh TO HANDLE FL3W TRANSIENT - C3NTROLLER DESIGN PROSLEM - 53 PORVS
PAGE 38 J
i FAILED TO OPEN - OUT 3F CALI5 RAT!31 CATAWBA 1 - JUN 85 - 645 P0JER
/
P.06.E.0615 : REACTOR, TUR5!NE TRIPS - FW PUMPS TRIPPED - FW ISOLATION #ALVE, IRM CHANNEL AND FJ FLOW TRANSMITTEP FAILURES 81
--- SUMMER - SEP $5 - 93% P0 DER P.06.E.0624 : REACTOR TRIP DN L3S5 3F VITAL BUS - LOSS 3F FW SYSTE1 11TIGRITY
()
- WATER HAMMER - 1AIN FW CHECK VALVES FAILED PWRS IN SENER4L - JAN 86 --- 54N DN0FRE 1 - NOW $5 - 60%
POWER
()
P.36.E.3630 : STOP CHECK V4tVES IN LUX FW PUMP STE41 SUPPLY FAILED 3N L3W FLOW - DISC 435E19LY DEGRADED - VIBRATION PWR$ IN 3ENERAL - FE3 86 --- TJR(EY PT. 3 8 4 - N3V 85-JAN
$$ - VARIOUS POWER LEVELS ((
P.06.E.3634 : 53 BLOWD3WN 13NITOR INOPERABLE - FW IN BLOWDOWN LINE - CHECK VALVE BACKLEAKAGE
--- YANKEE R)WE - DEC 85 - OC POWER (STARTUP)
J P.06.E.3644 : AUX FW 5 TEAM SUPPLY ISCLATION VALVES IN0PERA8LE - GUIDE SFUDS BROKEN - HIG4 CYCLE FATIGUE TURKEY PT. 3 & 4 - JAN 86 - 103, 1005 POWER P.06.E.3646 : CHECK VALVES OMITTED-FROM EFJ PUMP STEAM SUPPLY - IN4)EQU4TE J
DESIGN REVIEJ ARKANSAS ONE 1 - JAN 56 - 90% P0JER Ih P.06.E.0648 : P3TENTIAL FAILURE OF ANCHDR DARLING CHECK VALVES - INTERN 4L TACK WELDS MISSING - MFG ERR 3R 4
/
SUMMER - FEB 86 - 103% P3WER g)
P.06.E.0673t REACTOR /TUR9INE TRIPS, FW IS3LATI31$ - SG FW FLOJ C01TR3L PROBLEMS, C01 TROL CARD N3T ADJUSTED, POT, VALVE FAILURES
/.
bI CATAWBA 2 - MAY 86 - #ARIOUS P3 DER LEVELS (PREOPERATI3NAL) y P.06.E.36748: REACTOR TRIP DN MAIN FW PUMP TRIP - HEATER DRAIN PUMP TRIPPED -
CHECK VALVE STUCK OPEN - SALLING 34 HINGE PIN 1
3CONEE 1 - MAY 86 - 45% POWER P.06.E.0675t TURBINE, REACTOR TRIPS ON HIGH SG LEVEL - FEED' PUMP DISCHARGE LINE CHECK VALVE ALL3dED REVERSE FLOJ ON STARTUP - DISC P!VOT PINS MISSING - SETSCREWS NOT ANCH3RE)
INDIAN PT. 3 - MAY 86 - 25% P3JER P.07.A.3401 : SI CHECK VALVE SEAL WELD LEAK - INADEQUATE APPLICATION 3F WELDING AND GRINDING TECHNIQUES FES 83 - SHUTDOJN
- - 1AINE YA1CEE P.07.4.0407 : COLD LEG INJECTION ACCUMULAT3R CHECK VALVES LEAKED 1CGUIRE 1 - MAY 83 - STARTUP, 35 & SOE P3WER 4
OI P.07.A.3420 : HPI STOP CHECK VALVE STUCK - DISC TO SEAL CONTACT ARE4 T03 WIDE DESIGN DEFICIENCY 3 AVIS-9 ESSE - SEP 83 - SHUTDOWN
()
P.07.A.3437 : $1 CHECK VALVE LEAKING - INC3MPLETE CONTACT BETWEEN DISC & SEAT
--- FARLEY 2 - SEP $3 - REFUELINS gg
m_m PAG 59 J
P.37.A.3447 : VELA 4 CHECK VALVE ANTI-R3 TAT!0N STOP JAMMING - DESIGN PR03LE1 DAVIS-SESSE - OCT 84 - REFUELING e
P.07.A.3452 : EXCESSIVE $1T CHECK VALVE LEAK 4GE - SEAL PLATE C3CKED, V4.VE 2
SEAT COMPENS4 TIN 3 JOINT 34LL GALLED FM gj ST. LUCI! 2 - DEC 34 - H3T STA1037 P.07.A.3456 : EXCESSIVE $1T CHECK VALVE LE4KAGE - ETHYLENE PROPYLENE 3-TING
()
M4TERIAL DEGRADATION CALVERT CLIFFS 1 - J AN 85 - 133% POWER P.07.A.0457 : UNIT SHUTD3W1 ~ LOW ACCUMJLATOR 33RON CONCENTRATION LEACINS VALVES 1CGUIRE 1 - APR.$5 - 33% POWER y
P.07.A.346$
LPSI PUMP, $1 TANC INOPER4BLE - MAINTENA4CE, CHECK VALVE EAKAGE
--- PALISADES - N3V 85 - 78% POWER J
P.07.A.3474 : HIGH SI TANK LEVEL, LOW B3 ROM CONCENTRATION - PRIMARY C30LANT LEAX4GE THROJ3H $! CHECK VALVE PALISADES - N3V 85 - PS% POWER P.07.A.0485 : RCS NOT ISOLLOLE FR01 LP COOLANT SYSTEMS - PCIV LEARAGE PWR$ IN 3ENERAL - JUN 86 ((8.37.D.406))
P.07.8.31678: IDDIVE REMOV4L SYSTEM IN0PER43LE - N40H TANK DILUTED - CWICK VALVES LEACE), SIGHT GLASS INACCURATE ST. LUCIE 1 - FES 86 - 130% P3WER I
P.07.C.3126 : RCFC MOT 04 H( N00$!N3 DRAIN CHECK VALVES INSTALLED BACKdARDS
^
IION 2 - NOV 85 - REFUELING i
e I)
P.07.D.3061 : CONTAINMENT ISOLATION VALVES LEAKED - SEAT. ADJUSTMENT, SItT, 3
JACKIN3 SCREd ADJUSTMENT ST LUCIE 1 - MAR 83 - REFUELING
/.
)
P.07.D.0063 : CIV LEAK RATES EXESSIVE - DEBRIS, SE4 TING SURFACE DE3RATI34, 3R0 KEN PLUNGER SPRIN3 $ PIN NORTH ANNA 2 - MAY 83 - REFUELIN3 J
P.07.D.0065 : CONT 4!NMENT 3YPASS LE4KAGE P4TH MECH 4NICAL PENETRAT!31 V4LVES
/
LEAKED - PERSONNEL ERROR, GENERAL WE4R, DIRTY COMTACTS, F3 REIGN MATERIAL 1CGUIRE 2 - JUN $ JUL 83 - COLD SHUTD3WN P.07 D.0073 : CIVS LEAKED - VARIOUS CAUSES i
JAN To MAR 83 - REFUELI4G CONN TANCEE (HADDAM NECK) l P.07.D.0076 : CIV LEAKAGE - FM, INTERNAL C3MPONENT FAILURE
--- PT 8EACH 1 - OCT 83 - REFUELIN3 P 37.0.3059 : EXCESSIVE CIV LEAKAGE DURING LLRT - SEAT DAMAGE, P4CKIN3 EAC
--- SALEM 1 - MAR-AUG 84 - REFUELI1G P.07.0.0095t EXCESSIVE cit LEAKAGE - V4RI3US CAUSES
---' CONN. YA1(EE (HADDAM 1ECC)
JAN 86 - SHUTDodN C)
P.07.E.0168 : FIRE PUMP FAJLTY - LEVEL SWITCH, CHECK VALVE CORRODED
--- MAINE YA4KEE - APR 83 - 100% POWER
- m. _
m PAGE 40 4
P.08.A.3689 : BAST'S OUT OF SPEC FOR 27 HR - AMd LEAKAGE, SAMPLING PR3CEDURES GINNA - SEPT S3 - 100E POWER P.05.A.3729 : CHARSING PUMPS Ig3PERABLE - DISCHARGE CHECK VALVE FAILED, FM IN DISCHARGE RELIEF VALVE gj SAN ONOFRE 3 - APR 84 - 100% POWER P.05.A.3747 : BORIC ACID C3NCENTRATIONS BELOW SPECS - CHECK VALVE LEACES
()
SALEM 2 - JUL 85 - 235 P3WER P.05.A.3757 : CHAR 3!NG PUM?$ GAS BOUN3, VCT GUTLET CHECC VALVE STUCC CL3 SED -
()
PULSATION DA1PENER BLADDER LEAKED PALO VERDE 1 - FEB 86 - 100% P3WER P.35.A.376A : ADDITION TO VIII.A.757 RE: CHARGING PUMPS GAS B0UND, VCT 30TLET PULSATION DAMPENER BLADDER LEAKED -
CHECC VALVE STUCK CLOSED MONTHLY PM.141TIATED
--- PALO VER3E 1 - FES 86 - 100% P0'JER P.05.8.3569 : C3NTAINMENT SPRAY RHR CHECK VALVE EXCEEDED ALLOWA8LE LEAK RATE -
j COOK 1 - AUG $3 - REFJELING P.03.8.0574 : CHARGING PUMP SW PUMP DISCHARGE PRES $URE PROBLEMS -
J INSUFFICIENT NPSH DUE TO SYSTEM DESIGN, CHECK VALVE SEATI1G FAILJRE & CL3GGED STRAINER $ DUE T3 F1 IN SW SURRY 1 $ 2 - JUN TO 3CT S3 - 103% POJER & COLD SHUTD3WN P.35.8.3609 : SEAWATER PUMP DISCHARGE CHECC VALVE FAILES - CORROSI31 CRYSTAL 11VER 3 - MAY 84 - 94% P3WER gj P.03.B.3634 : CCW $UPPLY CIV LEAKAGE EXCEEDED LIMIT - STICKING DISC - 53SHING HAD SHIFTED TOWARD Sd!NG ARM I
/
()
PT. BEACH.2 - APR 53 L OCT 84 - REFUELING P.03.8.0639 : CCW PUMPS TRIPPED - M3 TORS FLOODED - SW LEAKED THR3 UGH SW PIPE
/'
gg OPENING s
--- INDIAN PT. 2 - AUG 84 - COLD SHUTDOWN J
P.03.8.0648 : Sd REALIGNED TO STAND 8Y Sd P3ND - PUMPHOUSE PIT ISOLATI34 VALVE FAILED TO OPEN - TORQUE SdITCH SET L3Ws P3SITION INDICATI3N MALFUNCTIONES
--- CATAW84 1 - APR 85 - COLD SHUTD0dN P.08.C.3033 : POTENTIAL DA1 AGE TO REDUNDANT SAFETY EQUIPMENT FROM SAC (FLOW THROUGH EQUI
- MENT AND FLO3R DRAIN SYSTEMS CALVERT CLIFFS 1 & 2 --- PWR'S IN GENERAL - JUL 33 J
P 09.D.3297 : TURBINE / REACTOR TRIP - CONSERVATIVE DISCHARGE PRESSURE SWITCH SETP31NTS - FW IS3LATION VALVE FAILURE, ROD CONTROL ALARM, S3URCE RANGE DETECTOR FAILURE k
l
--- SUMMER - APR 84 - 12% P0 DER P.09.D.3334 : REACTOR TRIP - LOW SUCTI31 PRES $URE
- PNEUMATIC TRANSMITTER 00 FAILED, TURS!NE CONTROL SYSTEM DI3DE FAILED - SG PORV & TSAFP DISCHARGE CHECK VALVE FAILURES - PRESSURE INSTRUMENT II P.09.D.3355 : REACTOR TRIP DN HIGH SG LEVEL - Fd PJMP SPEED CONTROL FAILED -
CHECC VALVE STUCK - CONTROL DIL IMPJRITIES JATERFOR) 3 - JUL 85 - 55% P3 DER l
/
F"
)
PAGE 41 2
P.09.H.0761 : HYDR 3 GEN SAM
- LING PUMP SEAL JATER CHECK VALVE INTERNALS REQUIRED CLEANING C00K 1 S 2 - APR A JU1 83 - 133% POWER
'J P.07.H.0793 : C3NTROL.400M BOTTLED AIR PRESSURII!D SYSTEM PRES $URE 3EL3d gj LIMITS - HEA3ER LEAKAGE, COMPRESSOR FAILURE, INADVERTENT $1
--- 10RTH ANMA 1 S 2 - MAY, QCT 83 - 100% POWER P.07.H.3907 : SINGLE FAILUIE SEISMIC CRITERION 437 MET 3Y Z HR BACKJP
!ssTRUMENT AIR SYSTEM - DESIGN ERR 3R THREE NI.E IS. 1 - MAR 85 - C3LD SHUTDOJN
()
P.11.A.3683 : CHECC VALVE FAILURES IN 03 RAW WATER COOLING SYSTENS - INTERNAL DISASSEMBLY - ABtASION, C3RR35104
--- 8WR'S IN GENERAL - MAR 83 ((5.11.A.4933)
P.11.A.3792 : DGS INOPEt4BLE - CONSTANT VENT CHECK VALVE BINDING, DISTRIBUTOR
.HAD EXCESSIVE SU3HING & R3 TOR WEAR
--- FARLEY 1 & 2 - APR & AUG $3 - 103% POWER P.11.8.0561 : LOSP - TRANSFORMER FAILURE - CABLE FAULT. INSULATION DEGRADED -
AGES ENVIRO 1 MENTAL DAMAGE
--- PWR$ IN 3ENERAL - JUN 86 --- SAN ONOFRE 1 - NOV $5
((8.11.B.526))
P.15.A.0455 : PDTENTIAL FOR P! PING SYSTEM $ PIPE S'JPPORT DAMAGE - PIPE BENDING - TEMP DIFFERENCES INDUCED Bf STRATIFIED FLOJ
--- PWR$ IN 3ENERAL - DEC 84 (CB.15.3003)
Ih' P.15.A.0457 : ADDITION TO XV.448 RE: SNUBBER SELF-ALIGNING BALL SUSHINGS FAILED - SNU39ERS RESUILT ROBINSON 2 - MAY 84
/
g)
P.15.A.3455 : CONTAMINATIO1 0F 3REATHING AIR SYSTE15 - AIR COMPRES$3RS FAILED, RADICACTIVE GAS S FM IN AIR SURGE AT SYSTEM STARTUP'
/.
ROBINSON 2 - JUL S3 - COLD SHUTD3WN --- PWR$ IN GENERAL %
53 s
JAN $5 ((
'P.15.A.0468 : LIMITORQUE ACTUATOR WORM SHAFT GEAR FAILURES - IMPACT LDADING OF CLUTCH MECHANISM DURINS M3DE CHAN3ES AT HIGH SPEE3 1
PWR$ IN 3ENERAL - AUG 85 --- DTHER UNITS AS LISTED IN ARTICLE ((B.15.A.310))
P.15.A.3472 : LIMITORQUE M370R OPERATOR LITHIUM 3ASE LUBRICANT N3T ENVIRONMENTALLY QUALIFIED - CALCIU1 3ASE GREASE TO 8E USE3 SYRON 1 - JUL 85 - C3LD SHUTD3dN P.16.C.1890 : PORY FAILED TO FULLY STR0(E - INSTRU1ENT AIR CHECK VALVE INSTALLED SACKWARDS SURRY 1 - FEB 83 - 85% P3WER, SHJTDOWN g
,j P.15.C.2524 : AUX FW PUMP AUTO STARTS DURING CHECK V ALVE T ESTING - LOJ SG LEVEL - WRITTEN INSTRUCT!3NS NOT !$$UED - ADMIN /PR3CEDURAL' gg DEFICIENCY CATAW84 1 - N3V HOT STAND 5Y I) '
i P.15.C.2620 : CONTAINMENT SPRAY ACTUATI3N DURING TEST - INITIATION RELAf5 40T RESET - RCPS SECUREDs SEAL D414GE - CCW PIPING PRES $URE SPIKE
~~- dATERFORD 3 - FE8 85 - H3T CTANDSV
($
PAGE 42 P.15.C.2747 : ECCs TRAINS IN0PERABLE - CONTAINMENT RECIRC SUMP !$0LATIO1 VALVES NOT TESTED PR3PERLY - PER$31NEL ERROR
/
--- PALO VER3E 1 - APR 85 - HOT SHUTDOWN P.15.C.2313 : REACTOR, TUR3!NE TRIP, ESF ACTUATION 5 - HIGH MSR LEVEL - 1 EATER gj DRAIN TANC DRAIN PATHS IS3 LATED
---.CALLAWAY 1 - AUG 85 - 25% POWER P.16.C.2853 : REACTOR TRIPS - SPURIOUS CLOSURES 3F MSL EXCESS FLOW CHECC VALVES - RUPTURE DISC FAILURE 5 - 93U1 TING 83LTS UNDERT040JED
--- 1AINE YA1CIE - OCT 85 - 25, 331 POWER
()
P.16.C.2933 : $! DURING ACCUMULATOR DISCHARGE CHECC VALVE FLOW VERIFICATION -
DISCHARGE VALVES OPENED - $$PS FUSES REMOVED - PROCE33RAL, PER$3NNEL ERR 3R NORTH AN1A 1 - DEC 85 - COLD SHUT 80WM
............. ALL DONE, PRESS < RETURN > CEY TO G3 T3 MENU.............*..'..T THERE ARE 19 ARTICLE (S) T3 SE DISPLAYED:
19-AUG-1985 14:13:118 PWR - CHECC' VALVES - RCS SYSTEM j
P.05.e.0018 : RC MAKEUP LINE CRACK $, THERMAL $LEEVE DAMAGE - FATIGUE FAILURE SUSPECTED l
--- CRYSTAL RIVER 3 - JAN 82 --- 3C01EE 1, 2 5 3 --- PWR'S IN GENEaAL
'P.05.a.0020 : RCS CHECK VALVES LEACED 1CGUIRE l - FEB 82 - HOT SHUTDOWN I
P.05.C.3047 : P3RV STUCK OPEN - SOLEN 0ID OPERATED CONTR3L VALVE LOCCED IN ENERGIIED STATE Bf VENT RESTRICTID1 1
--- GINNA - JAN 82 - RAPID C33LDOJN 4
,/
()
P.05.C.0050 : PORVS INOPERABLE - LEAKIN3 DIAPHRA3M, AIR SUPPLY CHECC VALVE OCT 82 - SHJTD3WN
/
g
$URRY 1 P.05.C.0059 : RCS LEAKAGE - PORV BLOCK VALVE, SI CHECK WALVE, SG TU5E 2
1ILLSTONE 2 - MAR 83 - 130% POWER P.05.E.0063 : PORV LEAKAGE PROBLEMS, C01TR3L AIR ISOLATED, PCS VENT P AT4 L357 RAND 0M MAI1TENANCE PROSLEM - PORV LEAKAGE PROBLEM U1RES3LVED
--- SALEM 2 - JAN $3 - C03LD3WN, S4UTDOWN P.05.F.0004 : PROSLEMS WIT 1 VALVE OPERATING SYSTEM
--- SHIPPINGPORT - 1960'S P.05.F.0016 : CHECC VALVE INTERNALS REPLACEb WITH MORE DURABLE MATERIALS J
YANKEE R3WE - AUG 74 - REFUELING SHUTDOWN P.05.F.0043 : PRIMARY COOLANT EYSTEM LEAKA3E - CHECK VALVES FLUSHED T3
{
d IMPR3VE SEATING o
PALI $4 DES - JUL S4 - HOT STAND 3Y N
P.05.F0052 : PLANT SHUTDOJN - EXCESSIVE RCS LEACA3E - VALVE PACCING, RCP FLAN 3E LEACE)
--- CATAWar 1 - OCT 85 - 103% POJER
()
P.05.E.0594 : REACTOR TRIP DN MAIN FW PUMP TRIP - FW PU1P RECIRC VALVE FAILED TO HANDLE FL3J TRANSIENT - C3NTROLLER DESIGN PR05LEM - $$ PORVS 5}
FAILED TO OPEN - DUT 3F CALIARAT!31
l e
PAGE 43
--- CATAWBA 1 JUN 85 - 64% POWER P.37.A.3457 : UNIT SHUTDOW1 - LOW ACCUMULATOR 83R01 CONCENTRATION - LEACINS VALVES 1CGUIRE 1 - APR 85 - 331 POWER g
P.07.A.3474 : HIGH SI TANK LEVEL, L3W S3R01 CONCENTRATION - PRIMARf C30LANT LEAKAGE THROUGH SI CHECK VALVE
()
--- *ALISADES - NOV 85 - 981 POWER P.07.A.3455 : RCS NOT IS3L48LE FROM LP COOLANT SYSTEMS - PCIV LEAKAGE
()
--- PWR$ IN 3ENERAL - JUN 86 ((8.37.D.406))
P.07.D.3073 : CIVS LEAKED - VARIOUS CAUSE3 JAM TO MAR 83 - REFUELING CONN YANCEE (HADDAM N!CK)
P.07.D.0089 : EXCES$1VE CIV LEACAGE DURING LLRT - SEAT DAMAGE, PACCIN3 LEAC SALEM 1 - MAR *AUG 84 - REFUELING P.16.C.1297 : CHECC VALVE CL3SURE WEIGHT MISSING
--- NORTH AN1A 1 - MAR 81 - SHUTD0JN
'j P.16.C.1890 : PORV FAILED T3 FULLY STR0(E - INSTRU1ENT AIR CHECK VALVE INSTALLED SACCWARDS SURRY 1 - FEa 83 - 85! P3WER, SHUTDOWN P.15.C.2620 : CONTAINMENT SPRAY ACTUAT!3N DURING TEST - INITIATION RELAfS NOT RESET - PCPS SECURED, SEAL DAMAGE - CCW PIPING P4ES$URE $8IKE JATiRFOR) 3 - FE3 85 - H3T STAND 3Y g
............. ALL DONE, PR!$$ < RETURN) (EY.T0 GO T3 M NU..
7...
.........ft THERE ARE 95 ARTICLE (S) TD BE DISPLAfED:
19-4U3-1986 14:15:43 PWR-CHECCVALVES-SI/RECIRC/CONTfSPRAY
(
P.34.8.3089 : CAD SUIDE TU3E SUPPORT P!1 FAILED - SGS DAMAGED JITHIM M3JRS~
$3
--- JESTIN3H3USE PWR'S IN GENERAL - JUL 82 y
P.05.5.3018 : RC MAKEUP LIVE CRACKS, THERMAL SLEEVE DAMAGE - FATIGUE FAILURE SUSPECTED J
--- CRYSTAL RIVER 3 - JAN 82 --- OC01EE 1, 2 8 3 --- PWR'S IN GENERAL P.05.C.0059 : RCS LEAKAGE - PORV 8 LOCK VALVE, SI CHECK VALVE, SG TU3E j
11LLSTONE 2 - MAR 83 - 1303 P3JER
.e P.07.A.0013 : CHECC VALVE ASSEMBLED INC3RRECTLY.
PALISADES - MAY 72 P.07.A.0020 : INDICATIONS IN VALVES 1AINE YA1KEE - SUMMER 1972 (AFTER HOT FUNCTI3NAL TESTING) i 1,
J P.07.A.3025 : WELD SLAG UN)ER VALVE SEAT - SI TANK CONCENTRATION L3J
--- MAINE YA1(EE - DEC 72 P.07.A.0032 : CHECC VALVE SOFT SEAT FAILURES TURKEY PT 4 - MAY 73
)
P.07.A.0060 : CRACC INDICATIONS IN VELAN CHECK VALVES CALVERT CLIFFS 1 - SEP 73 (FUNCTIONAL TESTIN3)
o PAGE 44 P.07.A.0063 : CHECC VALVE LEAKAGE - ACCUMULATOR DILUTION
$1NNA - SEP 74 P.OT.A.0085 : CHECC VALVES D!D NOT SEAT, PROCEDURAL PROSLEM - 50 ROM CONCENTRAT!01 LOW e>
SURRY 1 - AUG 75 - HOT STAND 8Y P.07.A.0105 : LEAKING CHECC VALVE - ACCJMULATOR 30ADN L3W
()
--- ROBINSON 2 - J AN 76 - 103% POWER P.07.4.3114 : LEACING CHECC VALVES CAUSED 50RON DILUTIO1 IN SI ACCU 1DLBTOR
()
SURRY 1 JUL 76 - 103% POWER P.07,A.3120 83RON DILUT!3N IN SI ACCUMULATOR CAUSED SY LEAKING CHECC #ALVES j
SURRY Z - AUG 76 - 103% POWER P.07.4.0122 : SACK LEAKAGE IN ACCU 1ULAT3R DISCHARGE LINE CHECK VALVES IION 1 - JUN 76 - 50% POWER P.07.A.0126 : WRONG SIZE GASKET INSTALLED IN ACCUM*LATOR CHECK VALVE J
ZION 2 - 3CT 75 - 30% P0JER P.07.A.3143 : 19PR3PER SEATING OF 5!T DUTLET VALVES CAUSED DIL'JTION OF 34T'S j
--- ZION 2 - MAR 77 - 50% P04ER, APR 77 - 48% POWER P 07.A.3172 : 33LTS & BRACCET F3R CHECC VALVE DISC MIS $1NG j
--- INDIAN PT. 2 - MAY 78 (L3W POWER PHYSICS TESTS)
P.07.A.0175 : TILTING DISC CHECC VALVE INSTALLED I4 WRONG ATTITUDE gj SAN ON3FRE 1 - MAY 75 - APPR3XIMATELY 55% POWER j
P.07.A.0182 : 90 ROM CONCENTRATION LOW SI TANC CHECK VALVES'LEACE3
/
()
CALVERT CLIFFS 2 - SEP 75 - 98% POWER P.07.A.3187 : STOP CHECK DISC SPRING JAMME) - SPRI4G RETAINER ADDED
/'
g)
ARKANSAS ONE 2 - JUL 78 - PRE 3P TESTING y
P.07.A.3200 : CHECC VALVE LEAKED - VALVE 83DY OdERSIZED
~'
--- ARKANSAS ONE 2 - JAN 79 - 19% POJER (POWER ESCALATI3N TESTING)
P.07.4.3204 : INCORRECT VALVE WEIGHTS USED IN PIPE STRESS CALCUL4T!3NS -
-HANGERS MODIFIED
--- 40RTH ANNA 1 - MAR 79 - 92% POWER P.07.A.3225 : $1 TANK 80404 CONCENTRATI3N LOW - CHECK VALVE LEAKED MILLSTONE 2 - APR 77 - 99% POWER a
P.07.4.3229 : SAT 50RON C01CENTRAT!0N L3W - VALVE LEAKED - LIMIT SWITCHES READJUSTED
--- ZION 2
.3CT 79 - 94% POWER
{
J P.07.A.0230 : SWIN3 CHECK VALVE BEARING CAP 0-RING FAILED b
--- 1AINE YAMCEE - OCI 79 - 97% POWER P.07.A.3240 : HPI CHECK VALVE SEAT HOLD-00JNS F3JND LOOSE g)
THREE MILE IS 1 - FE3 30 - SHUTD3WN a.07.A.3256 : C3NTAINMENT PENETRATION CHECC VALVE SEATS LEAKED b)
--- CONN VANCEE (MADDAM NECK) - M4f 50 - REFUELING j
............. ~. _.
m PAGE 65 P.07.A.3260 : HPI VALVES LEAKED - P! PINS HEATED EXCESSIVELY 3CONEE 1 - JUN 83 - 73% POWER P.07.A.3262 : FAILED CHECK VALVE, 3 PEN ISOLATION VALVE
- 20 GAL RELEASE CRYSTAL RIVER 3 - JUL 80
- SHUTD3WN P.07.A.3266 : SIT RECIRC LINE LEAKED - dELD CRACKED - V4LVE CONTROLLER FAILED 91LLST3NE 2 - JUL 80 - 100% P3JER
()
P.07.A.3270 : CHECK VALVE STUCK OPEN - DISC WELD / VALVE SODY INTERFERE 1CE SEQUOYAH 1 - SEP 50
- HOT STANDSV
()
P.07.A.3271 :. SIT PRESSURE LOW - NITR03EN CONTROL VALVE FAILED ARKANSAS 3NE 2 - OCT 50 - 1001 P3WER P.07.A.0273 : CFT 3VERPRESSURIIED - CHECK VALVE IMPROPERLY ASSEMSLED DAVIS-BESSE 1 - OCT 50 - HOT STAND 8Y P.07.A.3275 : HIGH HEAD SI PUM* CHECK VALVE ANTI-R3 TAT!3N DEVICE BINDI43 DUE T3 FLOW OR PRESSURE SURGES
--- SEAVER VALLEY 1 - OCT 80 - SHUTD3WN P.07.A.D.
8 : ADDITION T3 #11.A.273 RE: CHECK VALVE
--- DAVIS-9 ESSE 1 - OCT 83 - HOT STAND 8Y P.07.A.3283 : SI TANK LEVEL LOW - 53LEN3!D VALVE LEAKED
--- CALVERT CLIFFS 2 - N3V 83 - 133% POWER P.37.A.3285 : SI SYSTEM CHECK VALVE STUCK 3 PEN SALEM 1 - DEC 80 - SHUTD3WN J
P.07.A.3291 : SI ACCUMULAT3R BORON DILUTED - CHECK VALVE LEA ED i
()
--- SURRY Z - JAN 81 - 103% POWER
/
52 P.07.A.3296 : LPI CHECK VALVE DISC C0CKED 8Y DEPOSIT BUILDUP s
--- 3CONEE 1 - FE3 81 - DE P3WER CHEATUP)
J P.07.4.3302 : ADDITION TO WII.A.296 RE: CHECK VALVE
--- 3CONEE 3 - P.AR 81 - SHUTDOWN P.07.A.0306 : ACCU 1ULAT04 DISCHARGE CHECK VALVES LEAKED
--- 9CGUIRE 1 - APR 81 (PRIOR TO I1ITIAL CRITICALITY) - H3T STAND 8Y P.07.A.3307 : COLD LEG INJECTION ACCUMULAT3R DISCH4RGE CHECK VALVES LEA <ED MCGUIRE 1 - APR 81 (PRIOR TO INITIAL CRITICALITY)
-H3T STANDBY P.07.A.3310 : LEAK TEST DA1 AGED UNI ACC'JMULATOR VALVES - SEAT TYPE CH4NSED 1CGUIRE 1 - 9AY 31 (PRIOR To I1ITIAL CRITICALITY) - C3LD SHUTDOWN b
P.07.A.3311 : ACCU 1ULAT04 CHECK VALVE FAILED - 83TH RHR TRAINS IN0PER4BLE IN COLD SHUTDOW1 1CGUIRE 1 - MAY $1 (PRIOR TO INITIAL CRITICALITY) - C3LD
)
SHUTDOWN P.07.A.3315 : SI CHECK VALVES STUCK OPE 4 i
--- PT BEACH 1 - JUL 81 - SHUTDOWN
)
PAGE 46 P.07.A.3316 : ADDITION To WII.A.243 RE: TILTING DISC CHECK VALVES THREE MILE IS 1 - Jug 81 - COLD SHUTD3WN P.37.A.0339 : CHECK VALVE 31SC MIS $1NG - C3NSTRUCTION ERROR gj COOK 2 - OCT 81 - SHUTD0JN P.07.A.3341 : SI ACCUMULAT3R BORON CONCENTRATION HIGH - VALVES LEACED (3
PROCEDURES REVISED NORTH AN14 1
- N3V 81 - 79% PDJER P.07.A.0343 : LEAKING SI SYSTEM CHECK VALVES SE4TED FOLLOWING FLUSHINS
--- PT 8EACH 1 - DCT 81 - REFUELINS P.07.A.0345 : $1 ACCUMULAT3R 80RON CONCENTRATION HIGH - VALVES LEACED, FAULTY PROCEDURE
--- NORTH AN14 2 - OCT 81 - 97% POJER P.07.A.0346 : 4DDITION To dII.A.331,335 & 337 RE: 4CCUMJLATOR 5040N CONCENTRATI015
--- NORTH ANNA 1 & 2 - SE* 81 - 133% POWER P.0T.A.0352 : SIT INLET VALVE FAILED - SHAFT PACCING T03 TIGHT
~j
--- SuaRT.1 - DEC 81 - COLD SHUTDCJN P.07.A.0363 : LPSI SWING C4ECK VALVE VELR, DAMAGE - DESIGN y
--- PALISADES - SEP $1 - REFUELING P.07.A.0372 : HP 11JECTION STOP CHECK VALVES STUCK - UP TO 580 PSID REQJIRED TO ESTABLISH FLOW
--- DAVIS-9 ESSE - JUN 82 - REFUELING c
I)
P.07.t. 3375 : ADDITION TO WII.A.372 RE: HP INJECTI3N ST3P CHECK VALVE STUCK -
INCORRECT SEAT ANGLE
/
DAVIS-3 ESSE - JUN 82 - REFUELING --- PWR$ IN GENERAL - AUG GI
$2
.s P.07.A.3384 : SIT 3UTLET C4ECK VALVE LEAKED RINGS DETERIORATED J
--- CALVERT CLIFFS 1 - JUL 82 - STLRTUP P.07.A.0386 : COLD LEG INJECTION CHECK WALdE LEACED - CAR 80N STEEL CL3SJRE STUDS CORRODED j
NORTH ANN A 1 - SEP 82 - REFUELINS P.07.A.0390 : ADDITION TO VII.A.275 RE: VELAN CHECC VALVE BINDING PROBLEMS -
DISCS REPLACED 3EAVER V4LLEY 1 - OCT 82 - SPECIAL TESTING
.)
P.07.A.0392 : 51 CHECK VALVES STUCK - DISC STUD PR3TRUDED A80VE NUT, DISC MISALIGNED 4RKANS45 DNE 2 - OCT S2 - REFUELING P.07.A.0396 : MINOR LEAK 4GE INT 3 $!T C01 POUNDED BY $1T LEVEL 21DIC4 TING g
SYSTEM FAILU4E CAUSED TECH.PEC LI9ITS TO 8E EXCEEDED - V4LVE LEAK 4GE, TEMS EFFECTS ON REFERENCE LEG PALISADES - SEP-DEC B2 - 100E POJER
()
P.07.A.3397 : SI T4NK LEVEL LOW, BORON CONCENTR4T!3N LOW - VALVE LEAK 4GE
--- PALISADES - NOV 82 - 1005 POWER gg ;
I
~~
PAGE 47 I
P.07.A.3400 : ADDITION TO WII.A.399 RE: CHECK VALVE DISC MISSING COOK 2 - DEC 82 - REFUELING P.07.A.3401 : SI C4ECC VALdE SEAL dELD LEAC - INADEQUATE APPLICATI3N 3F dELDING AND GRINDING TECHNIQUES g) 9AINE YA1CEE - FEB 83 - SHUTD3dN P.07.4.0403 : ACCU 1ULATOR 30R01 CONCENTRATION L3d - CHECK VALVE LEA (ED
()
SURRY 2 - SEP $2 - 103% POWER P.07.A.3407 : COLD LEG INJECTION ACCUM3 TAT 3R CHECK VALVES LEAKED
()
1CGUIRE 1 - MAY $3 - STARTUP, 35 5 501 POWER P.37.A.0420 : HPI STOP CHECK VALVE STUCC - DISC TO SEAL CONTACT AREA T33 WIDE
'j
- DESIGN DEFICIENCY l
--- DAVIS-3 ESSE - SEP 83 - 5HUTD3d4 1
P.37.4.3422 : ADDITION TO WII.A.384 RE: SIT OUTLET CHECC VALVE LEACED I RINS$ DETERI3 RATED - 1ATERIAL CHAN3ED
--- CALVERT CLIFFS 1 8 2 - JUL 82 - $TARTUP P.07.A.3437 : SI CHECC VALWE LEAKINS - INc3MPLETE CONTACT BETWEEN DISC 1 $ EAT
--- FARLEY Z - SEP 83 - REFUELING P.37.A.3447 : VELAN CHECK WALVE ANTI-R3TATION STOP JAMMING - DESIGN P403LE1 DAv!$-BESSE - OCT 84 - REFUELING j
P.07.A.0452 : EXCESSIVE SIT CHECK VALVE LEAKAGE - SEAL PLATE C3CKEDa WALVE SE AT COMPENS ATING JOINT 3 ALL G ALLED - FM
(
--- ST. LUCIE 2 - DEC 84 - H3T STAND 3Y P.07.A.3456 : EXCES$1VE SIT CHECK VALVE LEAKAGE - ETHYLENE PROPYLENE 3-RING f
)
MATERIAL DE3RADATION
--- CALVERT CLIFFS 1 - JA1 85 - 103% POWER
/
C P.07.A.3457 : UNIT $HUTDOW4 - L3W ACCUMJLATOR 80401 CONCENTRAT!0N - LEACING s
VALVES
--- 1CGUIRE : - APR 85 - 33% POWER j
P.07.A.0468 : LPSI PUMP, $! TANC INOPERABLE - MAINTENANCEr CHECE VALVE LEACAGE PALISADES - N3V 85 - 78% POWER A
P.07.A.3474 : HIGH SI TANC LEVEL, LOW 8340N CONCENTRAT!3N - PRIMARf C33 ANT LEAKAGE THROUGH $1 CHECK VALVE PALISADES - N3V 85 - 78% POWER P.07.A.3485 : RCS NOT 153LASLE FROM LP COOLANT SYSTEMS - PCIV LEAKAGE
--- PWR$ IN SENERAL
- JUN 86 ((8.37.0.4063)
P.07.8.0012 : CRACKED ROCK! HAFT 5 AND CRACKED BODY IN CHECK VALVES
--- $URRY 1 1 2 - NOV 74 k
J P.07.8.0053 : CTS 10!!LES LEAKED - VALVES NOT SHUT TIGHTLY.- CHECK VALVES N
INSTALLED SACCWARDS
--- C00c 2 - NOW 78 - sHurD0JN I)
P.07.8.0068 : SPR4f PUMP M3 TOR STAT 3R C3!LS/ PHASE LEAD CHAFED, $HORTE)
--- CALVERT CLIFFS 1 - SEP 77 - 100% POWER kh P.07.8.0080 i CHECC VALVE DISC AND RELATED PARTS N3T INSTALLED
PAGE 48 J
SEQUOYAH 1 - APR S0 - PREOPERAT!3NAL s
P.07.B.3036 : SJST LEVEL L3d - RBS SYSTEM VALVE FAILED TO RESEAT, ALARM FAILED OCONEE 3 - APR 83 - NEATUP t)
P.07.B.3123 : CONTAINMENT SPRAY ADDITIVE TANC CHECC VALVES LEACED RANCHO SECO - APR 82 - SHUToody P.07.8.0131 : CONTAINMENT SPRAY.S NA0H 4DDITIVE SYSTEM CHECK V4LVE FAILED -
SEAT & DISC 14CHINED
--- SINNA 1 - SEP S2 - 103% POWER C)
P.07.8.0134 : SPRAf PUMP CHECK VALVE PI4/$ DING ARM WORN 3INNA - JUN 82 - 100% P0 DER P.07.8.314'2 : CONTAINMENT SPRAY PUMP DISCH4RGE CHECK VALVE F'AULTf - PIN / SWING ARM JORN
--- 31NNA - JUL 82 - 100% P0 DER 4
P.07.8.31678: 10 DIME REMOV4L SYSTEM IN3PER:2LE - N40H T4NK DILUTED - CHECK VALVES LEACE), SIGHT GLASS IMACCUR4TE
- l'
--- ST. LUCIE 1 - FES 86 - 130% P3WER P.37.D.30758: EXCESSIVE cit LE4 CAGE - V4R13US CAUSES
--- CONN. YA1CEE (HADDAM 1ECC) 44N 86 - SHUTD0dN P.07.E.3237 : CORR 3SION FAILURES OF 413 SS VALVE SFEMS - IMPROPER HEAT l
TREATMENT
--- PWR$ IN SENERAL - JUL 85 --- 3C01EE 1 - DEC 71 --- FARLEY 1 -
U FE5 54 ((9.37.F.2713) j P.03.A.0230 : CHECC VALVE LEAKED - RELIEF VALVES, FLOW TRANS'MITTERS D4M4GED -
[
g)
CHECC VALVE REWORCED
--- DAVIS-5 ESSE 1 - JUL 77 (LFTER INITIAL FUEL L34 DINS) - HOT f
STANDBY P.05.A.0459 : WRONS INTERNALS USED IN VALVE NORTH 4N44 2 - SEPT 83 (PRIOR TO POWER OPERATION)
C3LD
$HUTDodN
-)
P.03.8.0265 : WELD CRACr5 FOUND FOUND I4 83 RATED W4TER PIPING
--- 3AN ONOFRE 1 - SEPT 77 - 100% PodER 4
I P.03.C.0033 : POTENTIAL DA14GE TO REDUNDANT SAFETY EQUIPMENT FROM SLCCFLOW THROUGH EQUIPMENT AND FL33R DRAIN SYSTEMS s
--- CALVERT CLIFFS 1 & 2 --- PWR'S I4 GENERAL - JUL 33 9.07.H.3798 : CONTROL ROOM 80TTLED AIR PRESSURIIED SYSTEM PRESSURE SELQJ J
LIMITS - HEADER LEAKAGE, COMPRESSOR FAILURE, INADVERTENT SI
--- NORTH ANIA 1 5 2 - MAT, 3CT 83 - 100% POWER
{
P.15.C.1776 : SAAT PUMP DISCHARGE VALVE FOUND CLOSED - PROCEDURE ERROR DAVIS-SESSE 1 - MAY 82 - COLD SHJTDOW1 P.15.C.2747 : ECCS TRAINS INOPERABLE - CONTAINMENT RECIRC SUMP !$0L4 TID 1 VALVES N3T TESTED PR3PERLY - PERS31NEL ERROR 84LO VERDE 1 - APR 85 - HDT SHUTDOWN P.16.C.2933 : $1 DJRING ACCUMULATOR DISCHARGE CH!CC VALWE FLOW VERIFIC4i!OV -
IIl P40CE33 RALE DISCHARGE VALVES OPENED * $$PS FUSES REMOWED
3 s
C'O O
3 3
a o
O a
o o
~
- Ib 4
W L1 air
,_e
- 9 De M
0 e
e O
e 0
s 0
8 4
0 3
0 0
S M
Z 3
W O
E o
>=
M 3
>=
T M
M L1 e
O
.J O
O b=
u W
w tr%
E3 A
Z u
m W
3 En
>=
w raf V
e M
O at M
M F w
KZ M
WW JE w >=
W 2&
Z ZO O
n7 n
M M $
J w4
.J
& 8 W
i 0
e 9
I 8
0 8
0 0
0 I
t 0
i 6
i
e e
d*
D e
e" g
h
- e b
ATTACHMENT B
't 9*
4 l
.j i
NPE I
PLANT I
PLANT I
i EVENT I FAILURE l I NUM3ER I
(DATE)
I STATUS 1 SYSTEM I DESCRIPTION I MODE I
I I
I 8
I I
I I
I I
I I
I I
I V.D.20 3 McCuire I Hot 1 RCS I Three RCS pressure isolation check IL I
I I (02/62)
I Shutdown i
I valves failed leak rate test (failed to i I
d I
I I
I I seat properig).
I I
i I
I I
I I
I I V.C.59 I Millstone 2 1 100% Power 1 81 i Metal gasket replaced on one check IL I
O I
I (03/83) i I
I valve.
I I
I I
I I
I I
I I V. C. 50 i Surry 1 1 Shutdown I PZR, 1 Air suppig check valve for PORVs IL I
O I
I (10/82) i I
i leaked.
I I
g.
I I
I i
i i
l I VI.D.52 i Surry 1 1
I RCS I The check valve in the bypass line IL i
J l
I (11/82)
I I
I around the trip valve of the main i
I l
I I
I I steam line C leaked--flaw in the carbon !
I I
I I
I I steel valve body.
I I
I I
I I
I I
i I VI.D.55 I Kewaunee 1 Construc-1 MS I Cracks in isolation check valve seats IL i
I I (Late 1972)
I tion i
I (two valves) (discovered during i
i I
i I
I I gua11tg records audit).
I l
8 I
I I
I I
l l VI.D.55 8 Zion I Construc-I MS 1 Cracks in two of the four MSI check IL i
I I (Earig 1973)
I tion i
I valves.
I I
I I
I I
I I
I I VI.D.60 t Zion I Cold 1 MS I Weld cracks in main steam check valve iL I
I I (09/73) l Shutdown i
I disc stops (three of twelve valves).
I I
1 I
I I
I I
I I VI.D.73 1 Maine Yankee I
I i Various problems in Shutte & Koerting iF i
W I
I (Late 1973) i I
I I
8 I
I I
I l
,'l I VI.D.73 i Surry 1 I
I I Various problems in Shutte & Koerting iF I
I I
I I (1973)
I I
I I
I I
I i
/
I I VI.D.78 I Fourt Calhoun I Power i MS I Thin-walled check valves.
1-I I (04/74)
I I
I I
I.
O f
1 I
3 I
I I
I line check valve shaft.
I-I s
I VI.D.148 I Trojan I Cold i MS I Main steam J
i I (08/76)
I Shutdown i
I I
I B
I I
I I
I I
I VI.D.264 I Robinson I Refueling I MS I Check valve rock shaft displaced 1-1 I
I (06/82) i I
I asially-set screw loosened.
I l
8 I
I I
I I
I I VI.D.274 i Segungah I Refueling i MS I Check valve disc / stem separation IF 1
J I
I (11/82) 1 I
I (cyclic fatigue failure).
I I
l I
I I
l I
I I VI.D.294 I
I I
I Additional information to VI.D.2740K.
IF i
I I
I I
I I
I I VI.D.290 1 Trojan I?
I MS I Main stream nonreturn check valves iF I
I I (03/83)
I I
I stuck open due to increased friction i
I I
I I
8 I due to overtightening of the packing I
I I
I I
I I gland (four valves affected).
8 J
l I
8 I
I I
b I VI.D.303 I North Anna I 0% Power i MS I Steam generator main steam trip valve
- I FC I
I I (09/83)
I I
I closed slow-corrosion on control I
I I
I I
I I venting check valve.
I I
O I
I I
I I
I I
Page 1
m
I VI.D.305 I TroJcn I-I-
1 Additian to VI.D.290.
IF 1
I i
1 1
I I
I I VI.E.22 i Surry I Power i FW I Check valve gasket fa!!ed (interesting IL I
I I (10/72) i I
I event).
I i
I I
I I
I I
I VI.E.29 I Surry 2 I-1 FW I Check valve in feedwater line a leaked IL i
I I (05/73) 1 I
I through or flaw in the valve body i
I I
I I
I I (small leakage).
I I
I i
8 8
8 I
i i VI.E.106 4 Palisadas t?
8 FW l Feed water check valve gasket leaked.
I FD I
g i
I (<1974)
I i
I I
i I
1 l
i I
I I VI.E.128 I Robinson i 100% Power I FW I Check valve fa!!ed to hold? (Burr on iF I
O I
I (08/77) 1 I
I hinge prevented proper operation).
1 I
I I
I I
I I
I I VI.E.155 l San Onofre 1 i Startup i FW I Feed water check valve pin broke I FC I
O I
I (04/76?
I I
I allowing the disc to fall into the I
I I
I I
I bottoming of the valve.
I I
I I
1 1
I I
I 2
I VI.E.171 i Trolan i Hot i FW I CST check valve stuck open (no cause i le 1
I I (07/78)
! Shutdown i
I identified).
I I
I I
I I
I I
I I VI.E.185 I Fourt Calhoun 1 Refueling l AFW I Check valve found to be oriented in I FC I
~
I I (11/78)
I I the wrong position.
I 1
I I
8 1
I I
I I
I I
I I
I I VI.E.278 8 Crystal River 31 Cold i FW 8 Check valve failed to close--steam IL I
I I (04/80) i Shutdown i
I generation feed water check valve dise 1 I
I I
I I
I retainer pin missing.
I I
1 I
I I
I I
I I VI.E.286 I Oconee 2 8 48% Power i FW I Steam leaked from feed water check IL I
W I
I (06/80)
I I
I valve.
I I
l i
I I
I I
.'I I VI.E.319 8 Cook 2 8 72% Power i FW I Feed water check valve leaked (seat i FC 1
O I
I (01/81)
I I
8 leakage).
I I
I I
I I
I I
/
I I VI.E.325 i Cook 1 1 100% Power i AFW I Auxiliary feed water check valve stuck I L 1
O 1
I I (02/81) 1 I
I ope (caused by dirt).
I f 1
I I
I I
I I
valve lea k ed-s ea t/
IL I
y O
I VI.E.331 I Sequoyah 1 8 Cold I FW l Feed water check i
I (03/81)
I Shutdown I disc required lapping.
1 I
I I
I I
I I
I I VI.E.340 i Turkey Point 3 I Refueling i NFW I Main feed water check valve disc stud iF I
s I
I (04/01)
I I
I nut missing due to the failure of its I
l I
I I
I associated locking device.
I I
I I
I I
I I
I J
I VI.E.362 I Davis Besse 1 i Startup I FS I Auxiliary feed pump check valve leaked. IL i
l I (03/00) i I
I I
I I
I I
I I
I I
I VI.E.370 3 Point Beach I >79% Power i FW I Feed water pump discharge check valve IL 1
I I (11/80) i I
I failed.
I I
I I
I I
I I
I I VI.E.378 i Surry 1 i Startup I FW I Check valve Feed water-89 disc I FC I
I I (04/80) 1 I
I retaining nut cotter pins f ailed-valve !
J l
I I
I I stuck open (improper material).
.I I
N I
I I
I i
I I
I VI.E.378 i Surry 1 i Startup i FW I Check valve feed water-27 disc was iF 1
I I (01/80)
I I
I found detached (cause unknown).
I l
8 I
I I
I I
I O
Pase 2
I VI.E.38le
! N2rth Anna I 100% Pawsr i FW l Steem suppig check valve pceted ledgsd iF I
I I (12/C1)
I I
I in sus FW purp trip valvs--disc nut I
I I
I I
I I retaining pin installation suspect.
I I
I I
I I
I I
I J
I VI.E.398 I Cook 1 I 100% Power i FW I Aust11ary feed water pump leakoff line i FC I
l I (04/82) i I
I check valve failed to seat completely.
I I
I I
I I
I I
I I
I VI.E.402 I Turkey Point 4 i Startup i FW l Feen water check valve stud nut iF I
I I (06/82) 1 I
I fastener fatted.
I I
I i
I I
I I
i g
i I
I I
I I
I I VI.E.412 i Arkansas One 2 i Refueling i EFW 8 EFW pump turbine steam suppig check iF l
8 I (10/02) i I
I valve failed--the disc stud was broken i I
g i
I I
I I and missing.
I I
I I
I I
I I
I I VI.E.417 i Salem 2 I B2% Power i EFW I EFW pump steam supp1g check valve disc IF I
J I
I I (10/82) i I
I stud broken. nut and cotter sin f
i l
I I
I I missing.
I I
I i
l i
i I
I J
I VI.E.444 I Crystal River 31 Refueling i FW 8 Feed water air accumulator check valve 1L I
l I (03/83)
I I
I leaked due to water in the instrument 1
l I
I I
I I air system.
I I
J I
I I
I I
I I
I VI.E.481 I Farley 1 1 100% Power i FW l NDAFWP discharge check valve leaking IL 1
I I (12/83) l I
I hinge pin bushings missing, worn.
I l
J 8
I l
l l
8 l
l VI.E.495 I NcCuire 2 3 25% Power i FW I So inlet check valve stuck open (cause i FC i
i I (05/04)
I I
I not known).
I I
I I
I I
I I
I I
I VI.E.524 i Millstone 2 I 62% Power i FW l Binding check valve stem-f eed water I FC E
I I
I I
I check valve to SC 2 failed to seat.
I I
SI I
I I
I I
I E
I VI.E.524 3 Millstone 2 8 62% Power i I Binding check valve stem.
IF
,'t I
I (11/84) l I
I I
I O
I I
I I
I I
I I VII.A.32 1 Turkey Point 4 1-I SI' I Seat failure (safety injection high IL /
l 1
I (05/73)
I I
I head lines to RCS cold leg)- "0" ring I
I O
f I
I I
I I missing ( ma x.
leak 1/3 gpm)
I.f I
I I
I I
I l
I VII.A.60 I Calvert I-I SI l Crack in check valve body.
IL I
y J
I I Cliffs 1 I
I
'l I
l l'
I I
I I
I I VII.A.63 I Cinna 1 10%/Hr.
I SI I Check valve in loop B cold leg IL I
s I
I (09/74)
I Power Red. I I leak ed-accumulator dilution.
I I
I I
I 1
8 8
I I Check valve associated with 8 FC I
t VII.A.85 I Surry 1 I Hot I SI s
I I (08/75) i Shutdown i
I accueulator IC did not seat.
I I
I I
I I
I I
I I V11.A.105 i Robinson 2 8 100% Power I SI l Check valve associated with IL I
J I
I (01/76)
I I
I accumulator B t ea k e d.
8 I
l I
i l
I I
I I VII.A.114 1 Surry 1 1 100% Power I SI I Two check va'ves associated with IL 1
s I
I (07/76) i I
I accumulator B leaked.
I I
I I
I I
I I
I VII.A.120 i Surry 2 8 100% Power 1 SI I Two check valves associated with IL I
U I
I (08/761 I
I
-I accumulator C leaked.
I I
l i
I I
I I
I I VII.A.'126 I Zion 2 1 30% Power i SI 1 Wrong site gasket installed in IL I
O I
I (10/75)
I I
I accuiulator check valve (0.25 gpm) 1 I
P.go 3
i
- -- --. l
. -. _. -. ~.. -
I I
I I
I (factary escochly serse) (m).
I I
l I
I I
I I
I I
8 VII.A.128 I Zion 1 1 50% Power i SI I Accumulator 1D discharge !!ne check iL I
I I (06/76) i I
I valve leaked.
I I
I I
I I
I I
I d
i VII.A.172 I Indian Point 2 i Low Power i SI l #21 HH SI discharge check valve IL 1
I I (05/78)
I Test 8
I lea k ed--b ol t s and brackets for disc I
I J
I I
I I
I missing.
I I
l I
I I
I I
I I VII.A.175 i San Onofre 1 55% Power I SI I Tilting disc check valve installed in i FC 1
O 8
I (05/78)
I 1
I vertical rather than horizontal pipe I
I I
I I
I I line - failed to close.
I I
I I
I i
i I
I O
I VII.A.182 I Calvert I 98% Power I SI I Safety injection tank check valve IL 1
I I Cliffs 2 I
I I leak.
I I
I I
I I
I I
I
-)
I VII.A.187 I Arkansas One 2 I Pre-op i LPSI I Check valve on pump discharge failed to 1F I
I I
I Testing i
i close (disc spring Jammed).
I I
I I
I I
I I
I J
I VII.A.200 l Arkansas One 2 1 19% (power I SI I Check valve body oversighted.
IL I
l i
I escalation i I
I I
I I
I testing)
I I
I I
J I
I I
I I
I I
l VII.A.240 I Three Mile i Shutdown l HPI 1 Seat hold-down found loose.
IL l
I I Island 1 1
I I
I I
J I
I (02/80)
I I
I I
l I
I I
I I
I I
I VII.A.236 I Haddam Neck I Refueling i HPSI I Containment penetration check v41ves IL i
J I
I (05/80)
I I
I leaked
(> technical specifications) due i I
l I
I I
I to uneven seating surface component i
I I
I I
I I cooling water check valve.
I I
SI I
I I
I I
I I
I VII.A.260 I Oconee 1 1 73% Power i HPI I
I
,'t I
I (06/80) i I
I I
I O
I I
I I
9 I
I I
I VII.A.262 I Crystal River 38 Shutdown i SI' I Core flood check valve failed (cause iL /
I I
I (07/80)
I I
I core flood /n2 check valve isolation i
I O
I I
I I
'I unknown).
I f l
I 8
8 1
I I
t I VII.A.273 1 Davis Besse I list i SI I Cross back leakage through cone flood IL I
?
J I
I (10/80)
I Shutdown i
I check valve n200 (unknown)- valve i
I l
i I
I I improperly assembled.
I I
I I
I I
I I
I J
I VII.A.275 I Beaver Valleg i Shutdown I SI I HH SI pump check valve antirotation IF I
I i (10/80)
I I
I device binding due to flow or pressure I I
I i
i i
I surges (check valve did not reseat).
I l
J 8
I I
I I
l I
l I
I I
l l
I I VII.A.270 i Seguovah I Hot i
I I
I I
I (09/80)
I Shutdown i SI I Check valve stuck open disc weld /
8L I
~*
I I
I I
I valve body interference.
I I
I I
i l
i i
I I VII.A.230 t Maine Yankee 1 97% Power 1 St I O-ting failure on the bearing cap of IL I
I I (10/79)
I I
I the swing check valve.
I l
l 8
I I
I I
I b
I VII.A.285 I Salem 1 i Startup i SI I Check valve in the interference between i FC I
j I
I (12/80) i I
I the RCS Hot Log II and Safety Injection i I
i 8
I I
I pumps failed to close during test.
1 I
O 1
I I
I I
I I
l 0
Page 4
l
I I
I I
I I
i I VII.A.291 I Surry 2 8 100% Pcw3r 3 CI I SI cccurulette check valve Isaked IL 1
I I (01/81)
I I
I through causing boron dillution in i
I I
I I
I I accumulator (normally closed manual i
I I
I I
I I valves were left open intentionally i
I I
I I
I I making the check valves value the i
I I
I I
I I interface).
I I
I I
I I
I i
1 3 VII.A.294 8 Oconee 1 8 0% Power I LPI I Check valves leaked excessively = disc i L 1
I I (02/80) l I
I cocked by deposit butidup (leaking i
i g
I i
l I
I valve was the final value in LPI loop B I I
I I
I I
I before reaching reactor vessel).
I I
l I
I I
i 1
1 b
I I
I I
I I
I I VII.A.306 8 McCuire i Power i SI I Accumulator discharge check valves IL 1
I I 404/01)
I I
I peaked (two values on cold leg i
I s
I I
I I
I injection accumulator al.
1 I
1 I
I I
I I
I I VII.A.307 i McCuire i Power i SI I Discharged check valves associated with IL I
J I
I (04/81) 1 I
I accumulators d and c leaked.
I I
l i
I I
I I
I I VII.A.311 l McCutre i Power i SI I Accumulator check valve failed IF 1
J I
I (05/81)
I Cold I
i (both RHR trains inoperable in cold i
I l
i I Shutdown i
I shut down).
I I
I I
I I
I I
I I VII.A.315 I Point Beach i Startup I SI I SI check valves I-853 e and d.(unknown) IL I
I I (07/81) i I
I (first check valves between the RCS I
I I
I I
I I and the low head SI core) leaked i
I I
I I
I I excessively.
I I
I I
I I
I I
I I VII.A.316 i Three Mile I Cold I MU i Addition to VII a.240 re.
1-1 8'
I I Island 1 1 Shutdown i
I tilting disc check valves.
I I
I I (06/81)
I I
I l
I I
I I
I I
I I
O I VII.A.339 i Cook 2 3 Startup i SI.
I Check valve disc missing - Construction IL I
I I (10/81) i I
I error (discovered during valve test).
t
/
I I
I I
I I
I I
0 I
I I
I I
I 8
y i VII.A.343 I Point Beach I Refueling I SI I Check valve leaked during ten. (first iL I
I I (10/81)
I I
I check valve from RCS for (minor) low I
I y
i I
I I
I head SI cone deluge line).
I I
i I
I E
I I
I I VII.A.363 I Palisades i Refueling I LPSI I Excessive wear in swing check valve.
IL I
I I (09/81)
I I
I I
I 1
I I
i I
I I
I VII.A.372 I Davis Besse i Refueling i HPI I HP Injection stop check valve stuck lF I
e I
I (06/82)
I I
I (up to 590 pst required to establish l
I i
I I
I I flow) cause--incorrect seat angle.
I I
I I
I I
I I
I I VII.A.375 I Davis Besse I Refueling i HPI I Addition to VII.A.372.
I-I I
I I
I I
I I
leaked IL I
I VII.A.384 i Calvert i Startup I SI I SI tank outlet check valve I
I Cliffs 1 I
I I ring deteriorated (200 gym).
8 I
1 I (07/82)
I i
I i
(f I
I I
I I
I
'l b
i VII.A.386 I North Anna 1 I Refueling a SI I Cold leg injection check valve leaked-IL I
I I (09/92)
I I
I - carbon steel closure studs corroded.
I I
I I
I I
I I
I I
I I
I I
I I
I 0
Page 5
i l
l 8 VII.A.370 t Esaver Valleg il Special i HPI I Additian to VII.A.275.
I-I I
I (10/02) i Testing i
I I
I I
I I
I I
I I
I I
I I
I I
I I VII.A.392 I Arkansas One 2 I Refueling i SI
- I Two safety injection valves stuck in i FC I
I I (10/82)
I I
l the open position - two different i
l I
I I
I I
I I causes.
d I
I I
i I
l i VII.A.401 1 Maine Yankee I Startup i SI I Safety injection check valve seal weld IL I
I I (02/83) i I
I leaked (check valve in the loop 1 RCS I
I I
I I
I I safety injection line closest to the 1
I I
I I
I I cold leg).
I i
I i
i I
O i VII.A.403 i Surry 2 1 100% Power I SI I Accumulator check valve leaked.
IL 1
I I (09/92) i i
l I
I I
I I
i 1
1 I
J I VII.A.407 I McCuire 1 i Startup i SI I Two cold leg injection accumulator iL I
l I (05/83)
I 358.50%
1 I check valves leaked (unit was shutdown I I
I I
I Power I
i to repair valves).
1 I
J I
I I
I I
I I
1 I VII.A.420 I Davis Besse I Startup i SI I HPI stop check valve stuck-disc to IF I
I I (09/83)
I I
I seal contact area too wide--design i
I 2
I I
I I
I deficiency. Value opened on 3 to 4 psid 1 I
I I
I I
I instead of 100 psid.
3 l
8 8
8 I
I I
i J
! VII.A.422 8 Calvert i Startup i SI I Addition to VII.A.384.
I-1 B
l Cliffs 1 I
I I
I I
I I (07/82)
I I
I I
I J
I I
I I
I I
I l VII.A.437 I Farleg 2 8 Refueling i SI I RCS pressure isolation leaked--
IL i
I I (09/83)
I I incomplete contact between disc and 1
I 4'
I I
I I
I seat.
I I
I I
I I
I I
.'l I VII.A.452 1 Saint Lucie 2 I Hot i SI I Safety injection tank check valve IL i
O I leaked.
Seal plate cockeds valve seat i I
l I (12/84)
I Standby I
f 8
I I
I I compensating Joint ball galed.
I
/
I l
5 8
I I
i I
I O
l VII.A.456 I Calvert Cliffs i 100% Power i SI I Escessive SIT check valve leaked due to I1 I
(
I i 1 and 2 I
I I O-ring patorial degradation, t
- 9..
I I
I (01/85)
I I
I (Unit 1 = 1.6 gym - Unit 2 =27.2 gpm.)
I I
s J
I I
I I
l 8
l VI.E.462 i Surry 2 8 100% Power 1 AFW I Steam cut setse general check valves.
IL I
I I
I I
I I
I I
J f
I VI.E.412 l Arkansas One 2 i Refueling i EFW I EFW pump turbine steam suppig check i FD I
I I (10/82)
I I
I valve falied--internal damage.
I I
I I
I i
I I
i l
i I VI.E.415 i San Onofre 2 I Preop i MFW I Low N2 actuator pressure on steam 1L i
l I (09/82) 1 Testing I
I generator feedwater isolation valve.
I I
I I
I l
i N2 leakage on plug.
I I
i J
I I
I I
I I
I l VI.E.333 I Robinson 2 1 100% Power 1 AFW I Valve fails to open, Auxiliary feedwateel L I
I I (06/01) i I
I discharge to steam generator A.
Caused 1 1
I I
I I
I by heat due to back--leakage of down-1 I
J l
l I
I I stream check valve.
I J
l i
I I
I I
I N
I VI.E.199 8 Beaver Valleg il Startup I AFW l Steam supply trip value would not close.I FC i
I I (03/79)
I 100% Power I I
I i
I I
I I
I I
I O
I VI.E.32 i Ginna I Power I MFW I
"B" steam generator feedwater iL 1
O Page 6
l l
~ ~ _ _.
J l
I (07/73)
I I
I cantest valvo plug esperated 1
I I
I I
l I fraz etEm. ccu3Gd 19e3 ef f3Sdwat5r I
i I
I I
I I and water hammer I
I I
I I
I I
I I
I VII.B.12 I Surry 1 and 2 i Refueling I CS I Cracked rockshafts and cracked body in iL I
I I (04/82)
I and Power I i check valves (CS and RS).
I I
I I
I I
I I
E I VII.B.123 I Rancho Seco I Starting 1 SI I Various problems on check valves caused iL 1
I I (04/82) i I
I backflow from BWST to borated water I
I i
l i
i I additive tanks.
I I
g i
8 I
i i
I I
I VII.B.131 8 Cinna i 100% Power I CS I Core sprag pump check valve failed l FC 1
I I (09/82) l I
I to close seat and disc machined.
I I
g i
I I
I I
I I
i VII.B.134 I Cinna i 100% Power 1 CS l Spray pump check valve pin / swing are i FC I
I I (C6/82) l I
I worn.
I I
j J
I I
I I
I I
I B VII.B.142 8 Cinna I 100% Power I CS I Core spray pump discharge check valve 1
1 1
I (07/82) i I
I faulty.
Pin / swing arm worn.
Air-I I
l I
I I
I I operated valve.
I I
I I
I I
I I
I I VII.D.34 i Cook 1 1 100% Power i CI I containment isolation valves VCR-10 i FC i
J i
I (9-12/79) 8 I
I and 20 would not close.
I I
l 8
I I
I I
I I
I I
I I December 1979--valve VCR-10 failed to I FC 1
J I
I I
I I close' I
I l
I I
I I
I I
I I Cook 2 1 75% Power I CI I Containment Isolation check valve I WP I
J I
I (06/80) i I
I VCR-156 installed backward 1
I l
I I
I I
I I
I I Cook 1 1
I I November 1980. VCR-20 failed to close I FC I
M I
I (11/80)
I I
I I
I I
1
.'I I
I I
I I containment steam heat suppig CI check i L I
I VII.D.47 I Yankee Rowe i 100% Power i CI O
I I (04/82)
I I
e I valve flanges leaked. Relaxation of I
I I
I I
I I flange bolts.
I
/
I I
I I
I I
I I
O I
I (03/80) 8 I
I I ~
I I VII.E.78 I Arkansas One 1 1 100% Power I i Check valve failed to open - corrosion i FO I
E I
I I
I i
1 y
I VII.E.28 I Davis Besse 1 I Refueling I CI I Sir CI check valves leaked. Discs and iL I
l I (06/78) i I
I seats cleaned and lapped.
I I
I I
I I
I I
I I VII.E.20 t Davis Besse 1 1 68% Power i I Hydrogen dilution blower check valves I FO I
I I (04/78)
I I
I stuck (suction check valves)
I I
I I
I I
I I
I I VII.B.28 I Rancho Seco I Hot I DHR I Two stop check valves stuck closed -
1F I
I I (19743 I Functional i I Design error I
I I
I 8 Testing i
I I
I I
I I
I I
I I
I VII.B.43 I Robinson 2 1 100% Power I CCW l Debris check valve caused CCWP trip.
IL I
I I (01/753 1
1 I cracked pump seal.
I I
I I
I I
I I
I l VIII.B.82 I Point Beach 2 i Refueling I CI I CI Check Valve in CCW Supp1g to "A"
IL I
I I (03/763 I
I I RCP Leaked.
Dirt Accumulation on Seat. I i
I I
I I
I I
I 8 VIII.B.118 i Palisades i 100% Power i SW l SW Pump Discharge CKV stuck.
I FC 1
i i
I (02/773 8
I I Lubrication.
I I
O I
I I
i i
l i
1 0
Pale 7
e i VIII.B.12a 1 Cook 1 I R3fu211ng I SW I SW CKV Locked. D;pesits frge Lcke IL E
I (02/77) l I
I Watst.
I I
I I
I i
1 I
I I VIII.B.127 i Calvert Cliffs 8 100% Power i SW l SW Pump Discharge Check Valve Stuck i FO I
I I (04/77)
I I
I I
I I
I I
I I
I I
I VIII.B.151 1 Calvert Clif's ! Power i SW I SW Pump Discharge CKV Stuck Corroded i FO I
I I
I I
I Flapper Hinge Pin.
I I
1 I
i i
l I
I i VIII.B.182 1 Davis Desse 1 I Power l SW l SW System CKV Rusted and Inoperable 1 FC 3
g i
E I
I I
I I
I I
I I
I i
1 VIII.B.183 I Palisades I 9C% Fswer i SW l SW Pump Discharge CKV Stuck Shut.
I FO 1
0 8
I (04/77)
I I
I Corroded Hinge.
I I
I I
I I
I I
I I VIII.B.228 i Cook 2 1 100% Power i SW I SW Supp1g CKV to Diesel Failed to Seat. I FC I
I I (01/%9) i I
I Disk Wear.
I I
i 8
i i
i I
I I VIII.B.233 Calvert
! 94% Power i SW I Saltwater Pump Discharge CKV Failed to I FO I
~
I I Cliffs 1 I
I I Op en.
Foreign Debris.
I I
I I (02/79) i I
I I
I I
I I
I I
I I
I VIII.B.290 i Cook 2 i Refueling i SW I NESW Check Valves Leaked. Sand Deposits IL 1
1 I (12/79)
I I
I Pitted Seats I
I I
I I
I I
I s
I VIII.B.308 Connecticut 1 Refueling I CI I Containment Penetration Check Valve
!L l
i I Yankee l
I I Seats Leaked. Weld Problems.
I I
I I (05/B0)
I I
I I
i i
I I
I I
I i VIII.B.380 i Beaver Valleg 18 Starting i SW I River Water Check Valves Deteriorated i FD I
I I (04/80) 1 I
I I
I O
I I
I I
I i
1 I VIII.B.411 I Cook i Refueling i SW I Leaking NESW Check Valves. To Be IL 1
I I (07/81)
I I
I Replaced With Diaphragm Type.
I I
O i
8 I
i
+
3 I
I I VIII.B.423 I Surry 2 1 100% Power I SW I Charging Pump SW Check Valve Failed to I FC /
I I
I (08/81) l I
I Seat.
I I
O I
I I
E I
I I
8Fb i
I VIII.B.461 1 Surry 2 8 100% Power i SW I Charging Pump SW Check Valve Worn.
I I (08/81) i I
I Stuck Open.
I i
y i
I I
I I
I I
I VIII.B.493 1 Davis Besse 1 i Cold i SW I SW Swing Check Valve Disc Seized Fu119 i FC I
I I (06/823)
I Shutdown i
I Open. Corrosion Deposits. Disc Stud 1
I I
I I
I I Penetrated.
I I
I I
I t
i I
I I Containment spray and RHR Chec k Valve IL l
I VIII.B.569 I Cook 1 I Refueling i CS 8
I (08/93)
I I
I Leaked Disc Lapped.
I I
1 I
I I
I pump SW discharge)
I 1
1 I
I I
I I
I l
i VIII.B.574 5 Surry 1 and 2 1 100% Power i SW I Check Valve seeding failure.
(Charging iL i
1 I (6-10/83 8 and i
I pump SW dischare)
E I
l I
I I Shutdown I
I I
I I
I I
I I
i 1 VIII.B.609 I Cegstal River 31 94% Power i SW I Seawater Pump Discharge Check valve I FC
,1 I
I (04/84) l I
I Failed. Corrosion I
't b
I I
i I
I I
i I IX.F.101 i Rancho Seco I 95% Power I SI I Nake-up Tank Pressure Above N2 Header I-1 I
I (10/82) 1 I
I Pressure Cause Value/ Leakage. (Not a I
i O
I I
I I
I Value Problem)
I 0
Page 8
1
o._.....
l I
I I
I I
l 1 IX.F.120 i NcGuire I 50% Power 1-1 VUC3T (Iow tats 11 gr creatic chcc v51va I L I
I I (8 and 9/82)
I I
I leaked. loss of 11guid at the flow I
I I
I I
I I element.
I I
I I
I I
I I
I I IX.F.44 1 Maine Yankee 1 90% Power IContain-l Check Valve on Containment APD pump iL 8
I I (04/75)
I iment APDI failed. Wern seats.
I I
I I
I I
i 1
1 8 IX.F.689 I Qinna i 100% Power Icontain-1 Containment Oas R/A monitor Check Valve !L 3
8 I (6 and 9/82)
I Iment Cast Leaked. FM on the Seat.
I I
g i
I i
1R/A I
I I
I I
I IMonitor I I
I I
I i
I i
i I
O I IX.F.78 i Cook 2 i Refueling i SW I il NS*J Check Valves Leaked. Lakewater iL I
I I (02/77)
I I
Deposits on Seating Surfaces.
I I
I I
I I
I 2) Containment Purge Isolation Valves I
l I
I I
I I
Leaked. Loose Actuators.
I I
LEGEND:
-i FC = Fati to Close F0 = Fall to Open FD = Failed on Demand F = Failed L = Leakage j
WP = Wrong Position J
v C
O e
/
A d
i i
J l
I e
e O
Page 9
~
_ ~
\\
~
ATTACHMENT C
\\
SAN ONOFRE EVENT OF NOVEMBER 1985 (Reference 2) e.
N 624.
Reactor Trip on foes of vital Bus - toss of 6
W System Integrity - Water Hammer - Main W g'
Check Valves Failed
,N N
in Gene 1 - Jan 86 I
San onofre 1 - v 85 - 60% power l
g on 6 Jan, the NRC issued IE Information Notice 86-01, "Tailure of Main Feedwater Check Valves Causes The reactor trip also caused level shrink in the SCs, causing SC level to drop below the actuation Loss of Feedwater System Integrity And Water-Mammer level for the aux W ( AW) pumps.
The electric-Damage," to inform recipients of a recent event ally driven AW pump received an actuation signal, caused by 5 main W (MW) check valve failures at a WR plant.
The failures resulted in a loss of Mg but no longer had electric power available.
The steau-driven AW pump, after a 3 min auto warmup system integrity and significant water hammer dam.
period, began to deliver relatively cold W to the age.
SCa at a point in each of the M W IInes between the NW regulating valve discharge-check valve and the on 21 Now, San Onofra 1 was operating at 60% power SC.
The AW to the SC W lines initially flowed when an aux transformer failed, resulting in a loss backward through the failed check valves and of power to a vital bus and to the bus feeding the forward.through long horis runs of W pipe in east (electric) MW pump.
The west (electric) MW pump remained energized from the unit main primary containment.
Although the operators were unaware that the check valves had failed, they then generator because of an abnormal electrical lineup.
When the east MW pump tripped, its discharge-check closed all MW regulating valves and their associated isolation valves per procedures.
(The valve (WS-438) failed to seat properly.
As a result of the discharge-check valve failure, the effect of the MW valve closure had yet to be determined.) Contact between steam in the W lines west MW pump supplied W backward through the and the cool AN In the horis pipe resulted in a discharge-check valve and uverpressurized the east
~
The water hammer caused damage to W heater-condensate train.
Several tubes appar.
the W line pipe supports and stretched the bonnet ently ruptured in the east W train fifth stage bolts on the
'B' W regulating va %e bypass line (LP) W heater as a result of the over.
pressurlastion, causing the shell side of the check velve (INS-378), causing the anetal valve heater to rupture.
Also, several main turbine gasket to extrude. The flapper on the 4 In. check rupture dirce failed.
Following these events, the valve was later found to have been damaged by the water hammer impact.
De extrusion of the valve operators tripped the teactor and turbine per procedure because of the loss of power to a vital gasket resulted in a substantial steam-water leak bus.
That also caused the west MW pump to trip.
from SC 'B' to the N messanine area and the sta Both 12 In. MW pump discharge-check valves (WS.
which was not isolable for some time because of the proximity of the associated isolation valve to the 438 and -439) were later found cocked
- open, supported by their disc antirotation lugs that had leak. SC 'B' boiled dry because all ATV 'B' flow was carried out through the leak.
Plant personnel rotated under the check valve hinge arm.
finally able to close valves TWS-342 and WS-were 376 to isolate the leak and continue the plant When the west HW pump tripped, all 3 SC MW regu.
cooldova - 6 he af ter the event started.
lating valve discharge-check valves (FWS-345, -346 and -398) also failed to seat. Two of those 10 in.
De NRC sent a 3-member incident investigation team check valves were later found to have their flap-( 117) to the San Onofre ! site shortly after the pers loose in the bottom of the valve body with incident. The licensee agreed to hold in abeyance their nuts missing.
The third check valve was later found to have failed in the sase mode as WS, any work in progress or planned (as allowed by 438 and -439.
The check valve failures in the MrW plant safety considerations) until the licensee and the FRC had an opportunity to evaluate the event.
system resulted in leak paths from the SCs backward through the MFW regulating valves and the east My The IIcensee also agreed to keep linit I shut down until concurrence was received from the NRC to re-pump to the ruptured east train W heater.
- Also, the west M W train might have been pressurised from turn to power. The IIT coepleted a preliminary in-the SCs.
The net effect was that the inventory in vestigation of the event and expec ted to issue a report la Jan 86.
(wwg) all 3 SCs began to blow steam and hot water back through the east MW train.
ENCLOSURE No. 3 TO SBN-1207 FREQUENCY OF UNFILLED RHR PIPING t
I e
1 1
s i
O
~
FREQUENCY OF UNFILLED RHR PIPING The enclosed writeup provides a conservative bounding analysis of the.
f requency of RHR piping f ailure due to an undetected presence of air in the syst ect during an interf acing system LOCA.
This analysis f ully justifies the assessments made in RMEPS that such events, even when conservatively analyzed, make insignificant contributions to risk in relation to those already in the model.
l i
-l
FREQUENCY OF UNFILLED RHR PIPING In the interfacing systems LOCA analysis given in Reference 1, a piping fragility analysis was performed in which it was assumed that the RHR system piping was initially water solid. Under these conditions, analyses were performed that showed that the peak stresses in RHR piping and components are those that result from RHR system pressurization to a value that is bounded by the initial RCS pressure of 2,250 psia. An independent verification of this peak pressure, which includes dynamic and static contributions, was performed in the study in Reference 2.
The-frequency of failure of RHR piping due to pressurization to 2,250 psia was estimated to be 6 x 10-3 per demand, based on a simple fragility model described in Reference 1.
The above analysis did not explicitly include a postulated failure mode in which the RHR piping, or major sections of it, is initially assumed to be unfilled with water.
In these cases, there may be additional loads on the piping and supports associated with water hammer effects or other effects associated with the acceleration of a slug of water through the system.
It was judged in the Reference 1 analysis that such failure modes, if explicitly considered, would make an insignificant contribution to risk in comparison to the scenarios that were modeled. There were two principal reasons for this judgment. One is that the frequency of having an unfilled RHR system, or significant segments of the piping unfilled, is believed to be extremely small and certainly much frequencyalreadyassignedtopipingfailure,6x10gessthanthe per demand when pressurized to 2,250 psia. The second reason is that even if such an unfilled condition were postulated, it is very doubtful that major ruptures of the piping would result. As far as small leaks are concerned, the addition of new, lower frequency scenarios with small leaks would also have a negligible risk impact because the Reference 1 analysis already assumes that even when the piping remains intact, small leaks through the RHR pump seals are almost certain to develop. The purpose of this note is to revisit this question by more quantitatively addressing the frequency of unfilled RHR system piping.
Two independent approaches were taken to obtain a bound on the frequency of unfilled RHR piping. The first was to examine experience data with PWR RHR systems to identify problems associated with air ingress, improper venting and filling procedures, inappropriate operating of valves, and other events in which there was evidence that the RHR system was not properly filled and vented. The results of this experience review provided a basis for estimating the frequency of unfilled piping.
The second approach was to estimate this frequency using on a model that considers the frequency of maintenance operations on RHR components and the potential for human errors in restoring the system to its normal, solid status.
1 A review of about 280 reactor years of PW RHR experience was conducted, based on Reference 3.
From this experience, a total of 12 events at 6 different plants was identified in which problems with air in the RHR 1
1420P091686
system was experienced. Four of the events happened at one plant. These events are summarized in Table 1 and event summaries provided in the attachment.
In most of these events, the presence of air in the system was discovered because of the indications of RHR or safety injection pump cavitation or air binding, as the air void was circulated into the pump. Of the 12 events,10 happened during refueling or cold shutdown,1 at 100%
power, and 1 at hot standby. These data cover about 276 reactor years of experience on a calendar basis. Assuming our average value of refueling intervals of 1.5 years, this covers about 184 refueling outages.
Therefore, the average frequency of RHR air ingress problems has been about (10/184) =.064 events per refueling outage. Over the period of this same data base, the amount of reactor power operation, assuming a 60% average capacity factor, has been (0.6) x (276) = 166 reactor years or 166 x 8,766 = 1.5 x 106 hou rs. The average frequency of RHR problems, while at power in hot standby, is then 1.5 x 10-D per hour.
An interfacing system LOCA scenario requires that the RCS be initially pressurized. For an air ingress event that occurred during refueling to continue to exist during power operation, it is necessary to postulate failure to perform preoperational testing of the RHR system, which would be carried out prior to plant operation.
It is also necessary to postulate failure to perform the monthly surveillance procedures to test the RHR pumps in the miniflow mode and the monthly venting tests to check for the pressure of air, either of which will detect the presence of significant amounts of air. Furthermore, in the event that the initiating event occurs more than 2 months after the previous refueling outage, it is necessary to postulate failure of multiple monthly surveillance tests.
It is conservatively assumed that the frequency of failure to detect and correct RHR air ingress during refueling is 1 x 10-3 per demand.
It is further and conservatively assumed that the frequency of failure to detect this problem during the first monthly surveillance test is 1 x 10-2 and again conservatively assumed that if it is not picked up in the first test, it is not picked up in any subsequent tests, depending on when the initiating event occurs. Over the entire refueling cycle of 12 months, the conditional frequency that the initiating event occurs before the first surveillance test is 1/12.
Hence, the frequency with which the RHR system is not properly filled at the time of the initiating event due to an undetected air ingress that occurred during the last refueling is.054 air ingress events per refueling outage x.001 failure to detect air in preoperational testing x [(1/12) + (11/12) x.01 failures to detect air in surveillance tests] = 6 x 10-6, To calculate the frequency of an initially unfilled RHR due to causes other than refueling, two approaches can be used. One is to use the experience alone to calculate the frequency with which an air ingress event during plant operation is unfortuitously present at the time of the initiating event. The above data identified two events in 1.5 x106 hours of reactor operation. One event lasted 45 minutes and the other an unspecified amount of time.
If we conservatively assume that each event lasted 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the fraction of time of RHR air presence l
2 1420P091686
during power operation was (2 x 24)/(1.5 x 10-6) = 3.2 x 10-5 This is the frequency with which we would estimate that the RHR system is unfilled at the time of the initiating event due to events that occurred previously during power operation.
An independent conservative assessment of the operation contribution can be obtained as follows. The RHR system includes the following components which have a significant frequency of scheduled and unscheduled maintenance.
Maintenance Frequency
- Number Component per Component (events / hour) 2 Motor-Driven Pumps 8.4 x 10-b 2
Heat Exchangers 2.7 x 10-5 16 Motor-Operated Valves 2.7 x 10-5 Total for System 6.5 x 10-4
- Taken from PLG-0300 (SSPSA), Table 6.4-1.
The above frequency, which is based on the PLG generic data base that was used in ttle SSPSA, is considered highly conservative because 6.5 x 10-5 events per hour translates into one maintenance event every 2 months on the system. Only for a smalI fraction of the maintenance events would the RHR system be required to be drained or partially drained. Even when this is required, maintenance procedures require refilling, venting, and preoperational testing of the system to verify no air in the system. Assuming tne same error rates as before (namely,
.001 for failure to perform the post-maintenance test properly and
.01 for failure to detect air in surveillance tests if the maintenance event was performed more than 1 month prior to the initiating event) the likelihood of an undetected air bound RHR system due to maintenance events other than at refueling is (6.5 x 10-4/hr)(720 hr)(.001)[(1/12) + (11/12).01) = 4.3 x 10-6 This estimate is in good agreement with that obtained more directly from plant experience data. Using the latter result, the total frequency of an unfilled or partially unfilled RHR system at the time of the initiating event due to both refueling and nonrefueling undetected air ingress events is 5 x 10-6 + 4.3 x 10-6 = 5 x 10-b per RHR pressurization event. This value is only about 1% of the frequency already assigned to the failure of the RHR piping due to overpressurization when the system is filled. Because of the numerous conservative assumptions in the above analysis and because an unfilled RHR system does not automatically mean that the system would lose its integrity if pressurized, the original judgment that unfilled piping is 3
1420P091686
not important is fully justified. Stated another way, if we conservatively add the above bounding estimate of RHR failure frequency l
due to unfilled piping to our current fragility model, there would be an insignificant contribution to our results.
REFERENCES 1.
Pickard, Lowe and Garrick, Inc., Westinghouse Electric Corporation, and Fauske and Associates, Inc., "Seabrook Station Risk Management and Emergency Plan Study," prepared for the New Hampshire Yankee Division of Public Service Company of New Hampshire, PLG-0432, December 1985.
2.
Frank, M. V., " Shock Wave Effec; Following... RHR System, PLG J.
memorandum to K. N. Fleming, August 22, 1986.
j 3.
Nuclear Power Experience, Vol. PWR-2, S. M. Stoller, Albuquerque, New Mexico, 1979-1985.
1 1
1 1
1 I
1 4
b 4
4 1420P091686
TABLE 1.
EVENTS INVOLVING UNFILLED OR UNVENTED RHR SYSTEMS IN U.S. PWR PLANTS FROM JANUARY 1,.1971, TO JULY 1986 (TAKEN FROM REFERENCE 3)
Sheet 1 of 3 Event Plant Status Date NPE Reference 1.
Safety Injection Pump Cavitation.
Turkey Point 3 Cold Shutdown May 1974 PWR VII.A.46 Air ingress believed to be caused (ECCS) by prior modification of RHR pump
' suction stop valve from sump, which required draining of line. Valve cycling test performed I week later could have allowed trapped air to enter the safety injection pumps via the RHR system piping.
2.
a.
During draining of the refueling McGuire 1 Refueling April 5, 1983 PWR XV.421 cavity, water level in the RPV PWR XVI.B.70 dropped to level of the hot legs--led to cavitation of HHR pumps.
(RHR system was not solid.)
b.
RHR Pump Cavitation. Voids McGuire 1 Refueling April 21, 1983 PWR XV.421 existed in the piping of the nonoperating RHR train due to physical layout of the RHR system and vessel water level.
(Top of RHR heat exchanger tubes was ~ Elevation 758' and RPV level was
~ Elevation 740'.)
3.
Air in RHR system was discovered Ginna Cold Shutdown May 1972 PWR XVI.C.2
'when plant shut down and depressur-ized and on shutdown cooling.
1421P091686
TABLE 1 (continued)
Sheet 2 of 3 Event Plant Status Date NPE Reference 4
4.
RHR system was operating at low RCS McGuire 1 Cold Shutdown March 1982 PWR XVI.C.1731 water level (4 to 6 inches below that specified for RHR system operation). This was allowing air
'into the RHR system suction.
5.. A containment spray system vent.
McGuire 1 100% Power June 1984 PWR XVI.C.2425 valve was not closed. During a valve stroking test, 35 gallons of water from the RHR system were inadvertently drained.
6.
Simultaneous venting and draining Cook 1 Hot' Standby April 1985 PWR XVI.C.2673 of the RHR system and RHR heat exchanger created a flow path between containment and outside atmosphere.
7.
RHR pumps were air bound.
Beaver Valley 1 Cold Shutdown September 1978 PWR VIII.B.200 8.
Air binding of the LPSI pumps.
Calvert Cliffs 2 Cold Shutdown October 1978 PWR VIII.B.213 resulted when air being used to transfer purification system resin leaked into the shutdown cooling system.
t 9.
Air binding of RHR pumps resulted Beaver Valley 1 Refueling January 1980 PWR VIII.B.294 from the operation of the vessel vent eductor system.
- 10. RHR pumps 'were air bound.
Beaver Va'lley 1-Cold Shutdown April 1980 PWR VIII.B.296 PWR VIII.B.304 i
.1421P091686
l TABLE 1 (continued)
Sheet 3 of 3 Event Plant Status Date NPE Reference
- 11. RHR pump was cavitated..RHR pump Trojan Cold Shutdown June 1981 PWR VIII.B.404 was operating at low RCS level.
Level indication was incorrect due to pressurizer vent valve being closed during the draining of RCS--
personnel error for not opening the vent valve.
- 12. Partial draining of RHR system Trojan Cold Shutdown April 1978 PWR XVI.C.605 occurred because of open drain valve.
1
%J 1421P091686
_,mww w
_w w,
w-
_e
_a__
i ATTACHMENT 4
1
-~ -
Vs1. WR-2 VII. Szfsty Systems 44 BORIC ACID TRANSFER PUMP SHAM FA11.t'RES A. Emerg. Core Cool.
- p. 13 Robinson 2 - Aug & Dec 73, Mar 74' Two previous incidents of boric acid transfer pump shaft failures (in vicinity of shaft keyway) had been experienced (in Aug and Dec 73). Prior to this the "B" Boric Acid Transfer Pump was in' service, while at 1007, power, recirculating "B" Boric Acid Storage Tank (BAT) with Boron Injection Tank (BIT).
The "A" Pump was in service recirculating the contents of "A" BAT.
O The "B" pump breaker then tripped due to thermal overload. n e breaker was reset and the pump re*
46.
AIR IN SI PUMP started. Pump discharge pressure was checked and found abnormally low. D e BIT closed with no Turkey Pt. 3 - May 74 appreciable increase in pump discharge pressure.
The pump was secured and declared inoperable.
n e 3A SI system pump was started in order to add water to the SI accumulators. De pump started but ne pump motor was checked electrically and found to indicated abnormally low running current and low be sound, therefore, it was apparent the inteller discharge pressure. The pump was imediately had separated from shaft or the shaft had broken.
stopped and valve lineup was verified to be correct.
Investigation proved the latter to be true. ne With maintenance personnel present, the pump was pump was replaced with a spare from stock. The pump restarted with similar results. The ensuing which failed was to be rebuilt with certified parts investigation revealed the pump casing contained or replaced as a unit. D e "A" pump had a modified air. It was vented via 2 installed casing vents pump shaf t as a result of previous pump failures.
which are normally blind flanged. A small amount ne "B" pump was to be replaced with a pump with of air was released from the vent at the suction a modified shaft as soon as it could be obtained end of the casing, and a larger volume of air was from the vendor.
(aob) released from the vent at the discharge end of the pump casing. After the pump was vented, it was then tested satisfactorily. The remaining Unit 3 6 45.
BORIC ACID DANSFER PUMP SMAM FA11,URE 4 SI pumps were. tested and performed satisfactorily.
(
Robinson 2 - Apr 74 h e SI and RHR system piping and pumps were clso vented to ensure they were free of trapped air, ney were operating at 150 MWe when a high tempera
- Appropriate work documentation was initiated for ture alarm was received on "A" Boric Acid Transfer the installation of additional vent lines and.
Pump, initiated by Heat Trace Recorder No. 1.
An valves in approved locations in the SI and RHR operator icraediately checked the pump and found it systems.
stopped, ne pump breaker was normal, therefore, it was determined the pump had tripped due to ther-ne exact cause of the air in-leakage could not be mal overload. D e pump was restarted. Current determined, however, subsequent investigation readings indicated abnormally low, therefore, a revealed the most likely cause was from maintenance broken shaft or impeller separation was suspected.
performed earlier on the SI system. About 5 wk The pump was secured.
earlier a modification was performed on the RHR pump suction stop valve from the containment post The pump was disassembled and a broken shaft in the accident recirc sump. The modification required vicinity of the keyway was determined to be the draining of a line. About,1 mo earlier a valve cause of the failure.
The pu:np was replaced with was cycled according to the SI periodic test which a spare pump from stock and was to be repaired or could have allowed the trapped air to enter the RHR replaced for future use.
system piping. Operating logs indicated the pump was run periodically during the month to add water Past failures (see VII.A.44) of this type pump led to the accumulators. This could have drawn the to modifications which were expected to increase trapped air in the RHR piping into the SI pump reliability. One anticipated improvement was by way of the comon suction piping.
(aqt) redesigning the shaft keyway, from a square to a round shape to reduce fatigue in that area.
How-ever they did experience a failure of a modified 47 PIPING VIBRATIONS DURINC Fil.I.INC OF SI TANK shaf t, therefore, they felt this was not the ultimate solution. Representatives of Chempump (Crane)
Ft. Calhoun - Apr 74 (power escalation testing) were contacted and were planning to continue their N 's"haf t failuns.
8 " ^. 58 s 67 fcr subsequent During a shutdown, and white fitting a SI tank (aqq) using a high pressure SI pump, the operator noticed a pressure drop on a pressure indicator. An increase was also noted in the Containment Sump level. About 2 days later the same thing happened again. On both occasions the operator discontinued filling the tank by closing a valve, at which time the pressure returned to normal. A drain valve was e
e==
e Aug 79
Vol. PWR-2 XVI. Operational Problems
- 5. Refueling
- p. 28 6
69.
H.P. Technician Seriously Injured, been closed during an attempt to reduce leakage Contaminated in 40 Ft Fall From Crane into the RCDT, isolating the RPV level gauge used To Reactor Cavity Floor to indicate that the level was approaching the ve*-
set fla,ge.
Procedures Jid not require visual mt.-
Cinna - Mar d3 - refueling itoring of cavity level.
The cavity was refilled and the ND system was vented and declared operable.
During the early morning hours of 31 Mar, an H.P.
Procedures were to be revised and personnel were to technician was installing a radiation monitor on be counseled.
See IV..e 21 for additional conse-the refueling bridge of the manipulator crane in quences of this event.
(sab) preparation for refueling operation, above the dry reactor cavity.
The manipulator crane was located at the south end of the reactor cavity against its 71.
Addition To XVI.B.66 ret FAs Damaged south travel stops above the fuel trans fer mechan-During Core Loading - Reasedial Action ism upender.
During the performance of this task, the technician fell, unbeknown to others, to the Turkey Pt. 4 - Apr 83 - refueling bottom of the reactor cavity floor about 40 ft below.
On 17 Apr during spiral loading of the Unit 4 core, they observed that FA X-04 had leaned across the He was. found by one of the fire watches (between empty core center and was resting on FA W-51.
0056 and 0702 hr), who gave information to one of Eight assemblies were removed for inspection, and 3 the security guards with subsequent notification to of those were rejected or held out for additional the control room.
The medical emergency team was tests or repairs.
activated to provide necessary help.
An ambulance and the plant medical physician were summoned.
The in Nov they reported procedural modifications to technician was placed on a stretcher and removed prevent fuel handling errors.
A TV camera (for from the reactor cavity for t rans po r ta t ion to the accessible core positions) and visual verification hospital via ambulance. The subsequent report from of FA position were to be required.
When. loading the hospital was that the technician had a collap-fuel into the core, no FA was to be unlatched until sed
- lung, broken ribs, ' broken
- leg, internal proper alignment with the guide pins was verified.
bleeding, and spinal injury.
His condition was Requirements were added for adequate lighting at serious.
all times during fuel movement. Any problem exper-ienced with the lighting was to be corrected prior The H.P. technician was found on the fuel transfer to any further fuel movement.
Malfunctions of mechanism where the dose rates were from 100 to 200 lighting equipment or other refueling equipment was arem/hr.
After he was removed from the area his to be reported to the plant supervisor-nuclear.
self reading pocket dosimeter indicated he had received 170 mrem.
Film badge and TLD readings Also, during refueling, a detailed turnover of the were to be processed to verify this dose but the refueling activities were to be given by the of f-170 mrem was considered consistent with the esti-going refueling shift.
The relieving shift would mated stay time and dose rate.
The contamination then verify that all equipment was functionally levels in the area were very high (greater than 106 operational before resuming refueling operations.
2 dpm/100 cm ).
Air sa:nples taken during the removal Additional checks were added to the various refuel-
)
indicated airborne contamination levels at about ing procedures to ensure that the periodic tests of 252 of MPC for CobO, the only isotope identi fiable the in-core, mast-mounted TV camera and the manipu-on the air sample.
Nose smears indicated minor lator crane were performed as required.
Operator contamination so. no significant internal contamina-requalification classes were to review fuel handl-tion was suspected.
Surface contamination levels ing procedures, provide classroom training on fuel were in the range of 200 cpm on the neck to 14,000 handling, and provide hands-on training in the cpm on the head to 20,000 cpm on the leg. Contam-spent fuel pit moving burnable poison assemblies ination was confined to small localized areas.
and thimble plugs.
(sjc)
Decontamination was performed in the hospital radiation emergency area.and the patient was then transported to the normal emergency wing of the N 72.
Spent Fuel Pool Cate Traveled Over hospital.
( ru f)
Irradiated FAs - Procedures inadequate I
N Millstone 2 - Oct 83 - refueling 10.
RCS Drained Down, RHR Pumps Cavitated &
Stopped - Level Not FComitored - Operator On 29 Oct, the spent fuel pool gate, a load of Valving Error about 4500 lb, traveled over irradiated FAs in the spent fuel pool.
The gate was being moved from the McGuire 1 - Apr 83 - refueling transfer canal to the cask laydown pat when discovery was made.
The load was then returned On 5 Apr, while the Unit I refueling cavity was over irradiated fuel to a sa fe Position by the being draired so that the RPV head could be placed transfer canal. Due to physical limitations on the in position, the RHR (ND) pumps began to cavitate.
travel of the crane, the load could not be Eventually both ND pumps were stopped.
Since the positioned over the designated safe load path.
To ND system was inoperable, RCS loop avg temp prevent recurrence, special rigging procedures were increased 280 The chem and vol control system was incorporated into the crane travel procedure to available to provide esol makeup water.
The reac-allow the gate to travel over designated sa fe load tor coolant drain tank (RCDT) isolation valve. had paths.
(szz) come=em Mar 84
A i
Vol. PWR-2 XV. Misc Systems l
I p.
113 iwo datterent analyses were performed on the fluid 420. ADutTION Tt) XV.192 re: SNUBBEst DEFICIENCIKS by the Florida Pow. r Corp. (FPC) umt minwt s:
one i
Ma t e r i a l s Tec h nology Dept. and the other by the snub-Millstone 2 - Mar 63
[
ber afr.
The Materials Technology report listed par-i ticle, of up to about 770 microns in size.
Their As of 22 Mar 83, all International Nuclear Safeguards I'
ennclusion was that the particles of Fe, Ni, and Cr Co.
(INC) mechanical snubbers installed on hangers cmdd. oe from SS.
The others appeared to be of that supported safety related piping systems had been extra vous origin (based upon review of materials replaced with mechanical snubbers manufactured by making up various snubber parts).
The report also Pacific Scientific Co.
(ofx,rve) y enneluwa that some of the elements could be of psint pigment origin.
421. RHR HANGER FAILURES - SNUBBERS thCKED UP -
/
WATER HAMMER DAMAGE In th-Power Piping'Co. analysis (done by Atmospheric Rnea rch Organization, Inc.), it was determined that
[
Nrticultte chemistries indicated welding metal and McGuire 1 - Mar, Apr & May 83 - shutdown flux, paint chips, and " dirt" as contaminants.
The-
)
r* port also showd an elevated phosphorus level found On 27 Mar, during unrelated maintenance, hanger IMCA-l in an unknown particle type.
It was stated that this ND-H260 was discovered to have pulled loose from its s hmastry was "not incompatible with welding rod wall anchors.
All RHR (ND) system mechanical snub-flax."
The one major dif f erence in the 2 reports was be's were subsequently full stroked on 27 Mar to en-r was performed sure sure their operability.
Two snubbers (IMCA-ND-t u nt the Materials Technology analysis nn fluid samples taken trom snobbera removed f rom the H177 & -H317) were found to be locked up.
A walkdown jl rr crue buitaing.
The Power Piping analysis was inspection of suspect piping was performed.
On 7 l
yrtormeo on fluid fras a snubber shipped to the Apr. hangers ND-H273 & -H287 were discovered. pulled 1
tield (Detroit Edison Co.) and returned to Power loose with broken concrete around their wall anchors.
Piping Co.
after about 3 mo without having been Continuation of the system walkdown on 8 Apr revealed jl 6 more hanger failures (ND-H1,
-H21,
-H274, -H282, installed.
-H306 & -H308).
The exact time that damage occurred I
FPG t onc luded that environmental conditions. might could not be determined or attributed to a particular have contributed to the contamination problem.
How-occurrence of water hammer (overloading the
-ver, it was their opinion that there was enough hangers / snubbers). These hangers had been inspected uv-silable J ta to say. that sufficient quantities of and approved in Jan 83.
At least 2 ND water haimsers cuqtawainants were present in the snubbers as supplied were thought to have occurred since that time.
l by Power hping Co. to render them inoperable.
An inoperable snubber in this failure mode (zero bleed-On 22 Jan, during shutdown, the ND system was aligned i
rate) could induce excessive amounts of stress in to provida normal decay heat cooling to the reactor l
piping that was supported by the snubber. Therefore, core.
It was theorized that voids formed by flashing IPC determined this issue to be reportable per 10 CFR of an isolated vol~of water between valves ND-1 &
j 11 because the - snabbers contained defects which con-ND-2 during leak testing might have contributed to a I
st a tuted a substant ial sa fety hazard.
water hammer at that time.
The problem of fluid contamination had not been noted On 5 Apr, during draining of the refueling cavity, until their then most recent refueling outage.
No water level in the stPV dropped to the hot' legs, and zero bated-ratt problems had been documented Juring the ND pumps lost suction (see XVI.B.170).
In order the 2 previous refueling outages. Due to the reser-to vent the loop, 6 pump start attempts were made votrs k ing a non pressurised type, vented to the while the system was not water solid.
. It was atm, tL :re was the possibility that conditions at the believed that this might have caused water 'hasusers in l
plaat bring storage, maintenance and/or operation the ND piping, damaging or contributing to existing 7
contributed to the contamination problem.
damage on the hangers.
J-Siace this problem was identified, they had developed Snubbers IMCA-ND-H177 & - H317 were replaced.' Anchor pr en in waich all fluid reservotrs, both old and bolts for hanger IMCA-ND H260 were retorqued and a completely flushed prior to filling with brace was added.
Repairs to hangers -H273 & -H287
- new,
.+
tresh t h t 4.
This cleaning process had revealed that (moving baseplates to undamaged concrete) were not nyw reser m rs, obtained from Power Paping Co. within completed within the 72 hr Tech Specs limit. The ND the proious yeir, contained varying amounts of what
'A' train was declared inoperable on 10 Apr.
On 12 appear 3 to be wlJ slag, small beads of metal and
- Apr, repairs were completed and
'A'-
train was fine metal tilings.
'they were also to institute a declared operable.
I pro gra:a to.: hec k the fluid of any snubbers removed tor testing and/or rebutiding, for evtdence of con-On 21 Apr, a maintenance worker on 733 ft elevation heard a loud ' crack' and witnessed the agitated move-tinued contamination.
ment of ND piping throw dust into the air.
Upon
% ring the lhi retualing outage, when this problem checking ND system supports, he tound t h.a t the anchor ns found, all snubbers inside cont.ainment were com-bolts on hanger ND-H21 had partially pulled loose
-j plately rebu ilt,
i.e.,
reservoirs flushsd, new seals,-
from the wall and -H273 bolts had also pulled out, new slaid, and required hardware added.
In addition, breaking concrete.
This occurrence had coincided 27 of the r.rsa ining 95 anabbers outside the contain-with the starting of ND pump
'IA'.
Farther. inspec-I ment. re rebuilt prior to %de 1
- 3. owe r operation.
tion revealed hanger -H287 had pulled loose from the l
From a review of piping analyses, tnese 27 snubbers wall as well.
Fur some time before the swapping of f
J c<hld have affected piping integrity due to zero ND trains, Pump
'lB' had been running with the RPV bleed-rate during tuermal movement.
fhe rebuilding level at about 710 ft elevation. The 2 ND trains were r-maining e snobbers was to be completed by 1 at the (ruo)
. uc t 32.
cmas=ca== * - e oct 13
,-,_..---n
Vol..PWR-2 i
XV. Misc Systems p.
!!4 isolated. on the discharge side of the 2 pumps.
been in service for b yr.
The f ailed ' snubber was PM ' program These conditions, combined with the physical layout replaced.
To prevent r ecu r renc.i. a was of the ND. system (top of the ND HX tubes-was about to be instituted following the next refueling to 758 ft elevation), could have contributed to voids rebuild hydraulic snubbers I throu gh O and 62 in the piping of the train not in operation.
The through 64 on a 5 yr cycle.
presence of such voids and the start of the
'A' pump was postulated as having caused the water On 27 May 82, during routine surveillance of hammer and subsequent hanger damage.
hydraulic snubbers (Bergen Paterson), it was deter-mined that snubbers No. 32 and 45 (Engineering Ac t ion taken included declaring ND 'A' train inoP*
Safeguards Pump Suction) contained insufficient oil etable, full-stroking of ND system mechanical, snub-and were therefore declared inoperable.
The snuo-bers, ins pe c tion of ND system hydraulic snubbers bers were found to have f ailed due to deteriorated and restoration of the Jamaged hangers.
- Further, o-rings.
These snubbers had been in service for i
S1 (NI) system piping was inspected and mechanical 6 yr.
The failed snubbers were replaced.
snubbers were full-stroked.
Insulation at the nearest elbows to the damaged hangers was removed On 29 Jul 82, during routine surveillat e of to allow closer inspection. Anchor bolts on -H287 hydraulic snubbers, it was determined that Bergen
& -H21 were retorqued, restoring the hangers.
Paterson snubber No. 31 (GC-8 Shutdown C wling LPSI Problems with -H273 could not be immediately Pump Suction) containad insufficient oil and was resolved due to broken concrete around the anchor therefore declared inoperable.
The snubber. was Dolts.
However, ND
'A' Train was improperly found to have deteriorated o-rings.
This snubber declared operable on 23 Apr, following an engineer-had been in service for 6 yr.
The failed snubbur ing evaluation regarding the effects of unrepaired was replaced.
hanger ND-H273.
The engineering evaluation deter-mined that the ND piping was sis t fic ien tly On 21 Sep 82, during routine surveillanc.e of Bergen restrained and supported without the immediate Patersun hydraulic snubbers, it was determined that restoration of -H273; however, a factor not consid-Snubber No. 39 (GC-d Shutdown Cooling LPSI Po mte ered in declaring the ND
'A' train operable was the Suction) contained insufficient oil ait des there-snubber support function of -H273.
Tech Specs fore declared inope rable. Failure was found to be required that ND system safety-related snubbers be due to fluiJ loss through scores in als cyt;nJer operable in any mode requiring the ND system to be wall of the snubber. The scores were attributed to operable.
The I:ydraulic snubber of -H273 was a minor vibrations over a 6 yr period.
The failed safety-related 13.750 lb snubber, and was covered snubber was replaced.
(rwc) by this. Tech Spec.
With the 'A' train improperly declared operable, Unit I changed f rois mode 5 to mode 4 on 28 Apr in isolation of Tech Specs.
O 423.
Hydraulic Snubber 1,eaked - Piston Restoration of -H273 was underway on 6 May when it Rod 0-Ring Failed was cascovered that the
'A' train had been improp-erly declared operable. Work was completed and the O
Pt. Beach 1 - Dec 82 - 77% power Sanger and snubber were restored on 7 Apr.
A memo-tandum containing guidelines to insure performance A duty shift supervisor noticed an oil spilt in the of all proper evaluations prior to removing any area below the
'A' SC main steam line snubbers.
4 hangers from service was distributed to appropriate The snubber oil reservoir w.is checked and found to personnel.
be empty.
Also, a pressure gauge atts_5ed to the 4
oil supply line was reading slightly lower than Procedures were to be revised to prevent further normal.
These findings led to the discovery t'ia t water hammers upon aligning to the NDS system dut*
snubber No.
lHS-1 was leaking oil past its shift ing shutdown'.
Loss of RHR system procedare was seal.
The snubber was manufactured by Anker-Hosth, modified to indicate that the system had to be Model No. 21.1250.007, and had a capacity of 50,000 vented between valves IND-1B & IND-2AC.
RHR oper-Ib.
ating procedure was changed to prescribe opening a cross-tie line on the discharge side of the ND The snubber was removed and it was found thit pumps priar to swapping trains.
This was intended although the oil reservoir was empty there was to fill any voids in the non-operating train with-still oil in the supply line. to the snubber.
The out causing water hammers.
(rva, me.s) leak was located above the cylinder ho.ly and the snubber was still full of oil when it was removed.
It was found that the le ak was caused by a damaged a 422.
Hydraulic Snubbers Inoperable Rings piston rod o-ring.
The snubber was repaired h1 Deteriorateu. Cylinder Scored installing a new shaft seal kit, which incluled tha piston rod o-ring.
After repairs, the snuSher was 1
o Palisades - May to Sep 82 - 0% to 100% power functionally tested and returned to service within 2 days.
As a result of this event, the allowed On 19 May 82, durind routine surveillance of interval until the next-required visual snubber was reduced from 18 to 12 no in sccor-hydraulic snubbers (Bergen Paterson), it was deter-inspection
.nined that snubber No. 17 (main steam line upstream dance with Tech Specs.
(rwu) of ats dump valve CV-0780) contained insufficient oil and was therefore declared inuperable.
The snubber failure was found to be due to deteriorated o-rings.
The normal service life for the o-rin>rs was considered to be 5-10 yr.
This snuboer h.4J c
g
,e a-Snv R3
C.
MMCEI. LANE 0t'S (OPERATIONAL PROBLEMS)
Vol. WTw2 XVI. Op;r. Prob 1ces C.
Misc.
1.
COWAINVENT INTECRITY VIOLATION
- p. 1 Cinna - Jan 70 (1) During the period of steam generator inspec-During a cold shutdown to replace 6 failed reactor tion, the minimum level in the primary coolant
/
coolant system RTD's, both personnel airlock doors pipe will be 10 in, above the centerline to assure were opened to permit flow of maintenance tools nf ficient NPSH to the RHR pumps.
and equipment. During partial draining of the (2) A hot leg tetrperature indicator was added so reactor coolant system to replace the RTD's, it was noticed that in-core temperatures were-40*F that a more comptshensive monitoring of loop temp-above the Tech Spec limit for lack of containment erature could be attained.
integrity. One door was shut icr:ediately and the i
door interlock system was returned to service within (3) A temporary recorder to monitor for any flow i
I hour.
fluctuations was connected to the RHR system.
To prevent recurrence, procedures were changed that Tech Spec violations that were probably violated
" *E#
Y "** "
$8 prescribe running the RHR system during draining; a
e ee shu h n con W u M M F, aM replacement of one RTD at a time; ma,s;taining con-in the coolant tinuns surveillance of in-core T/C and setting in the hot leg nozzle probably exceed-ed the heatup and cooldown rates of 60 and 50'F/hr a max limit of 130 F on reactor coolant system above which cooling must be initiated when containment respectively. (ft) integrity is not maintained.
(fr, fz) 3.
PRIMARY C00MW DILUTION TECH SPEC VIOMTION 2.
AIR IN RHR SYSTEM - TECH SPEC VIOLATIONS Ginna - July 70 Ginna - May 72 During a shutdown to repair a steam leak in a The plant was in cold shutdown with both manways to high pressure gage line, other maintenance work the "A" steam generator open and a strong back on was performed which required stopping both coolant pumps. Work on the turbine stop valves used steam, one of the "B" generator manways off. The loop plugs we g
e an e puatwe. To maintak W had not been installed in the "A" generator. In or, der to enter the generators, the water level had been c ntr 1 r ds at their position and the power con-lowered to 4 in. above the centerline of the reactor stant, the boric acid concentration was reduced primary coolant nozzles.
while the pumps were off. Tech Specs require one pump to be running to insure proper mixing. This Operators noted an increase in loop 4 T and that the Prompted considerable procedural training and or-RHR system flow and discharge pressure were both zero.
E*" *** "" *
""E***
- }
j The RHR pump was running smoothly. A valve to the 4
DECONTAMINATION OF EWIRE PRIMARY SYSTEM refueling water storage tara was opened, flow and pressure were established, and the water level was raised 4 in.
Shipping port - Mar 64 The RHR pumps were stopped and personnel entered the During Core 1 operations, radiation levels from "A" steam generator to verify that no material could main coolant system components increased because have fallen into the "A" hot leg piping which is the of conosion product activation within the core suction for the RHR system.
and subsequent redistribution of the activated crud throughout the system. The Core 1 to Core 2 At one time a temperature of ^-212'F was indicated.
m dification program, which consisted of extensive The level was raised another 14 in and RHR flow and alteration efforts on piping and components which pressure were established.
had been subjected to contaminated coolant for 67 mos of operation, required construction personnel It was believed that because of the configuration of to be in close proximity to the sources of radia-the RHR system suction piping, there are 2 potential tion. Radiation level estimates indicated that locations where the system could become air bound exposures to personnel would adversely affect the if the RHR system is operated at levels which could modification schedule, allow air into "A" loop hot leg and through vortex-ing into the suction of the RHR system. This is what In addition, reactor plant decontamination would could have occurred to cause the interruption of the reduce general area radiation levels during the RHR system. It is thought that some air was probably testing of the plant following modification, during removed when water from the refueling water storage future power operations, and during any plant main-vas injected into the "A" loop hot leg through the tenance that might be required during these periods RPR suction line and that with the loop water level of operation and testing.
raised, the system functioned properly and swept the remainir.g air from the system. Subsequent vent.
Two major objectives were:
ing at both locations where air could be trapped in-dicated no air to be present. Since the possibility
- To determine the effect of decontamination on exists that the air buildup could have taken place subsequent plant operation and on radiation build-over a long period a scheduled venting of the system up during Core 2.
was established while at those conditions. Other a
preventive measures instituted were:
- To determine the effectiveness of decontamination on a nuclear power plant which has been in ser-vice > 6 years and had accumulated > 2 5.000 EFPH.
Vol. PWR-2 XVI. Ope r. Problems source of make-up water) opened briefly. An attempt C. Misc.
was made to verify level at or above the RTD r.ani-
- p. 448 fold which would have insured proper level for pump operation. Water was reported present when the RTD manifold vent was opened, and the level in a tygon
" Operations Standing Order" was issued for both tube vented to the atmosphere agreed with the level units which prevented either unit f rom being operated gauge. Although cause for the pump's behavior could in conditions where the Specs were applicable until not be determined, it was returned to service as procedures could be revised to reflect the require-soon as possible. Pump parameters soon began to ments of this amendment.
(pfh) oscillate, and flow was throttled to 1,500 spm.
1.a t e r, flow was throttled to just above 500 spm af ter parameters had begun to oscillate again.
N 1730. SUPPLY AIR REGULATOR FOR A/C After flow was throttled to 500 gpe, level was THERN0 STAT F06D MISADJUSTED observed (using tygon tubing vented to the RTD man-ifold vent) below that specified for ND operation.
N Calvert Ciiffs i _ Mar 82 - iOO: power Tse operator that checked level earlier had been misled by a small amount of water present in the The No. 12 Control Room air conditioner (A/C) failed vent piping. The low water level was allowing air to start. The supply air pressure for the A/C into the ND suction, causing partial air-binding unit's thermostat was found set at 12 psig higher of the pump, with symptoms similar to cavitation.
g than the proper 20 psig. This caused a continuous As level was later increased. ND flow was increased, j
false high pressure output signal to the A/C com-without signs of air-binding, until level was 175 pressor control pressure switch. The pressure in, and flow % 3,000 gpm.
switch would not have actuated to start the compres-sor as normally required by a rise in Control Room The only instrument normally indicating water level temperature. The No. 11 Control Room A/C unit, con-while draining or filling the NC system was the trolled by redundant instruments and thermostat, control board level gauge. The transmitter for this wss available^for cperation.
gauge was referenced to atmosphere because its design function was to indicate water level during Rasdjustment of the thermostat supply air regulator refueling. Pressures other than 0 psig were pos-(Norgrea) restored control operation for No.12 A/C sible during draining and filling because the t
allowing its compressor to be started by the demand system was not directly vented to atmosphere.
signal. It had not been determined how the supply During draining, NC was vented to the PRT where an air regulator output was raised. The regulator N2 overpressure could exist per procedure, " Drain-responded properly during corrective action for ing the Reactor. coolant System". The PRT was mon-this event. in the event that the regulator was itored using a 0-100 psig pressure transmitter intdvertently adjusted, a facility change to add and the computer video display (low pressures could temper-proof, locked metal guards to enclose pneumatic not be accurately read on this 0-100 psig control control panels and thermostats in the Control Roon board gauge) to insure no pressure deviation occur,
A/C System was to be completed.
(pfn) red that would significantly affect indicated level. NC pressure could differ significantly from PRT pressure if the drain or fill rate was high N im. = P= m not-D - tmt mNsmTm bmuse u. n mt m tag. 2 s- - '
MISAPPLIED the PORV discharge line inside the PRT. When ND
'h'***2'd
- 5
- P" '"' " ' '" ' "d t -
N cated 0 psig with the drain secured. 'A pressure of McGuire I - Mar 82 - cold shutdown 2.89 psig would cause a 170 in. level to indicate On 2 Mar, while draining the RCS (NC) System for SG 250 in, on the level gauge. Calibration of the pres-inspection, investigation of an RHR (ND) pump low sure transmitter on 3 Mar revealed it was within discharge pressure alarm resulted in ND pump IA being tolerance.
stopped due to signs of cavitation. During the 50 sin it was secured, no means existed for removing In the 50 min ND was of f, incore thermocouples core residual heat because the redundant pump IB was indicated core coolant temperature rose from %
0 out of service for manitenance. Upon restart, normal 105 to 130 F.
A conservative linear extrapolation ND flow could not be maintained without signs of showed that ND could be shut down % 4 hr before cavitation reappearing. The control board level the onset of boiling. This boiling would not be stuge indicated that NC system water level was 250-detrimental provided make-up was available (one 270 in. above the bottom of the ND pump suction Centrifugal charging pump was cperable) and any loops (175 in.was the minimum for ND operation).
vapor generated was not released to containment How1ver, a later check of tygon tubing connected to (vapor would be condensed in the PRT).
a loop drain and vented to a RTD manifold vent indicated that level was 4 to 6 in. below that lamediate corrective action was to raise level to sptcified for ND operation. Draining (to hot leg 175 in. and resume normal ND flow. After its N2 centerline - 175 in. on the level gauge) was being supply was secured, the PRT was vented to contain-accomplished per procedure, with an N2 overpressure ment atmosphere to insure the control board level on the Pressurizer Relief Tank (PRT) and Power gauge would indicate level accurately. Statements Operated Relief Valve (PORV) open to admit N2 to the were to be incorporated into the procedure " Draining pressurizer (PZR). This incident resulted from the Reactor Coolant System" warning that the 5-7 intecurate indication by the transmitter for the psig PRT overpressure could greatly affect the level gauge, due to a misapplication of the level levels indicated on the control board gauge. A trsnsmitter.
modification to have the reference leg of the level transmitter connected to the PORV discharge line To insure pump 1A could be restarted safely af ter was planned. This avoided problems associated with showing signs of cavitation, it was vented and the a pressurizer tap and was satisfactory since a PORV fuiling water storage tank supply to ND valve (a was required open during draining. The modification
^
w Jul 82
Val. FWR-2 IVI. Operational Problems C. Misc
- p. 714 N 2425. CS Vest Valve Opened - RER Water Dralmed verified. The closure of that damper had a direct Out - Persommel Error effect on the concentration of CO2 in the cable vault area. The last operability inspection of the N
NcCuire 1 - Jun 84 - 100% power fire danipers was conducted on 21 May.
Conduit was not found in the control room cable vault damper, A 3/4 in. vent valve in the containment spray (CS) but a section of conduit was found in a fire damper cystem was found open on 27 Jan, during a valve for the DG fuel oil day tank enclosure. The damper stroke timing test.
About 35 gal of water from the was in a partially closed position with the ob-RHR system drained onto the mechanical penetration structing conduit positioned perpendicular to the room floor. The test procedure required that each damper at the bottom sill of the opening. The con-chick valve in the system be determined operable by duit section then was removed and acceptable damper lifing the check valve disc slightly using instru-operation was verified.
The condition was not ment air.
If water was in the lines, the air pres-reported at that time except for a note on the aire could cause safety and radiological concerns original data sheet.
by forcing high velocity water out of the discharge g
vnt.
On 16 Apr, personnel placed a drain hose on An investigation revealed that personnel resetting g
scch spray header to determine if the header was the dampers after surveillance tests or actual frse of water.
A drain valve and vent valve were actuations occasionally used conduit to hold the cptned to perform the task.
dampers open, to aid in reattaching the latching mechanism. Not all ;ituations that required reset-It was theorized that when the header was drained, ting the fire dampers were covered by procedural manual, packless 3/4 in. vent valve (Kerotest) INS-controls.
There was no conclusive evidence as to 68 was not reclosed. No documentation was found to who was responsible for the blockages.
Admin substantiate the theory that the vent valve was policy changes were established to ensure that fire 1 sit open during preliminary draining an 27 Jun.
dampers were operable and unobstructed following Th'a pipe cap was found sitting next to the vent testing, maintenance or actuations that disturb a vz1ve, and a reducer coupler and a 1/2 in. nipple damper's readiness.
A safety evaluation that were attached to the valve.
Interviews with per-addressed the obstruction of the dampers concluded sennel who were in the same room as the vent valve that adequate fire protection existed in both and who applied air pressure to the spray header, instances despite damper closure problems.
(ugf) disclosed that no abnormal conditions were observed.
Applying air pressure to drain valve INS-39 with vent valve INS-68 open should have N 2427. BER Trains Simultaneously Placed In Racire crused high velocity audible air to exit the open Mode - Operator Error i
vent valve, if the vent valve was open.
- However,
/]
brckground noise might have prevented personnel N
Farley 1 - Feb 84 - cooldown
_/
from hearing tha noise.
Before entering Mode 4, on 10 Feb, the shift super-tha radioactive spill was cleaned up successfully, visor placed the RHR system in the recire mode to without any workers receiving an above-limit dose.
ensure chat the boron concentration in the RHR sys-Bowever, the performance test supervisor's shoe was ten was greater than or equal to that in the RCS.
l crntaminated when he closed vent valve INS-68.
That required shutting the RHR to RCS cold leg MOV Operations checked other vent and drain valves on for each train to prevent auto low head SI flow CS spray headers.
The pipe cap on INS-68 was without operator action. On 11 Feb, the supervisor reinstalled.
The importance of the removal and and the plant operator decided to place both trains rsstoration procedure in removing station equipment of RHR in recire. Normally, the procedure was per-nt covered by an established operating procedure formed I train at a time.
The decision was dis-tu s to be reviewed with all operators.
The cussed during shift turnover with the evening arrangement for non-operators to cycle valves in supervisor who recognized it as a Tech Spec viola-isolated systems and verify valve positions was to tion.
The trains were returned to normal service.
be discussed.
Procedures were revised to apply The day supervisor and plant operator were re-independent verification to vent, drain and test instructed and procedures were revised to state etnnection valves for train ' A'
& 'B' on the RHR &
that only I train at a time shwld be prepared for CS spray headers. Also, a review was to be done to cooldown until the unit was below Mode 3.
(uhx) ditermine if independent verification might be rpplicable to other header valves.
(uqe)
E 2428. Inadvertent SI - Wrona Vital Bus Synchronisina Switch Used - Operator Error 3 2426. Fire namper Obstructed by Conduit -
Personnel Error N
Surry 1 - Mar 84 - cold shutdown 3
Cook 1 - Aug 84 - 100% power While operators were rennving vital bus transformer (501.A glectric Co.) I-I on 1 Mar, vital buses I&
On 22 Aug, the control room cable vault supply fan III were mistakenly cross-connected out of phase, damper was found inoperable during the 18 no sur-which resulted in a voltage transient on the buses, viillance test of the carbon dioxide fire protec-causing spurious containment HP and high steam flow tien system.
An inspection found a short section signals, and the initiation of an SI signal.
The cf conduit positioned so that it prohibited the vital bus synchronizing switch for vital buses II &
drper from functioning.
The obstruction was IV had been used rather than the switch for the removed, and the operability of the damper was buses being crosatied.
- 84
Vol. PWR-2 XVI. Operational Problems C. Misc q
- p. 801 I
accidentally caused a short circuit. That actuated The shunt trip circuit was a back-up for the UV a relay and initiated the load shed feature of trip tircuit.
4160V bus IF.
All major loads supplied by that bus were de-energized.
To prevent recurrence, the The modification made in this event was to add a testing procedure was revised to require the use of separately fused power source to the shunt trip tools which were sufficiently insulated.
(vts) circuitry. The modification did not involve exten-sive wiring changes but did require internal cabi-net wiring changes. The changes on Unit I had been E 2673. Cantaimmat 1stearity Tiolated During implemented. The Unit 2 modification was in prog-Rydre Test - Simultammone Test And Draim ress when the specialist doing the work identified Of EER System &6J EER EI - Procedure Error a problem with the wire routing and questioned how the electrical separation criteria could be main-E Cook 1 - Apr 85 - hot standby tained. The technicians identified the train sep-aration problem when they noticed the differently
[
On 12 Apr, an equipment clearance was being placed colored cables entering the bottom of the cabinet.
j on the RHR system in order to remove blank flanges It was determined that the channel separation cri-which were installed on the safety valves for a teria had not been maintained during the Unit I hydro test.
An operator inside containment was in-installation and the RTBs were declared inoperable.
structed to vent and drain the system so the safety valve inside containment could be reinstalled. At The RTB cabinets were supplied with wiring troughs the same time, an operator outside containment was to provide electrical separation.
Wiring troughs to vent and drain the RHR HX so the safety valve at the top of the cabinet provided a, wiring path outside containment could be reinstalled.
That from the enclosed back of the cabinet to operating created a potential flow path between containmar.1 c e.At s in the front of the cabinet.
On the and the outside atu.
The operator inside ceatain-front side J the cabinet, train
'A' wiring was ment noticed a greater than expected flow of water routed from the cuOide edges of the cabinets into and notified the unit supervisor.
The supervisor the operating compones.O.
Train
'B' wiring was directed the operator outside containment to close routed down the center of cabinets 1 and 2 and out-the vent and drain to stop the flow.
However, the ward to the components.
Personeel were not aware operator inside containment closed the vents and of the wiring trough routing path in the top of the drains.
cabinets when the Unit 1 modification was being made. The train 'A' trough was difficult to access Although the system never completed draining, a and required removing cabinet panels when routing potential flow path was establisned between con-new wires. The procedures used by the technicians tainment and the outside atm AQ$, min.
The po-did not reference the installation specs or proper tential for obtaining the flow path also existed electrical' separation criteria. The wiring troughs during the flange installations but was not noted and wiring were not color coded or identified as at the time.
To prevent recutrence, the equipment train-related.
QC inspections during the Unit I clearance was required to include guidance on the modification did not reveal the error, installation and removal of the blank flanges, and the test was to include a precaution on maintaining While corrections were being made to the wire rout-containment integrity.
(vet) ing, an electricain accidentally depressed a limit switch inside the reactor trip breaker 'A' cabinet which initiated a main FW isolation signal.
E 2674. Reactor Trip switchaear Wirina out of sper -
Procedural, Persommel Errors An admin and procedural deficiency was found to be the cause of the incident because: 1) the wiring D
McGuire 1 & 2 - Feb 85 - shutdown, refueling change procedure used for the modification did not refer to the wiring separation criteria procedure on 6 Feb, an electrical separation criteria defi-or installation specs; and 2) the technicians in-ciency was discovered in the reactor trip switch-volved in the modification had not been trained on j
gear cabinets. The error involved a single train the installation specs manual.
A personnel error
'A' wire which had been routed with the train
'f' caused the unplanned main FW isolation signal, wiring during a previour electrical modification on 26 Mar 84.
The installation did not conform to the As corrective action, the procedure for wiring criteria which required a 6 in. free air space changes on systems and components was to be between redundant safety devices in separate trains enhanced.
The training dept was reviewing the when barriers were not used.
Installation spec and planned to develop the necessary training on applicable specs.
(vtz)
The RTB cabinet modification was a result of an NRC evaluation of another, previously installed modifi-cation which had added a shunt trip actuation cir-N 2675. Control acom Air Intake Radiation Element cult to the auto breaker trip circuits. The spe-Dypassed Too toma - Persammet Error J
cific problem with the original shunt trip instal-lation was that the safety-related shunt trip cir-E Wolf Creek - Mar 85 - initial fuel loading cuits shared a common fuse with nonsafety-related aux circuits.
The shunt trip circuit required on 13 Mar, control room air intake radiation power to operate and if the single fuse had blown elesent CK-RE-05 was bypassed for routine filter because of an aux circuit component failure, then rep!acement. Tech Specs allowed that condition for j
the shunt trip circuit would have been inoperable.
$ 1 br, af ter which time the control room emergency cosnem g mue Aug 85
Vol. PVR-2 Vill. Aux Systems B. Aux. Cooling N201.
CoRRooEn ToRoCE SWITCH, MAtio2USTEo POSITION
- p. 46.
LIMIT SWITCH ON SERVICE WATEh OPERATORS N
oconee 3 - Aug 78 - 1002 power 198.
SW PUMP BEARING WORN - COOLING LUBRICATICN On 22 Aug, the torque switch for valve 3LPSW-24 SYSTEM REDESIGNED failed, causing it to be inoperable in the closed position. On 23 Aug during the performance of a Kewaunee - Aug 78 - 95% power test the position indicator for 3LPSW-21 indicated An operator investigating intermittent vibrations that the valve had failed in an intermediate position.
These 2 apparent valve failures removed 2 of the 3 c ring from a service water (SW) pump (Wheelabrator-Reactor Building cooling units (RBCU) from service.
Frya) found the seal assembly was out of position.
A unit shutdown (102/hr) was commenced at 0830 hr Although the pump could be operated, it was taken as required by the Specs. At 0935 hr it was deter-out of service, placing the facility under an LCO. Only 2 SW pumps were needed during post LOCA mined that 3LPSW-21 was inits full open position and the apparent valve failure was actually a failure c nditions; therefore the 3 operable SW pumps of the position indicator and the unit shutdown was provided full capability.
terminated.
Tha failed SW pump was replaced with a spare pump The cauce of the inoperability of 3LPSW-24 was the cnd returned to service within the time allowed failure of the Limitorque valve operator torque by the LCO. The f ailed pump was found to have switch. It failed due to corrosion and it was worn bearings which allowed the seal assembly to replaced. The cause of the apparent inoperability thift. This failure was believed to be caused by of 3LPSW-21 was the failure of the valve's position insufficient lubricating / cooling water supply to indicator to properly show the valve's position tha bearings. A vendor design change to the pump bearing cooling water supply was being eval-because of maladjustment of the open/close limit switch (which was readjusted).
(gnd) urtid and was to be implemented when possible.
(gko)
N202.
RHR vAtvE STEx olSENcAcEo rRos oISC 199.
RHR PUMP SHAFT SLEEVE CASKET LEAKED, SEAL Surry 1 - June 78 - refueling FACES WORN buring Type C Leak testing, it was found that the Indian Pt. 3 - Aug 78 - hot shutdown 6 in. RHR discharge gate valve to the Refueling Water During preparation for plant startup, No. 32 RHR Storage Tank did not appear to be opening when act-uated. Further checking showed that no water would Pump Seal was found leaking. The pump (Ingersoll-Rcnd model 8X20W) was removed from service to pass the valve when the valve was opened manually.
idintify the cause of the leakage and No. 31 RER The valve was disassembled and was found with the stem disengaged from the disc and the disc fully Pump was verified operable. Upon disassembly of seated. During normal operation this valve was shut.
tha No. 32 RHR Pump Seal it was found that the shMt sleeve gasket was leaking and the seal faces It was opened only to provide a means of pumping v2ra worn. The pump seal and gasket were re-the reactor cavity to the RWST following refueling.
pltesd.
(gls)
The valve (Aloyco) had been previously disassembled during the outage for maintenance. The valve was 200.
RHR PUHPS AIRBOUND assembled by a qualified workman using an approved procedure, in the manipulation of reassembly the disc became disengaged from the stem. Investi-Beaver Valley 1 - Sept 78 - cold shutdown gation of the event indicated that the mechanic used due care Tha RCS was drained to the centerline of the in the assembly and was working accordiac to an approved procedure. The valve was reassembled cold legs. The primary side manways on SG's A and operated properly.
(ghg) cad B were removed for ISI and RCS temperature w:s M 1450F with the "A" RHR Pump in service when indicition of low RHR flow was received. The N203.
DC START TIME EXCESSIVE - COOLING WATER VALVE "A" RHR Pump was running with no discharge flow.
SEAT LEAKED Th2 "B" RHR Pump was started but it also produced n) discharge flow. An investigatien revealed b:th RHR pumps were air bound. The "A" RHR Pump Calvert Clif fs 2 - Mar 77 - 100T. power s: tis and casing were vented and the "A" RHR Pump During a weekly test DC No. 21 failed to reach w:s raturned to service N 1 hr later.
rated speed within 10 sec.
On 3 successive attempts,
!!ith mixing of the RCS re-established. Tavg the times to reach rated speed were 18.6, 16.0, a.d 11.0 sec.
The service water inlet valve v:s ditermined to be N1750F. The heatup from 1450F to 1750F for the water volume in the (Masonellan) 2-CV-1587, which provided cooling water to the DC was found to be leaking excessively re ctor vessel was consistent with the expected by its seat. This leskage resulted in low Jacket h utup rate for 2 MW of decay heat which w:s typical for 38 days into a shutdown from water cooling temperatures and thus was the apparent full power operation. The RHR pumps were for the sluggish operation of the diesel. The cause b2ing vented periodically and the RCS was to be valve was rebuilt with a new seat which corrected the r2fillid upon completion of maintenance activities.
leakage problem. The DC had was successfully test (git) started 5 times, over a 4 wk period, after this repair.
(gnh)
- s. - = e -=.
n my 78
f.. n t r <,, e t a b.
ru a u m. t tne -car rinn -ere rrt.uia& An Mat in C did wt valve 743 was installed.
propo nte around the impeller wear ring area.
Each of this new piping section, the mark h7,d a start and stop point. which was typical of 1
hydroid at 625 psig. Tests were some debris being trapped in the annulus. Shaft 4riormed on the new manual globe inspection by PT showed no indications. The inspection ed secticn was analyzed, and the covered the whole shaft and concentrated in the keyway to be cevitation and/or i: pinge-area near where previous failures occurred. Hub material y wts initiated to determine if was removed f rom the B pump impeller to further assure lud2 recurrence.
(gpe,hbs) adequate clearances as on the
'A' pump. NRC's review concluded the pumps were reliable and acceptable. (gsb)
~ FAILED - CASING I VEttER LASED 212.
SERVICE WATER VALVE ACTUATOR LINKAGE N"T MISSING 3 - Apr 7S - refueling Davis-Besse 1 - Oct 78 - 100% power e low flow condition existed The Control Room operator unsuccessfully attempted to Pump (DHP) 1-A.
Upon dis-open the CCW RX No. I Service Water Outlet Valve (Hammel varad that the pump shaft Dahl) SW 1424 to control the HX outlet temperature. He pillst end.
The pump mfr.
then attempted to manually stoke or operate SW 1424, sulted and pump shaft break-but it remained closed and CCW Loop No. I was declared initiated. Due to the severity inoperable.
nd occurrence (see VII.B.143) pump, the reliability and Upon visual inspection it was discovered that the nut ypi of pump was in question, on the actuator linkage for the valve was missing. The eps ba taken to establish valve had failed in the closed position since the ity end optrability of these positioner was not connected to the actuator. They replaced the linkage aru bolt and reset the stroke. Be valve operator was recalibrated, and the position Mit fcilure was primarily switches set. Af ter satisfactorily completing stroke so-irpsilar rubbing.
Sub-times, the valve was declared operable % Ik hr later.
rabuilt with a new shaft ca shived back to assure proper R Other SW valves associated with CCW HXs were inspected ctsing tnd the impeller. A for similar conditions with no problems noted. See
>rmed on this pump to establish R vIII.B.293 for adeitional information.
(as..hbh) 6xistad was no longer present.
1 iinid utilizing accelerometers in tha pump, connected piping 213.
LPSI PUMPS AIRBOUND FOLLOWING TRANSFER OF Thz tcquired data was reviewed PURIFICATION SYSTEM RESIN a d noic2 source identification
>th~r data sets to identify Calvert Clif f s 2 - Oct 78 - cold shutdown and trands in performance 41ts indictted that the A LPSI pump No 21 operating in shutdown cooling all ca th7 B pump. Also the (SDC) mode began cavitating and lost suction. LPSI n th) A pump casing showed pump No. 22 was immediately started but it too began 1sId on this data they cavitating and was secured. Both pumps were sub-ts ns non. lubricated metal.
sequently determined to be airbound. The pumps were tsd by attemely low noise vented and SDC was restored 2 hr later.
frsquincias, no high fre1uency n
Air binding of the LPSI pumps apparently resulted when air being used to transfer purification system resin leaked into the SDC system. To prevent recurrence of this event operating instructions were
)
being revised to prohibit transferring purification In July, the Unit 3 pump malfuncti system resin while the purification and SDC systems was dismantled and the shaft was f were cross connected.
(gsw) broken at the keyway. The pump wa tested satisfactorily.
214.
_ PUMP SEAL FAILED - HOTOR BEARINGS DAMAGED In July, the thrust bearings seize.
pump. The pump was dismantled and Surry 1 - Oct 78 - 100% power overheated thrust bearings. The b.
replaced and the pump tested satis Charging Pump Component Cooling Water Pump (Ingernoll Rand) 1-CC-P-2B tripped on overload.
The exact cause of the shaf t and b.
The redundant pump was placed in operation.
was not determined. It was suspect They found the overload was the result of corrosion have resulted from excessive interr damage to the motor bearings caused by water leak-by pump cavitation.
age from the pump seal. The failed pump was removed and disassembled for repair. The motor bearings A valve located at the SFP heat exc were replaced and the pump rebuilt with new seal is called the " system discharge lir material and new gaskets. The rebuilt pump was operation was very; noisy, indicatic valve". Operatie, personnel verifi reinstalled and tested satisfactorily.
(gvx) whenever this valve was > 50% open.
conducted an evaluation of system p 215.
SPENT FUEL PIT COOLING PUMP SEAL FAILED -
determine the optimum position of t INSUFFICIENT LUBRICATION - RELEASE charge line control valve. Evaluat indicated they could avoid cavitati Turkey Pt. 4 - Sept 78 - refueling provide cooling system flow to meet rent SFP heat transfer demand by ad:
A spent fuel pit cooling pump's mechanical seal limiting the valve position to abou failed causing leakage of radioactively open.
cerstamim w
_~
2%.
CORE FLOW I.0ST DUE TO RHR PUMP AIR BINDING Vol. PWR-2 VilI. Aux Systems Bsaver Valley 1 - Jan 80 - refueling B. Aux. Cooling
- p. 67 The plant was in Operational Mode 5 with RCS temp at 1000F and atm pressure. The RCS level was mid-loop with RHR flow at 1700 gpm. The reactor
.and full power operation was resumed while mainten-vessel vent eductor was in service in preparation ance ef forts to open POV-5 continued. At 0005 hr for refueling. A low flow alarm was received and low flow and low motor current were indicated.
on 11 Mar, POV-5 was opened and the SWCP C-13A placed in service. At 0010 hr, with the pumps C-13A A second pump was started and was also air-bound.
The pumps were vented, core flow was rapidly re-and C-13C determined to be operable, the pump C-13C stored and there were no implications to the health was stopped and unit operation proceeded in accor-dance with the 1,CO.
Throughout the incident, ade-and safety of the general public.
quate cooling was maintained to equipment normally Putting the vessel vent eductor system into service served by the CCW system during power operation.
The peak CCW HX exit temperature of 82*F was well was the root cause of the incident. The eductor was in service and caused a negative pressure in below the alarm condition setpoint of 97'F.
Tem-the RCS because the eductor air volume was greater peratures of pump bearings cooled by the CCW system were monitored during the incident and no signif t-than the influent air. This dp overcame the dP cant changes observed.
in the steam generator U-tube section and the excess steam generator water and entrained air drained into the loops. This entrained air later Investigation revealed that the SWCP C-138 shaft air-bound the pumps. The procedure was revised to failed due to apparent excessive vibration resulting constantly vent the RHR pumps when putting the from worn bearings. The pump shaft and bearings vessel vent eductor into service. This had been were repaired and the pump returned to service.
a recurring problem when loops were drained to mid-The failure of POV-5 to open automatically and the span while RHR flow was 3000-4000 gpm. With flow problem of insufficient priming of aux SWCP C-13C between 1000 and 2000 gpm, there had not been a were under investigation.
problem. A method was being researched to possibly constantly vent the RHR pumps automatically.
On 18 Mar during maintenance activities on the SWCP (Jam)
C-13B, pipe support 9 416-14 was found to be in a degraded condition. Corrosion was evident on the anchor bolts that attached the vertical support legs 205 SWC SYSTEM FAllt'RES - PIPE SUPPORT ANCHOR to the face of the concrete pedestal. These 2 BOLTS LOOSENED. VALVE & PUJP BEARINGS FAILED anchor bolts were found to be intact but loose in the concrete. Also a space ~ 1/4 to 1/2 in, wide San Onofre 1 - Mar 80 - 1007. power existed between the face of the concrete and the structual member. The pipe support was redesigned
-C At 2115 hr on 10 Mar with salt water cooling pump and rebuilt to transfer loads to the floor slab (SWCP) C-138 in operation, SWCP low flow and low rather than to the concrete pedestal. Following.
discharge pressure alarms were received on the 'nain completion of modifications made to the SWCP line support, SWCP C-13B was returned to service on 21 control roon aux board annunciator panel. Con-
- Mar, currently, SWCP C-13A automatically started due to i
low pressure in the discharge line of pump C-13B and pump C138 motor amperage was observed to be ir.-
Subsequent inspection of the corresponding pipe j
dicating low. They found both pumps were running support. 9-415-12, on the SWCP C-13A line revealed with discharge pressures of 0 and 40 psig indicated some signs of deterioration; however, this support
~
was in full contact with the concrete and determined at pumps C-13B and C-13A. respectively, and the pneumatically operated discharge valves (POV's 5 &
to be operable. On 12 Mar SWCP C-13A was removed
- 6) of each pump. were in the closed position. POV-5 from service to install a modified support. The 4
(discharge of pump C-13A) should have been open while modifications were identical to those made to support POV-6 (discharge of pump C-138) should have been 9 416-14 Installation was completed the same day.
closed. Efforts were then initiated to open POV-5.
Consistent with Tech Spec requirements 1 of the 2 SWCP's remained operational and the aux pump was At 2120 hr the aux SWCP C-13C was manually started from the control room. However, a low flow condi-tion as indicated by low pump motor amperage was observed. Investigation of the aux pump and piping 296.
RHR FLOW LOST WHILE INCREASING FLOW -
y system indicated that the low flow condition was PUMPS AIRBOl'ND, FLOWMETER RECALIBRATED due to apparent insuf ficient pump priming and the aux pump was then stopped. To re-establish SWCP reaver Valley 1 - Apr 80 - cold shutdown flow, the screen wash pumps were started from the local panel and valves manually aligned to discharge The SC's were drained and RCS level was at mid-to the bottom component cooling water (CCW) HX span in the loops. RCS temperature was 810F E-208 normally served by SWCP C-13A.
At 2133 hr, and the "B" RHR Pump was in service. A com-SVCP flow to E-20B was observed to be ~ 2000 gpm plete loss of RHR flow occurred when they were and CCW teeperature exiting E-20B was decreasinge attempting to increase RHR flow from 1000 to 3000 having reached a peak value of 62*F.
At 2156 hr' gpm so the RCS could be recirculated through adequate priming was restored to the aux SWCP C-13C, Mixed Bed Demineralizer CH-1-1A.
The resin in and the pump was placed in service
- CH-1-1A had been replaced and had not been used so a dilution of % 50 ppm was expected. Tech During the period described above, an 1.C0 was not Specs required 3000 gpm core flow when diluting.
cet and preparations were made to corr:ence an order-With RHR flow % 2500 gpm, the RHR Pump RH-P-1A ly shutdown of the Unit. However, at 2200 hr with ammeter started to oscillate and then dropped to zero. The operator then tried to start pump the aux SkCP restored to operation the Unit shut
- RH-P-1B but it also was air-bound. Operators down was terminated after a slight load reduction were dispatched to containment to vent both RHR pumps. RHR blow was restored within 35 min c wee e======== *==== a=== ea=
==
Mav 90
Vol. PWR-2 N299 SitTV4TeR/StXvrCe W4TtR 8x Test trAxro
$u C
N cat"n buns 2 - wu 80 - 7e rwu
- p. 68 Saltwater / service water HX (Foster-Wheeler) No. 22 with RCS temperature at 810F. Several subsequent was taken out of service to repair a leak. The HX attempts were made to increase flow; but, each time, was opened and a leaking tube was found and plugged.
the pump running current began to oscillate so The HX wa's returned to service within 2.2 hr. (jue) the operator returned flow to normal. Maintenance personnel were requested to check flow transmitter calibration and the operating procedure was changed O300 SWS FLOW TO CHARCINC PUMP COOLERS INADFQUATE to require an operator to vent at all times during this evolution. See VIII. B. 294 for a previous Q
North Anna 1 - Jan 79 - shutdown similar occurrence.
(jej)
N see vtti.B.304 for additional information.
A special test was conducted on the Sereice Water Sys-tem (SWS) to determine the flow characteristics of 297.
OPERATOR-TO-VALVE STEM LOCKNUT LOOSENED the system during accident conditions. The SWS was aligned to similate a LOCA on Unit 2 concurrent with a Robinson 2 - Mar 80 - cold shutdown cooldown ' on Unit 1.
The test results indicated that with a 95'F SWS, one p w p out for maintenance and fail-Component cooling water containment isolation ure of another to start, the design flow of 17 gpm to valve 626. RCP thermal barrier outlet, failed to the charging /high head SI pumps could not be met.
close during a test. The failure of the valve However, with only one unit operating and SWS tempera-to close was attributed to the loosening of the ture between 60 and 70'F. adequate flow to the charging operator locknut, allowing the operator shaft to pump coolers was available d aring accident conditions.
rotate freely and unattached to the valve stem.
Preliminary investigation indicated the pressure drop Valve CCW-626 was required to close as part of across the 2 in. carbon steel piping. charging pump phase 3 containment isolation. The locknut was skid piping, and lube oil cooler temperature control l
tightened and staked and the valve was operated valve was too,large to allow adequate flow. Subsequent electrically and returned to service within appropriate modifications were made to the SWS headers t 44 hr.
and associated instrumentation, piping for the HX's and instruments associated with the charging pumps.
l Various Limitorque SMB valve operators had main-(jvy) l tenance performed on them during the cycle VI-VII
]
refue.ing outage as follow-up to IE Circular 79-4 l
(see XV. 180) The operator-to-valve stem locknut N 30t.
SoCxtT-Wrua ttBm - rITT1NCS Sumrn CuCn -
1 for valve 626 was staked on 7 May 79 The lock-CENERAL MFC PROBLEM backed over the stake resulting in failure of N
nut the valve to close. The locknut was staked again and the repair was considered adequate.
St. Lucie 1 - Apr 80 - refueling
['
(
As additional action, a representative sample of PT of 45' socket-welded elbow-fittings (A182 sch. 40 1/2 the total locknuts staked during the 1979 refueling in. SS) in a non-class system showed cracks at the same outage wtre to be checked to verify similar location in 3 fittings. This was reported because the problems were not occurring on the other valve operators. Additional corrective action was then to fittings were certified for use in class 2 (safety-be taken af the results of these inspections indi-related) systems. The cracking appeared generic and PT might not always detect a crack. An undetected crack cated it,was necessary. These inspections and would lead to leakage. This was the first event of this any requtred corrective action were to be com-type.
Other FPL nuclear plants were notified and the pleted prior to the end of the 1980 refueling QA Dept, notified the vendor, outage.
(jsu)
The apparent cause was a generic mfg problem. All 3 fittings from heat code IPL, heat No. D20303, showed N m.
vitvr 02rRAT0a eso"Attrten tsvtROatsTAtty cracking. Three other fitting, (dif ferent heat) were -
ordered on the same Purchase Order but 2 were acceptable N
ocon e t - 'ar so - tor. Power nd the third was ta ccessible. The fittins, were or-t dered in 1977 No other deficient fittings were on site.
The otor operator (Rotork model 16 NA2) for valve ILP-105 (installed in 1976 as I of 2 flowpaths sided The inaccessible fitting was associated with the 1A In-to the decay heat line to assure dequate flow from take Cooling Water Pump (ICVP). It was located on the the RCS past-LOCA to prevent boron precipitation) lA ICVP lube water strainer drain line, downstream of was found to be unqualified for in-containment use an isolation valve and was open to atmospheric pressure.
during a review required by IE Bulletin 79-1B (see (jvm)
Vol. Ba -2. IX. E. 429). The valve operator was to be replaced during the next available outage.
(jud)
I l
- , e an so. -
' June
302.
DH PWP WT SITTIOV - LOW RCS VATER 1.EVEt. -
LEA %ING VALVE READJUSTFD Vol. PWR-2 VIII. Aux Systems Davis-Besse 1 - Apr 90 - cold shutdown B. Aux Cooling
- p. 69 On 18 Apr at 1915 hr they discovered that the RCS water level had decreased from 70 to 37 in. above the hot leg piping centerline as' read at 304 ADDITION TO VIII.B.296 re RHR PUMP the Re-fueling Level Indicator L1214 (This lower level corresponds to 78 in, above the top of the reactor Beaver Vsliey 1 - Apr 80 - cold shutdown core.)
The low RCS level caused an erratic decay I
heat -(03) flowrate due to inadequate DM pump suc-The SC's were drained and RCS level was mid-span in the loops. RCS te:nperature was 101'F and the "B" tion conditions. As a result. DH Pump 1-2 was RHR pump was in service. A complete loss of RHR flow e
stopped at 1920 hr.
At this time, the RCS level had dropped :o 35 in.
Operations began to check occurred while operators were increasing RHR Hx their isolation and drain valve lineup for DH Loop 1 flow by closing down on the Hx bypass valve.
When which was out of service a7d found water flowing beginning this evolution, pump flow dropped to t
from DM Cooler 1-1 Discharge Line Leak Test Connec-zero as the pump became air bound. At this time, tion Isolation Valve DH73, which had been opened the "B" RHR pump was shutdown and the "A" RHR Pump to f acilitate draining of DM Loop 1.
D973 was im-it too was shutdown. Both pumps were then vented was started.
"A" RHR Pump showed no flow also so 1
mediately closed. Further investigation of the loop isolation lineup revealed that D1 Cooler 1-2 plus the RCS loops levels were increased. The "B" s
Discharge to Makeup and Purification System Iso-punt was then started _satisfactority and normal lation Valve DH62 was 2 turns open. This talve, conditions were reached within 22 min.
RCS temp-normally operated by a manual remote valve opera-erature never increased > 907, during the incident.
tor, had to be closed at the valve stem. RCS makeup water was transferred from the Borated The apparent cause of the incident was that the Water Storage Tank (BWST) to the RCS via the DH pump was air bound. This incident sas not attri-i System to restore the RCS level to 75 in.
buted to lack of venting because the flow was being DH flow to the RCS was restored when DH Pornp 1-2 was re-diverted, but not increased. A procedure was im-Plemented for a total loss of RHR flow and a con-started at 1949 hr.
Tripping D9 Pump 1-2 placed tinous vent hose was installed and was to be put the unit in violation of the Tech Specs which
{
into operation as soon as possible (jyw) required that while in Mode $ at least one reactor coolant loop must be in operation with an associated q
P.P or D3 pump.
The DH pump was off for 29 min and N 30$.
M POSimN LIMIT SWITCHES RQUIRED US temperature remained significantly below the 1
ADJUSTMENT saturation temperature, rising frora ~ 93 to 103*F.
N-St. 1.ucie 1 - May 80 - refuelins The occurrence wss attributed to companent failure of the valve operator.
The remote valve operator (RVO) for DH62 was found to be out of adjustment See XI.A.400 for information.
(kbx) 3 ca ning the valve to remain I to 2 turns open when the RVO was on its close stop.
DH62 was to be re-N 306 j
adhsted.
VALVE OPERATOR RADIUS ARM OBSTRUCTED BY (jxf)
MOUNTING BOLTS j
N
- 303, VALVE OPERATOR I.IMIT SWITCH MISADJUSTED Davis Besse i - Apr.s0 Zion 1 - Apr SO - 100*. power While attempting to switch to 42 Decay Heat Re-3 moval Train for maintenance on #1. Decay Heat Re-moval Cooler No. 2 outlet flow control valve DH14A During surveillance both service water motor operat.
would not open beyond 207..
The improper operation ed valves (MOV's) to Aux W (AW) pump IC backup of valve DH14A was attributed to a mechanical com-supply failed to stroke properly. The Tech Specs Ponent failure. The proximity of the radius arm tied the operability of the pump to that of the of the valve operator linkage to the bracket mount-valves.
Because 1A AN pump was already out of service for maintenance 2 of 3 AW pumps were not ing bolts allowed only marginal clearance.
Any movement of the radius arm towards the mounting operable, as required by Tech Specs. One of the Limitorque (Type SMB Size 000) Limit switch rotors bracket caused binding and limited movement of the radius arm. The radius arm on the valve shaft was of IMOV-SW 103 was found out-of-position.
readjusted and the valve was tested satisfactorily, moved away from the mounting bracket, thus provid-It was ing clearance and allowing the valve to stroke-They investigated the total operating circuit of fully. After meeting stroke time acceptance cri--
IMOV-SW 105 and found no problems.
(jyj) teria. the valve was declared operable and return-ed to service ~ 20 1/4 hr later. They were in-vestigating further corrective action that might be required on this or other similar valves.
(kck)
- *** *== se-. s July 80
Vol. PVR-3 Vill.* Aux Systems B. Aux Cooling steam void in the system. He also noticed the A
- p. 91 laop wiue range hot leg RTDs were reading % 300'F, which was slightly above the 50 psig saturation temp. Spray was terminated and emergency FW flow information provided regarding the Crystal River was initiated to the A pnce through SC (OTSC),
and McGuire events in order to ascertain it they cooling the A loop hot leg. About 7 hr later suf-g factent cooling was achieved to drop pressurizer considered during review and revision of natural.
level. The plant depressurtzed without void form-
- circulation and shutdown cooling procedures
- I'"*
that were required in IC Circular 80-15, and I*
on 2 Jun, a similar event occurred at McCuire 1.
The unit had not achieved criticality but had heated operator was to be informed of the matters dis-cussed l~ the circular. 3) Licensees should con-up using RCPs.
While reducing RCS temp and pres-sioer including the information in training and re-sure to achieve a cold shutdown condition, a steam training classes.
(myp) void was apparently formed in the RPV head area when the system was vented. There was no decay heat to drive natural circulation as there would 403.
MINOR SV LEAK AT CONTAINMENT VENTil.ATION have been for an operating plant. The system was COOLING COIL DRAIN VALVE repressurized, the RCPs pumps were restarted to uniformly cool the system,and the plant was taken Cook 2 - Jul 81 - refueling to cold shutdown.
These events were apparently caused by insufficient 4 small SW system leak was found in the contain-cooling of the large masses of hot metal in re-ment ventilation cooling coil drain valve. The gions such as the RPV head, upper J leg (B&W),
leakage amounted to 3 drops per 5 seconds from SG walls, and RPV nozzles prior to initiating a bonnet to body leak on a I in. drain valve.
DHR cooling. Local stagnation and stratification They tightened the loose valve bonnet and.on the advice of one of the on-site NRC resident of the fluid in the upper head region while on DHR cooling may also have contributed to the pro-inspectors, reported this to NRC within 24 hr l
blem.
as required by IE Bulletin 80-24.
They requested clarification of the reporting requirements of IE Bulletin 80-24 from the NRC.
(azl)
While the coolant passing through the core was being maintained relatively cool by the DHR sys-tem, coolant in the RCS hot legs and in the up-per head region was essentially stagnant. This 404.
RRR PUMP CAVITATgD - INDICATED RCS 3
allowed the coolant to be heated to saturation LEVEL WRONG j
temp, or to remain at relatively high temp because of the stagnant conditions. This resulted in steam Trojan - Jun 81 - cold shutdown void formation when the system was depressurized.
The operators had correctly diagnosed the systes While reducing RCS level, with the RCS level voiding at both Crystal River 3 and McGuire I standpipe indicating % 66 ft, the RHR pump and took appropriate corrective actions to bring was cavitating as indicated by fluctuating motor current. Diverting to the hold up tank the plants to cold shutdown.
was terminated and, when the motor current did While these events were not a concern because of not stabilize, the RHR pump was stopped and the availability of the DHR system at both Crystal RCS charging was established.
River and McGuire and the lack of core heat pro-duction at McGuire, the NRC believed transmittal The standpipe level continued to decrease, and of the information was appropriate, since voiding further investigation found the pressurizer in the RCS was a concern if operators fail to vent valve isolated. The vent was opened and recognize these conditions. Also, normal natural standpipe level began dropping rapidly. The circulation flow was reduced by voids in hot legs RPV head vent was opened and standpipe level and sufficiently large RPV voids.
stabilized at 4 61 ft. The RHR suction tapped off the RCS piping at just above the 60 f t level.
IE Circular No. 80-15. Jun 80, and NRR Ceneric Previous experience had indicated that cavitation Letter No. 81-21, May 81, were directed at the may occur when RCS level approached that point.
possibility of voids being formed in the RPV head region during natural circulation cooldown. It was An attempt to restart the RHR pump failed due considered important that reactor operators to air entrained in the RHR suction line. The recognized that voiding could occur in other por-RHR hot leg suction isolation valves were closed tions of the RCS (e.g., in the "J-leg" of B&W and the RWST suction valve was opened in order reactors) and under conditions other than natural to provide a positive suction head to the RHR circulation cooldown ao that timely and correct pumps. Both pumps indicated flow when jogged action could be taken.
and the normal valve lineup was restored. The RHR pump was successfully restarted 75 min after NRC recoassended these actions for holders of an it was stopped.
operating license and near term operating licensees
- 1) Review operating procedures dealing with plant The cause of this occurrence was determined to cooldown at.d emergency and/or abnormal procedures be personnel error. The procedure for draining that address natural circulation to assure that the RCS had specified that the pressurizer vent sufficient information was available for operators valve be opened. However, personnel performing to recognize the symptoms of RCS voiding and take the evolution did not check to ensure that the appropriate actions to recover from a voided valve was open,and without the proper pressurizer condition. Special attention was directed to the vent path lineup, the RCS standpipe level would w e==p==a====
Aug 81
Vol. PWR-2 VIII. Aux Systems L. Aux Cooling
- p. 92 409.
SW CONTROL VALVE SEAT DAMAGED hive read higher than the actual RCS level. The immediate corrective action was to properly vent and While trying to adjust SW flow through No. 12 com-reestablish level in the RCS.
Long-term corrective Ponent cooling water heat exchanger (CCW Hx), the ections were to review the pertinent plant procedures operator noticed that flow control valve 125W127 would for adequacy and to review the importance of following not open and the valve was declared inoperable, approved plant operating procedures with the opera-The flow control valve was a Fisher Controls Co. 16 tions personnel.
(m2p) in. Vee-Ball, type 478-6-16-U-CAV5 cont rol hollow ball sealing against the open end of a tube bundle with a smooth, curved sealing surface. This sealing 405.
SW PUMP DISCHARCE PIPE CORRODED surface was found to be ma rred, restricting move-ment of the valve.
No. !! CCW Hx was placed in ser-Millstone 2 - Jun 81 - hot standby vice, and No. 12 CCW Hx removed from service. Valve 12SW127 was disassembled, and the tube bundle was Both SW headers were cross-tied to repair a small leak cut back to remove the marred surface and allowed on the "C" SW pump discharge pipe. The leak developed free operation. The' valve was reassembled, tested between the pump and pump discharge valve due to cor-satisfactorily, and returned to service within 22-1/2 rosion which was believed to be the result of coating hr.
The valve af r was being contacted for further failure on the carbon steel pipe. At that time, the R evaluation. Subsequent design changes eliminated the alternate SW pump was not available due to its strainer need for 12SW127; it was removed and a new Hx was being replaced.
(nam) installed.
(nec,pwd) 406.
VENTILATION UNIT SW COOLING COIL LEAKED Cook 1 - Jul 81 - refueling 410.
SW LEAKED AT CFCU DISSIMILAR METAL WELD During a surveillance test they discovered a pinhole Salem I - Jul 81 - 96% power Isak in the non-essential SW system to instrument room ventilation unit (American Air Filter) 1-HV-CIR-2.
During a routine containment inspection, SW was dis-The leak was in the integral return header of the cop-covered leaking (% 1.5 gal /hr) from a 3/4 in. vent cor.-
ptr cooling coil which was found to have a minor imper-nection weld on No. 15 containment fan coil unit fsetion, believed to have resulted from the mfg pro-(CFCU). The CFCU was removed from service and iso-cess. The affected area was repaired by silver brazing.
lated. The leak was found to be in a weld between a Characteristics of the defect did not indicate a generic 3/4 in. carbon steel pipe and a SS flange. There was problem and no further action was plannc.d.
(nat) an existing design chenge request to change the dis-similar metals in the system to all SS.
The 3/4 in.
carbon steel pipe was removed and replaced with SS.
407.
DECAY HEAT PUMP COOLING FAN MOTOR BEARINGS No. 15 CFCU was tested satisfactorily and returned to FAILED service within 33 hr.
The design change request was completed on No. 11, 12, 13 and 14 CFCU.
No. 15 CFCU Crystal River 3 - Jun 81 - 100% power was partially completed and was to be finished during It was discovered that cooling fan (Trane) AHF-ISB for decay beat closed cycle cooling pump IB was exces-sively noisy and the fan was shut down.
Investigation 411.
LEAKINC NESV CHECK VALVES TO BE REPLACED WIDI revealed worn bearings in the fan motor. The sealed DIAPHRACM TYPE VALVE bearings were replaced. The bearings were also re-placed in AHF-15A as a preventive measure.
(nax)
Cook 1 - Jul 81 - refueling While performing the B & C leak rate test, several 4.08.
CCMX DRAIN LINE NIPPLE IMPROPERLY INSTALLED valves exhibited excessive leakage causing the total leak rate to exceed the limit imposed by Tech Specs.
Calvert Cliffs 1 - Jun 81 - 84% power The leak rate testing indicating that valves (Marlin Mfg.) in the Non-Essential Service water (NESW)
To prepare to clean the tube side of the Struthers-system were the major contributor to the excessive Wells No. 11 component cooling HX (CCHX), No. 11 leakage rates. The excessive leak rates were attri-saltwater header was placed out of service. However, buted to a combination of sand deposits on the seating while pumping saltwater from No. 11 CCHX to No. 12 CCHX, surfaces and erosion of the valve seats. The valves the drain line on No. 12 CCHX became loose and began were repaired by cleaning and lapping the seating to leak.
No. 11 saltwater header was then returned surfaces, or gasket / valve replacement. All other to service and No. 12 header was placed out of service valves experiencing excessive leakage were repaired to repair the leak. It was found that the drain line and retested with acceptable leak rates. Thirteen of the separated from No. 12 CCHX due to the pipe :Qple not 14 NESW check valves were to be replaced witn a dia-teing installed properly. The nipple was rethreaded phragm type valve. An engineering review was in and reinstalled correctly.
(nay) progress for the remainder of these valves.
(nfa)
- e..r
_s Ju? 85
Vol. IVR-2 XVI. Oper. Problems C. Misc.
all parsonnat not to mova tha sstting unless cuthirissd p.
174 by the shift foreman. The OWP regarding this limiter and the governor system was to be revised to N m.
sFRvtCt VATFR SYSTEM RtDUNDMCY LOST BY incorporate a test to ensure proper setting on the IMPROPER TACCING limiter prior to returning the pump to serv 1Ct.
N min, vanute - May 78 -,3s,ove, a
One of t he 2 redundant Service Water system headers N 'os.
oRiin vAtv tart OPRn - RCS rtoW lost l
was incorrectiv !solated and tagged out to allow minor l
maintenance. The resul t of the incorrect tagging Trojan - Apr 78 - cold shutdown order was the loss of redundancy of the Service Water pumps. Within 3 hr the incorrect lineup was RCS flow was being supplied by the RER pumps with discovered and corrected. The operator who the plant in a partially drained condition.
In-wrote the tagging order failed to realize the signifi-mediately prior to this occurrence the RRR pump cance of isolating the header and both that header's discharge flow had been shifted from reactor cool-pumps. The incident was discussed with the personnel ant loop 1 and 2 cold legs to loop J and 4 cold involved and other operators were re*ninded of the legs. The Control Operator did not observe any significance of maintaining redundancy in all safe-abnormal indications during this evolution or in-I guards systems.
(fyp) mediately after it.
In addition, an operator was monitoring the RCS level standpipe inside the Containment and did not notice any abnormal level N m.
ArP TRieeto ou OvtRSetto - covtRs0R tmTtR changes. About 2o.in after the RaR flow wa.
READJUSTED - VALVING ERROR shif ted, the Control Operator received a report N
that level was increasing in both Containment sumps.
Robinson 2 - May 78 - 1007, power Concurrently the Control Operator observed oscil-lations on RRR pump motor currents. He launediately Mechanical Maintenance requested a clearance of valve restored RHR flow to all 4 reactor coolant loops V2-16A (discharge valve from the motor driven Aux and commenced charging
- makeup water to the RCS.
FW Pump (AFP) to SC A).
The steam driven AFP and Motor current oscillations were observed to be associated valves were tested as per the Operating increasing and the PRR pump was secured. Flow Work Permit (OWP) and an operator was dispatched to dropped to < 3,000 ppa (the Tech Spec limit.)
l turn off the breaker for valve V2-16A and tag it out The RER pump was restarted within 1 min and RCS flow of service. During this period it was realized that was reestablished.
an ISI timing test of valve V2-14A (steam driven AFP discharge valve to SC A) was needed and this test An investigation revealed that a drain valve on the was implemented. Mean.shile, the operator in the field SI test line near valve CV-8879D was open. To pre-noti fied the control room that valve V2-16A breaker vent re*urrence all vent and drain valves reposi-was off. Minutes later as valve V2-14A closed at tioned for work done under a system tagout were the end of its timing test, the steam driven AFP to be tagged to insure they were returned to their tripped due to overspeed. This resulted in the correct positions following the work.
(fakt inability to feed SC A with any AFP's.
Normal feed at the time was being supplied by the main FV pumps. A violation of a LCO was believed to have N-
"'S ""'5 5""" '"" - '*' 5 5"** 5"c" occurred and a reduction in power of 3/47./ min was MtSPOSITIONED started. The breaker for valve V2-16A, which had not been disabled, was returned to service. The Zion 1 - May 78 - 89% power power reduction was halted at 957, and the plant was returned to full power. It was later determined Vhile testing one of 4 pressurizer pressure Trans-that no violation of the limiting condition for mitters (IPT 457) the control selector switch was operation by Tech Specs did result until valve mispositioned. This caused the spray valves to V2-16A was returned to service.
open and RCS pressure to drop to 1930 psi (Tech Spec limit 2205 psih Pressure reduction stopped Daring previous maintenance of the steam driven when the control selector switch was placed in AFP overspeed trip, the limiter on the mechanical the proper position. Pressure was restored in 15 governor had been adjusted high so the overspeed min. Instrument Mechanics were reinstructed trip could be tested. The limiter adjustment on the on the importance of following procedures.
I Woodard governor was a knob type screw adjustment which ifal) could limit the speed of the steam turbine driving the AFP.
It was normally set to some value below
- N 6ai-oc ott PRzssuRz cAct t.luz tzArzo rottowtuc tha overspeed trip. This was done to prevent the CALIBRATION - LOOSE FIITING speed surge on starting, or when suddenly going to reduced flow condition which caused the speed to N
Beaver Valley 1 - June 78 - cold shutdown approach the overspeed trip point. This limiter speed was normally not reached because the turbine The lube oil pressure gage on the GM DC No. 2 was speed was regulated below the limit by the pump dis.
removed, calibrated and reinstal kd on 1 May 78.
charge pressure input to the governor. It appeared During a test on 1 June the generator was shut down that af ter the above test, however, the limit ad.
and declared inoperable because of a lube oil leak justment of the governor was lef t set too close in the engine control panel. The leak resulted to the trip speed. Therefore, when flow through from a loose Swagelock connection on the lube oil this pump was reduced suddenly, during timing of pressure gage apparently caused by not properly velve V2-14A, the pump did overspeed and tripped.
reinstalling the pressure gage after calibration.
The instrument line was installed correctly and DC The speed limiter was readjusted to its correct No. 2 was returned to operable status. Personnel setting and the system was returned to service.
were instructed to verify correct instrument instal-The speed limiter had a sign near it to advise lation prior to returning the equipment to service.
A functional check was to be performed following calibrations to insure system integrity.
(fro) ee-Ju1y 18
ENCLOSURE NO. 4 TO SBN-1207 SEABROOK STATION RISK MANAGEMENT AND EMERGENCY PLANNING STUDY, PLG-0432, DECEMBER 1983 i
i l
I 4
4 4
4 I
l i
1 4
i f
j-i
SEABROOK STATION RISK MANAGEMENT AND EMERGENCY PLANNING STUDY, PLG-0432, DECEMBER 1985 The following are responses to certain questions. that arose out of recent discussions about the V-sequence in the above ref erence.
W'e have chosen to respond by:
1.
Reviewing the relevant scenario phenomena.
2.
Considering single. valves that separate the RHR system from othe r systems and f rom potential release paths.
3.
Describing new scenarios that arose out of 1 and 2.
4.
.Providing order-of-magnitude estimates of the frequency of the new scenari os.
1.
REVIEW OF SCENARIO PHENOMENA e
A V-sequence is initiated by leakage of reactor coolant through valves that separate the reactor coolant system from the RHR system.
Leakage is of concern only if it is large enough to cause loss of inventory to exceed the i
4 l
4 5
- a..
pm makeup capacity of the CVCS (about 150 si
).
Leakage of sufficient size would initially reveal itself by causing R relief valves to open.
Reactor coolant would discharge through two relief valves in the RHR hot leg suction lines, whose setpoints are at 450 psig, into the pressurizer relief tank inside the containment. At'2,250 psi, these valves are capable of relieving a few thousand gallons per minute. Reactor coolant could also be discharged through four small relief valves, whose setpoints are 600 psig, in the cold leg injection lines during certain scenarios. These scenarios include those in which the RHR system remains intact and those in which leakage is greater than the relief capacity of the suction line relief valves. The cold leg relief valves discharge into the primary drain tank of the boron recovery system in the waste process building. At 2,250 psig, the four valves could relieve a total of about 450 gpm. At slightly above their setpoint, the valves are rated at 20 gprp each. At 600 psig, therefore, their total capacity would be less than 5% of the discharge through the two larger suction line relief valves. As the primary system pressure decays, these valves would be the first to and their discharge would essentially stop.
Continuous discharge through the suction line relief valves would eventually cause overpressurization of the PRT and begin to raise pressure in the containment. Therefore, even without further breaches of the RHR pressure boundary (e.g., pipe or seal rupture), the primary' system exhibits conditions symptomatic of a small break LOCA.
For small leaks, flow through the RHR system and out of the relief valves is governed by the valve rupture size. RHR system pressure is determined by the relief valves.
~
As the rupture size increases, causing the RHR system pressure to exceed about 700 to 800 psig, relief valve flow would become choked. The suction line relief valve discharge, which has an equivalent flow area of about 6 square inches, would determine flow rate, and the valve rupture size would determine the pressure.
The worst scenario would be a sudden, catastrophic failure of a pair of valves separating the RHR system from the reactor coolant system in a way that instantly causes a leakage size equivalent to the RHR system pipe diameter.
Previous work has shown that dynamic shock effects cannot cause more system j
pressurization than the equivalent of reactor coolant system pressure.
Therefore, the most likely damage scenario would be limited to that which is caused by hydrostatic pressurization to the primary system pressure.
In the worst scenario, initial flow out of the relief valves ~ is equivalent to a large fraction of the pressurizer volume per minute. Therefore, an "S"
signal on low pressurizer pressure and ECCS injection would be likely in.less than 10 seconds. Furthermore, the containment would pressurize to the 18-psig containment spray actuation setpoint in about an hour..The RHR pumps 2
r
%? -
~
i are not likely to be available because of pressure and temperature conditions well beyond their design basis.
MAAP calculations were performed for this severe situation with the added postulate that both RHR pump seals ruptured, creating an additional leakage path out of the system equivalent to 2.6 square inches. These calculations indicate that high pressure injection would occur in about 5 seconds, the PRT rupture disk would fail in less than 30 seconds, containment sprays would initiate in about I hour, and RCS pressure would decay to below the injection line relief valve setpoint in less than 12 minutes and to about 250 to 350 psi in a little more than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. During this time, approximately 2,500 gallons could be delivered to one of two primary drain tanks. Each drain tank has a capacity of 8,700 gallons. The RWST would not be expected to be depleted for about 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The primary system pressure would remain _ above 250 psi for this entire time without operator action to depressurize and cool down.
The event would be considered terminated by reducing the leak rate out of the primary system to less than 150 gpm. This is achieved either by isolating the leak or by causing RCS pressure to decrease below 50 psig. Emergency procedure ECA 1.1 instructs the operator to reduce reactor coolant temperature and pressure to below saturation at 1 atmosphere by continuing to dump steam from the steam generators.
The above discussion was intended to point out a few important features of the V-sequence. First, reactor system and, therefore, RHR system pressure is at its maximum of 2,235 psig for only a few seconds. The hydrostatic stresses experienced by RHR system components decay away quite rapidly, indicating that subsequent failures become less likely following the initial valve leakage.
Second, the response of the plant and its operators may initially be that for a small break LOCA.
If no breaches of the RHR system pressure boundary occur, the event may indeed be treated as a LOCA. Third, a secondary heat sink is required to remove decay ~ heat'during the LOCA and to terminate the' event by- - " ~-
~
depressurization of the reactor _ coolant system. Fourth, it is unlikely that the RHR pumps will be available for recirculation from the containment sump.
Finally, we believe that a simultaneous breach of more than one location in the RHR system is highly unlikely and was not modeled.
2.
RHR SYSTEM PIPING INTERFACES After breaching the interface between the reactor coolant system and the RHR system, the entire RHR system could become pressurized. Essentially every point in the system experiences some increased stress. Altho' ugh it is generally believed that the RHR pump seals are somewhat less able to withstand the pressure than piping, valve and pump casings, and heat exchanger tubes, it is interesting to consider the potential for reactor coolant discharge into-systems that connect to the RHR system and are separated from it by valves.
Table 1 delineates the type and normal position of valves that separate high design pressure. systems from the RHR system.
3
5 TABLE 1.
VALVE INTERFACES: RHR, CVCS, AND SAFETY INJECTION System Valve Position Chemical and Volume Control System CS-V502 Locked Closed Manual CS-V828 Locked Closed Manual CS-V496 Check Valve Safety Injection System (train A)
RH-V35 Normally Closed Interlocked M0V (train B)
RH-V38 Normally Closed Interlocked M0V The CVCS and safety injection systems are not susceptible to failures even if the interfacing valves are breached because they have a high pressure design.
Figure 1 shows the interface between the RHR and low design pressure systems and components: the containment spray system, the RWST, and the containment sump. High pressure water entering the containment spray system is unlikely to significantly pressurize it because it is an open system.
In fact, discharge into the containment through the spargers. is a likely outcome.
Similarly, discharge into the containment sump would dissipate the threat of release outside the containment. Discharge into the RWST requires failure of yet another closed. check valve and failure to close a normally open motor-operated valve.
4 3.
ADDITIONAL SCENARIOS
. The event trees contained in the Seabrook EPZ study have been depicted in the form of event sequence diagrams (ESD) to facilitate the visualization of accident-scenarios.- Figure-2-displays the scenarios caused by leakage paths
~
from the RHR suction line (called VS), and Figure 3 displays the scenarios caused by leakage paths from the RHR cold leg injection lines (called VI).
In the process of developing the ESDs, additional scenarios, which were not covered by the event trees in the above reference, were included. The additional scenarios had their genesis in the above-described reviews of relevant phenomena and RHR system interfaces. These reviews were initiated by questions that arose from recent discussions about the study. The new scenarios are shown in Figures 2 and 3 inside dotted lines. The discussion already presented in the reference is still applicable to those scenarios outside the dotted lines.
The ESDs first question if piping and heat exchangers remain intact; if they do not, mitigation is conservatively assumed to fail and core melt is assumed.
If they remain intact, the state of certain check valves is questioned.
If either check valve CBS-V55 or CBS-V56. is open and remains open (or ruptures), a leakage path into the RWST could occur.
If either check I
4
D valve CBS-V25 or CBS-V26 is open and remains open (or ruptures), a leakage path into the containment building spray system could occur.
If both sets of check valves remain closed and intact, the ESD questions the integrity of the RHR pump seals. The scenarios that follow this question are identical to those in the reference.
If RHR piping and heat exchangers withstand pressurization, the ESDs (see Figures 2 and 3) question whether reactor trip, ECCS actuation (S-signal), and the automatic or manual start of the ECCS pumps occur. Although failure of any one of these is recoverable and emergency procedures are provided for their occurrence, it was conservatively assumed that either loss of an S-signal or failure to start ECCS pumps would lead to core melt. Core melt would occur early with a dry containment. Since release paths would occur either through a water-filled RHR vault or through the RWST, these scenarios were assigned to the IFPV plant damage state. Failure of reactor trip was logically transferred to an ATWS event tree.
As described above, cooling of the steam generators would be necessary. The diagram questions its availability. Auxiliary feedwater pumps, main feedwater pumps, and steam dump and atmospheric relief valves could be used. Success would involve any situation in which at least one pump and either steam flow path operated. Although loss of steam generator cooling could occur any time during the V-sequence, it was conservatively assigned to an early core melt plant damage state, 1FpV.
Additional scenarios were developed to investigate failures of the interface
~
valves depicted in Figure 1 for scenarios in which the RHR p'iping and heat exchangers remain intact.
The ESD questions whether the check valves (V55 and V56) are initially open or initially seated._ _It then questions if the valves fail open, given each initial condition. Two separate paths are provided to represent two different check valve failure modes. Should check valves fail, a normally open motor-operated valve operable from the control room is available to prevent leakage into the RWST (see Figure 1). Operators must defeat an interlock to close this valve. Should the M0V remain open, a loss of fluid into the RWST would result.
It is judged to be extremely unlikely that the RWST would fail in a way that compromises its ability to provide water to the charging pumps.
The RWST vent has adequate capacity to relieve pressure buildup due to the insurgence of reactor coolant at well over 30,000 gpm. Charging pumps would be available to make up loss of inventory for valve ruptures equivalent to an 8-inch diameter break. In fact, since the reactor coolant system is discharging into the RWST,' a recirculation loop would be established so that RCS inventory could be replaced for an indefinite period of time.
Nevertheless, failure of the operators to prevent leakage into the RWST was conservatively assigned as a core melt, with a release path through water.
The 1FpV plant damage state was assigned.
i 5
h 1
Figure 1 indicates that isolation of the RWST via MOVs CBS-V5 or CBS-V2 leaves a source of water for both the charging and safety injection pumps. The figure also indicates that the containment building spray system would be the likely path for a discharge of RCS inventory following failure of CBS-V55 or CBS-V56 and closure of CBS-V5 and CBS-V2. Since the spray system is open through the sparger into containment, this scenario would be equivalent to a LOCA inside the containment.
Failure in an open position of either check valve CBS-V25 or CBS-V26, which separate the RHR system from the containment recirculation sump and the
~
containment spray system, would yield a LOCA inside the containment. If normally closed CBS-V14 and CBS-V8 remain closed, Figure 1 then indicates t
discharge would occur into the containment spray system. Otherwise, discharge would occur into the containment recirculation sump inside the containment.
The ESDs also question the ability of the RHR pumps to function. Essentially, the entire RHR system would pressurize following the initiating event and tend to close the check valves at the discharge of the _RHR pumps. Because the FCV-610 and FCV-611 are normally open, the discharge check valves would tend to close for either the discharge line (VI) or suction line (VS) sequence.
Following an S-signal, the RHR pumps would attempt to operate against the closed valves. Since they were not designed for " dead head" pumping or RCS temperatures, the quantification of the following scenarios assumed failure of~
the RHR pumps.
4.
ORDER-0F-MAGNITUDE ESTIMATES OF ADDITIONAL SCENARIOS
~
We have not performed a detailed quantification of these scenarios. We have instead assigned order-of-magnitude, point estimate frequencies to each event to provide an indication of their significance. We believe that the assigned values are conservatively higher than the mean values that a detailed uncertainty analysis would reveal. The point estimate analysis-indicates that j
the sum of the additional core melt scenarios would not be a significant
~
contributor to the frequency of the 1FPV release category. Furthermore, the added ATWS and LOCA scenarios would not be significant contributors to their i
respective initiating event frequencies.
Conservative values for demand failure of reactor trip (RT), S-signal (S),
ECCS pumps (EPMP), and cooling to the steam generators (SHS) are P(RT)
= 10-4/ demand.
P(S)
= 10-4/ demand.
1 j
P(EPMP) = 10-4/ demand.
P(SHS) = 10-4/ demand.
l i
6
h-Our estimate of P(EPMP) and P(SHS) recognize the ability of the operator to initiate them should automatic actuation fail. There is adequate time for operators to refer to procedures E-0 and E-1 and the appropriate functional Since response procedures to initiate ECCS or cooling to the steam generator.
these scenarios do not involve a piping rupture, the most likely situations would require only one high pressure pump. The flow equivalent of one AFWS pump would be enough for steam generator cooling. Main feedwater pumps could also be recovered and either steam dump (to the condenser) or atmospheric dump valves may be used.
We chose a very conservative approach to qua'ntification of the frequency of reactor coolant to the RWST via either failure of CBS-V55 or CBS-V56. We assumed that both check valves are initially open. Furthermore, we updated our data review and incorporated all relevant data pertaining to failure of check valves to close on demand, including the recent events at San Onofre.
~This approach is conservative because ruptures of check valve discs once they have been closed are f ar less likely. In other words, although the ESD shows two paths for check valve failures, we assumed all failures to be the path in which the check. valve is initially open to maximize the estimated failure frequency.
Our updated point estimate for a check valve to fai.1 open on demand is P(CVO) = 5.5 x 10-4 We judged that f ailure of either M0V CBS-V5 or CBS-V2 to close, given failure of the above check valves, would be dominated by the operator's reliability in this situation and assigned a value of P(OPMOV) = 0.1
~ dti conservative point estimate of all the additional scenarios that lead to plant damage state 1FPV may be obtained by (added 1FPV scenarios) = [f(VS) + f(VI)].[P(S) + P(EPMP) + P(SHS)
+ 2.P(CV0).P(0PMOV)]
where the factor of 2 in the last term accounts for both RHR lines to the RWST.
From Reference 1, f(VS) = 3.3 x 10-6 reactor year.
/
f(VI) = 4.5 x 10-6 reactor year.
/
Numerical substitution yields l
f(added 1FPV scenarios) = 3 x 10-9 reactor year.
/
i 7
n j,.
~
.~
uency of Thus, our conservative point estimate yields an increase in the freg/ reactor plant damage state IFPV of only about 10% of the frequency (3 x 10-year) found in the reference.
The frequency of a LOCA inside the containment initiated by a V-sequence cannot exceed the combined frequencies of VS and VI or f(LOCA inside containment due to V-sequence) < 7.8 x 10-6/ reactor year which is insignificant when compared to the medium and small LOCA frequency i
used in the SSPSA. The consequences of the V-sequence-initiated LOCAs are judged to be bounded by those in the SSPSA.
The additional ATWS frequency caused by failure to scram following a V-sequence may be estimated from f( ATWS due to V-sequence) = [f(VS) + f(VI)].P(RT).
~ 8 x 10-10/ reactor year.
which is also insignificant when compared to the ATWS frequency used in the SSPSA. Furthermore, the consequences of the V-sequence ATWS scenario would be bounded by those in the SSPSA for ATWS.
I hope this information is useful.
I remain available for further discussion or consultation at your discretion.
t
- l 9
8:
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INSIDE CONTAINMENT
. CONTAINMENT RECIRCULATON l
SUMP b' _ _ _
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FPV ASSUMED FPVASSUMED lI______________.___'_____________._________._.________.3;',*,,{*.
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SEAL LEAK D'.AGNOSES TERMINATES -
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- VALVELEAKAGE 70 Q1 02 4
END STATE:
7D
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CBS RHR, AND RELIEF VALVES END STATE:
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END STATE:
- ASSUMES THAT STEPS 24 AND 30 OF EMERGENCY PROCEDURE E-O ARE 1 FPV COMBINED INTOTHE SAME STEP.
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i FIGURE 3 (Sheet 4 of 6) l i
s-i 8
Q UMTG1 QTJMTUI TERMINATES OvCJCCES I
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RELEF RWSTMAKEUP
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VALW S EU N ED am LOC ENDSTATE:
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END STATE:
END STATE:
1 FPV 7FPV f
i FIGURE 3 (Sheet 5 of 6) t 9
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.=
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M TOR SAFELY OS. RHR, AND REUEF ESTABUSHES END STATE CBS PUMPS RHR PUMPS lt4JECTION SAFETY HJECTION VALVES RWST MAXEUP DLOC N
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END STATE:
7D FIGURE 3(Sheet 6of6)
ENCLOSURE NO. 5 TO SBN-1207 RHR PIPING FAILURE PROBABILITY f
RHR PIPING FAILURE PROBABILITY t
The calculation of he RHR piping f.tilure probability of 6 x 10-3 when the RHR is pressurized to 2250 psia was based on the conserv-ative assumptions of a lognormal f ailure distribution and the probability of failure at yield strength of the material to be 0.01, and the probability of f ailure at the ultimate strength of the material to be 0.99.
The American Society for Metals Handbook i
lists the yeild strength and ultimate strength for 304 stainless steel (the RHR piping material) as 35 KSI and 80 KSI, respectively.
Mark's Handbook.for Mechanical Engineers also lists the yield strength at 35 KSI, but the ultimate strength at 85 KSI.
The ASME Boiler and Pressure Vessel Code Section III lists the yield strength' and ultimate strength of 304 S/S as 30 KSI and 75 KSI, respectively. -
These values are conservatively low as the ASME code values are the values for 95% confidence level, with the average values being 25%
greater.
Hence, the average ASME values for yield strength and ultimate strength w6uld be 37.5 KSI and 93.8 KSI, respectively.
{
Pickard, Lowe and Garrick, Inc. (PLG), used a conservative yield strength value of 35 KSI and an altimate strength value of 80 KSI in its calculation of the failure probability, and further conserv-atively assigned a 1% failure probability to the yield strength.
There is no evidence that any failures of the ASME test specimens i
ever occurred at the yield strength, l
To account for undetected design errors, material defects, and in-j spection oversights, a conservative f ailure probability of 10-3 was assumed.
Consequently, because of the use of conservative value of yield strength and a high confidence that failure will not occur at yield, the area of the f ragility curve around 2250 psia is. strongly believed to be conservative.
i i
h i
ENCLOSURE NO. 6 TO SBN-1207 RHR PRESSURE BOUNDARY FAILURE MODES
RHR PRESSURE BOUNDARY FAILURE MODES A. wide spectrum of f ailure modes of the RHR system pressure boundary was considered in the enhanced interf acing system LOCA analysis of the RMEPS submittal.
In order of decreasing probability, the following failure modes were considered in the event of RHR system pressurization to pressures approaching 2250 psia.
Probability at 2250 psia o RHR Pump Seal Leakage
- little or no leakage
.01
- small seal leak (.0 - 0.09 in 2)
.08
- intermediate leak (.09 - 1.05 in 2)
.4
- large seal leak (1.05 - 2.6 in 2)
.5
- mechanical seal assembly failure
.01 o RHR Piping Rupture
~
.006 o RHR Heat Exchanger Failure
<<.006 0
While the exact response of the RHR pressure boundary to pressurization to 2250 psia is highly uncertain, the above characterization of the probabilities and f ailure modes is conservative and reflects this un-certainty. Moreover, the RMEPS results are not sensitive at all to a full range of alternative assumptions chosen to market the sensitivity to this uncertainty.
It is clear that the " weakest link" of the RHR pressure boundary when pressurized to 2250 psia is the RHR pump seals.
It is the only portion of the boundary not made of steel.
The materials that compose the seal l
include ethylene propylene elastomer o-ring material.
While the pump j
seals are initially shop tested to 1200 psia, at 2250 psia, it is expected to leak, although the leak rate is highly uncertain.
A' fact that supports this assessment is that during 8 RHR pressurization events on BWRs, the l
piping and heat exchangers remained intact but in several of these instances, RHR pump seal leakage occurred.
The uncertainty distribution assigned to the leak size in RMEPS includes the f ull spectrum from no leakage whatsoever to the maximum upper bound leak area that would result if the seal is completely blown out (2.6 in 2 for blowout of both l
pump seals). This range bounds the magnitude of possible leak sizes i
caused by failure of the primary seal. A larger leak would require.
failure of the mechanical seal assembly.
As noted on page 3-31 of RMEPS, such a failure is not credible because the stresses in the studs holding the mechanical seal assembly in place are less than half of the yield stress of that material during a pressurization event.
Nonetheless, a conservative assumption was made that the mechanical seal assembly would f all with a probability of.01 despite the f act we are quite confident that the assembly would remain intact.
We remain uncertain, however, as to whether the seal leakage will be anywhere in the range of zero to 2.6 in 2 for both pumps.
4 F
a
i RHR PRESSURE BOUNDARY FAILURE MODES (Continued)
To evaluate the sensitivity of our uncertainty about RHR pump seal leak to the RMEPS results we performed 2 sensitivity analyses on the interf acing system LOCA event tree quantifications.
Case A - assume no potential for pump seal leakage.
Case B - assume the pump seal leaks at the maximum leak rate with a probability of 1.
The results of these sensitivity cases is. shown in Table 1.
As can be seen, the results are generally insensitive to the assumptions about seal leak size from 0 to 2.6 in 2 there is only a very slight shift in plant damage state frequencies associated with this uncertainty.
Hence, it really doesn't matter whether the seal completely blows out or remains fully intact.
I J TABLE 3-14 RMEPS CASE A CASE B l
EVENT TREE EVENT TREE SI = 0.0 L1 = 1 0, L2 = 1.0 END STATE
.L3 = 0.0 i
VI l VS VI l VS VI l VS i
LOCA 4(-6) 3(-6) 5(-6) 3(-6) 4(-6) 3(-6)
DLOC 4(-7) 0 0
0 4(-7) 0 f
DILOC 3(-9) 3(-7) 0 0
3(-9) 3(-7) 8C 7(-10) 0 0
0 5(-10) 0 4
7D 5(-9) 0 0
0 2(-9) 0 7FPV 3(-9) 6(-9) 0 0
2(-9)
.3(-9)
IFPV 6(-10) 3(-8) 0 0
6(-10) 3(-8) i IFV 3(-9) 2(-9) 3(-9) 2(-9) 3(-9) 2(-9)
J Our assessment that the RHR pump seals represent the weak link to RRR
~
pressure boundary integrity during over pressurization is supported by our high confidence that the RHR piping, valve bodies, mechanical seal j
assembly and studs, and RHR heat exchanger would not fail at 2250 psia.
At this maximum value of pressurization the stresses in the limiting RHR piping are less than, but approaching yield stresses,.and the heat exchanger tubes and mechanical seal assembly studs are at small frac-i tions of their respective yield stresses.
l Our assessment that the RHR heat exchangers is probably the stronges t link in the pressure boundary is supported by calculations that show j
that even if the tubes are assumed,a be reduced by corrosion or erosion to 50% of their initial design thickness, stresses in the tubes are still less than the material yield stress.
On the one hand such losses are not anticipated, especially considering the tight water 4
chemistry controls on both sides of the heat exchanger (RHR water on one side, PCC water on the other, both are closed cycle and " clean" water systems).
On the other hand, the achievement of yield stresses i
does not in itself produce f ailure.
Finally, even if the heat ex-changer tubes leak, the impact on source terms and consequences is not believed to be significant.
J l
4 4
0
ENCLOSURE NO. 7 TO SBN-1207 RHR Hx DEGRADATION I
RHR Hx DEGRADATION The referenced study contains event trees relating to the V-sequence.
These trees ask questions about the integrity of piping and heat exchange r tubes following a pressurization of the RHR syst em.
Although the fragility of piping is discussed in the document, that of heat exchanger tubes is not.
The following is a brief discussion.
The Seabrook residual heat exchangers are vertical shell and with tube units designed to TEMA Class R requirements.
The shell side of the heat exchanger is designed to ASME Section III, Class 3 requirements with a design pressure of 150 psig and the tube side is designed to ASME Section III, Class 2 requirements with a design pressure of 600 psig.
The shell side is carbon steel, the tube side is stainless steel, the tube material is stainless steel with.049 inch wall thickness. A.005 inch corrosion allowance is. applied to the tube wall thickness.
Borated reactor coolant that meets reactor coolant chemistry specifications circ-ulates on the tube side and demineralized water containing a potassium chromate corrosion inhibitor circulates on the shell side.
The water chemistry on both the shell and the tube side of the heat l
exchanger is periodically sampled as part of the plant chemistry sur-veillance program. The tube side operating pressure is higher than that of the shell side.
Any through wall leakage would leak into the com-ponent cooling system (with closed loop cooling system, cooled by the Service Water System) and would be detected by a radiation monitor in the component cooling system and water chemistry samples.
Within the Seabrook residual heat removal heat exchanger design basis including normal and expected transient operating conditions, no RHR heat exchanger degradation is likely to occur beyond that con-sidered in the design basis.
l Under the maximum internal pressurization during a V-sequence (i.e.,
2,235 psig), the maximum hoop stress inside a tube would be about 16,000 psi, which is less than one-half of yield.
Since test specimens are.not known to fail at the yield stress, failure of heat exchanger tubes at the beginning of plant life would be extraordinarily unlikely.
We can, f urthermore, estimate the ef fect of plant operation by estimating the amount of tube thinning that must occur to increase the hoop stress to the yield stress.
At yield, the tubes would not fail but would begin to define (e.g., bubble) inelastically.
Any inelastic deformation would relieve the stress.
The following equation may be used to estimate the amount of tube thinning that mu.1t occur to yield the tubes because of internal pressure.
2b2 Shoop = P b'-a' G
+
..a
.I*
k-P gs 2 RHR Hx DEGRADATION (Continued) where Shoop = hoop stress.
P
= internal pressure.
b
= outer tube radius.
f a
= inner tube radius.
We conservatively assumed all tube thinning occurred on the inside of the tube. This maximizes the hoop stress.
We found that the heat exchanger tubes, which are initially about 0.05 inches thick, must thin by 53% to about 0.0235 inches thick for the hoop stress to approach 35,000 psi.
If the tubes exhibit a yield strength closer to the upper bound found in the data (e.g., 40,000 psi),
thinning must then be at least 58% to approach yield.
9 V
1 i
. ~.
ENCLOSURE NO. 8 TO SBN-1207 CAP / COP SYSTEM CONTAINMENT ISOLATION VALVES Note:
The following drawings referenced herein will be made available for Staff's use out of our Bethesda i
l Licensing Office.
o Posi-Seal Drawing Number 1300 Rev. H, 36" Class 150-150 WOG Valve Assembly & Matrix o
Posi-Seal Drawing Number 15484, Rev. I, 8" Class 150 - Wafer Valve Assembly & Matrix (Sheets 1 & 2) i T
I i
f I
s a
h i
4 4
4
-l j
1 i
I
e
\\
CAP / COP SYSTEM CONTAINMENT ISOLATION VALVES o
POSITION INDICATION (Ref:
FSAR 6.2.4.2.g)
Positions are displayed f r the main control room at both the control switch location and as part of the post-accident monitoring sys tem.
o PRESSURE INTEGRITY & TEMPERATURE
( P.ef :
SPEC 248-45) (Posi-Seal Butterfly Valves)
C0P 8" Class 150 wafer v1v., Matrix #26062 SR-60 f ail shut Design Press: 60 psi Design Temp:
300*F FP-91012 CAP 36" Class 150 - 150 wog viv, Matrix #45162 SR-60 Design Press: 60 psi Design Temp: 300*F Normal Press: 0.5 psig Normal Temp:
104*/120*F FP-90599 o
' LIMITATIONS (Ref:
Technical Specification 3.6.1.7)
CAP closed & locked during modes 1, 2, 3, & 4 00P closed (except during purge, modes 3 & 4), cannot be opened (Modes 1 & 2)
Admin. Controls Independent Verif. Lis t, OP 10.2, Program Manual SSMN o
SEAL HATERIAL (Ref:
FP-91012 & 90599)
Tef zel seal ring with EPR back-up o-ring o
LEAKAGE (Ref:
1-PT ( I) - 3 7. 2 Rev. 1)
No individual penetration leak rate shall exceed 46.6 SCFH PEN X-18 00P-V-l&2 (combined)
.5 SCFH 2/26/86 PEN X-16 00P-V-364 (combined) 1.2 SCFH 2/26/86 PEN X-HVAC-1 CAP-V-l&2 (combined) 1.9 SCFH 2/07/86 PEN X-HVAC-2 CAP-V-3&4 (combined) 6.0 SCFH 2/03/86 o
ACCIDENT / POST-ACCIDENT CONDITIONS (Ref:
SPEC 248-45)
, <. /.
7
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~
l i
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..fTACIIMENT 9.1 1-PT(I)-37.2 REV.
l' Uk Page 31 of 36 J.
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REMARKS / REPAIRS i
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%,,// Aseony GoA iNm AtS /o AT E INmAL5 / DAT E yz/ga.<i. a:.n As rOUM PC5 MON Of VALVES.
/_/C. /. De5 CONNECT ALL TEST EQULPMENT,CLOSE/ CAP V5.
TE (SCFH) ydf r/.1/rc. a ativRN vatvE uNEup To As roUND P05til0N.
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s D-SB 1 & 2 Amendment 56
]
FSAR November 1985 k
b E* Y* A g.
Actuation and Control Equipment l
Containment isolation valves are provided with actuation and control equipment appropriate to the valve type. For example, globe and diaphragm valves are generally fitted with air diaphragm l
l operators which will fail in the safe position on loss of 54 operating air. Cate valves are generally fitted with motor l
operators and are powered from emergency buses. On loss of offsite power, the power source is automatically switched to the i
diesel generators which feed the emergency buses. Motor-operated valves fail in the as-is position. No manual operation is required for immediate isolation.
If actuating power is lost, the automatic air-operated isolation valves assume the position (open or closed) that provides greater safety. Motor-operated isolation valves fail "as is".
Manual control switches provide a secondary mode of actuation for the j
automatic isolation valves. The positions of the automatic isolation valves and remote manual valves are displayed in the main control room at both their control switch location and as 6
part of the Post Accident Monitoring System.
j
(,
h.
Seismic Design Protection for containment isolation systems and components
)
against loss of function due to seismic event forces is provided.
l Containment isolation valves and their operators are designated as 56 i
seismic Category I.
Containment isolation provisions are capable of maintaining the isolation function during and after the Safe Shutdown Earthquake (SSE). The valves are capable of being-realigned after the Design Basis Event-and to withstand seismic.
r aftershocks following the SSE.
The containment isolation valves, their operators, and supports are designed to assure that they are capable of withstanding the Safe Shutdown Earthquake as recoassended by Regulatory Guide 1.29.
To assure their adequacy in this respect:
1.
Valves are located in a manner to reduce'their accelerations.
Valves suspended on piping spans are designed for the loads to which the span would be subjected. Valves are mounted in the position recommended by the manufacturer.
2.
Valve yokes are designed for adequacy and strengthened as required for the response of the valve operator to seismic l
loads.
1
! k.I i'
6.2-69
f.-
suc zue-gr g
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4.:
)
?!
- -ll 50 to 123 Y a) Ten:perature 0.5 psig b) Pressure
..s.30-902 c) Relative Hu:nidity Air d) Atmosphere 15 Mill-Rads /Hr e) Gamma radiiation over design life
'50*F f) Test Temperature 59 psig g) Test Pressure During accident and post accident conditions which may be 3.1.1.3.2 encountered at any time during the design life of the plant, the te=perature and pressure inside the reactor containment will be as shown in Figures 3.1.1.3.4 and 3.1.1.3.5 f
The time distribution gamma dose following a respe ctively.
loss-of-coolant-accident is conservatively established as shown below:
Bera Dose Gamna Dose Total Dose a) Elaesed Time
~.23 (+0 7 )
4 3.43 (+07) 7.9 8 (+06)
First 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 4.75 (+07) 1.07 (+07) 5.82 (+07)
Next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 5.74 (+07) 1.24 (+07) 6.98 (+07) 6.56 (+07) 1.36 (+07) 7.92 (+07)
First day 1.36 (408) 2.37 (+07) 1.60 (+08)
First 5 days After 10 days 1.72 (+08) 2.81 (+07) 2.00 (+08)
After first month 2.14 (+08) 3.44 (+07) 2.48 (+08)
After one year (avg.) 2.38 (+08) 4.48 (+07) 2.83 (+08) l 1
b) The safety class equiptnant located within the j
containement must form afe paction with ouweJ/.fiE*8
~=ma phis requirements will be j
71 4
valve data sheet.
tl I
i Spec. No. 9763-006-248-45 Page No. 8 I
sg.
l g,
1 s.
i
}
I c) During an accident condition, the atmosphere within e
the containment area vill consist of saturated stea=,
air, and hydrogen (4 or less). In addition, the I
valves could simultaneously be exposed to an aqueous l
spray of:
pH Boric Acid l$I Hydorxide Solution
.d 8.0 to 10.5 (Kin. / Max. )
.. E 0.63:
Sodium Hydroxide
'S
!j 3.1.1.3.3 The nor=al operating a=bient conditions for equipment located
' lj outside the reactor containment is defined as follows:
i a) Temperature 30 to 120 F
(;
it b) Pressure Atmospheric
' jf p
- l c) Relative Hu=idity 40-100 d) Atmosphere Air I.
e)
Ca==a Radiation 2.5. Milli-Rads /Br over design life 3
a f
- I I
- l. i i.
- l' f.
f)i)'
i.
a
- i i
)
Spec. No. 976}-006-24S-45 Page No. 9 4
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Spec. No. 9763-006-248 45 Pas;e No. 10
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9Y CONTAINMENT SYSTEMS m
PRIMARY CONTAINMENT CONTAINMENT VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.6.1.7 Each containment purge supply and exhaust isolation valve shall be OPERABLE and:
a.
Each 36-inch containment shutdown purge supply and exhaust isolation valve shall be closed and locked closed, and b.
The 8-inch containment purge supply and exhaust isolation valve (s) shall be sealed closed except when open for purge system operation for pressure control; for ALARA, respirable, and air quality consider-ations to facilitate personnel entry; and for surveillance tests that. require the valve (s) to be open.
APPLICABILITY:
MODES 1,* 2,* 3, and 4.
ACTION:
a.
With a 36-inch' containment purge supply or exhaust isolation valve open or not locked closed, close and lock close that valve or isolate the penetration (s) within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, otherwise be in at least HOT SHUT 00WN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLO SHt1TDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
b.
With one or more of the 8-inch containment purge supply or exhaust isolation valves open for reasons other than given in Specifica-tion 3.6.1.7.6 above, close the open 8-inch valve (s) or isolate the penetration (s) within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, otherwise be in at least HOT STANDEY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
With one or more containment purge supply or exhaust isolation c.
valves having a measured leakage rate in excess of the limits of Specifications 4.6.1.7.2 or 4.6.1.7.3, restore the inoperacle valve (s) to OPERABLE status or isolate the affected penetration (s) so'that the measured leakage rate does not exceed the limits of Specifications 4.6.1.7.2 or 4.6.1.7.3 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and close the purge sucoly if the affected penetration is the exnaust penetration, otherwise be in at least HOT SHUT 00WN within the ne,xt 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in COLD SHUT 00WN within the,following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
- The 8-inch containment purge supply and exhaust isolation valves may not be opened while in MCDE 1 or MODE 2. (7 5 5 /S jafeo/e/ /b f e, a fef>rjGd/a feSirschoh dub /N 4Nll be re.nsoVed read A y e to befo re tu e. are y
SEABROOK - UNIT 1 3/4 6-12 moc[e /)
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1 CONTAINMENT SYSTEMS PRIMARY CONTAINMENT CONTAINMENT VENTILATION SYSTEM SURVEILLANCE REOUIREMENTS 4.6.1.7.1 Each 36-inch containment purge supply and exhaust isolation valve shall be verified to be locked closed at least once per 31 days.
4.6.1.7.2 At least once per 6 months on a STAGGERED TEST BASIS each locked closed 36-inch containment purge supply and exhaust isolation valve with resiliant seals shall be demonstrated OPERABLE by verifying that the measured leakage rate is less than or equal to 0.05 L, when pressurized to P,.
4.6.1.7.3 At least once per 92 days each 8-inch containment purge supply and exhaust isolation valve with resilient material seals shall be demonstrated OPERABLE by verifying that the measured leakage rate is less than or equal to 0.01 L, when p'ressurized to P,.
4.6.1 7.4 Each 8-inch containment purge supply and exhaust isolation valve shall be verified to be sealed closed or open in accordance with Specifi-cation 3.6.1.7.6 at least once per 31 days.
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- PAGE 494 DESCRIPTIONS OF VALVE TYPES - BY SPEC. NO.
JOB NO. 9763-006 DATE 09/09/81 PUBLIC SERVICE OF NEW HAMPSHIRE PCS REPORT + 12 SPEC. NO. 248-45 SEA 8R00M STATION-00C.*
VALVE DESCRIPTION VALVE TYPE SIZE OTY
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ENCLOSURE NO. 9 TO SBN-1207 IE INFORMATION NOTICE NO. 86-60
I IE Information Notice No. 86-60: Unanalyzed Post-LOCA Release Paths The subject notice refers to two (2) potential post-LOCA paths from the containment.
Both of the release paths are via the RCP(s) No. I seal leak-off return line.
As noted, a potential release path exists when automatic isolation signals are removed f rom the containment isolation valves for the RCP No. I seal leak-off common return line.
The above concern is not applicable to Seabrook Unit 1.
The containment isolation valves associated with the No. 1 leak-off common return line are CS-V167 and CS-V168.
System design is such that both valves automatically close upon receipt of a Phase A containment isolation signal.
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