ML20213E652
| ML20213E652 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 11/07/1986 |
| From: | Devincentis J PUBLIC SERVICE CO. OF NEW HAMPSHIRE |
| To: | Long S Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20213E656 | List: |
| References | |
| SBN-1227, NUDOCS 8611130243 | |
| Download: ML20213E652 (264) | |
Text
SEABROOK STATION Engineering Offica
.) (-
J -LJI-November 7, 1986 Pubuc Service c4 New Mcmpshir, SBN- 1227 T.F. B7.1.2 NEW HAMPSHIRE YANKEE DIVISIO:7 United States Nuclear Regulatory Commission Washington, DC 20555 Attention: Mr. Steven M. Long , Project Manager PWR Project Directorate No. 5 Division of PWR Licensing - A
References:
(a) Facility Operating License NPF-56, Construction Permit CPPR-136, Docket Nos. 50-443 and 50/444 (b) USNRC Letter, dated October 8, 1986, " Request for Additional Information for Seabrook Station, Units 1 and 2, Emergency Planning Sensitivity Study", S. M. Long to R. J. Harrison (c) USNRC Letter, dated October 23, 1986, "Reques t for Additional Information for Seabrook Station, Units 1 and 2, Emergency Planning Sensitivity Study", S. M. Long to R. J. Harrison (d) PSNH Letter (SBN-1225), dated October 31, 1986,
" Response to Request for Additional Information (RAIs)", J. DeVincentis to S. M. Long Subj ect : Response to Request for Additional Information (RAIs)
Dear Sir:
Enclosed herewith are additional responses to the Requests for Additional Information forwarded in References (b) and (c). Previous responses were submitted in Reference (d). Attachment A identifies responses that are in-cluded in this transmittal. The responses are provided in Attachment B.
An additional submittal addressing the remainder of the RAIs will be forthcoming in the near future.
Very truly yours, 8611130243 861107 8 PDR ADOCK 05000443 F PDR John DeVincentis Lirector of Engineering Attachment cc: Atomic Safety and Licensing Board Service List Director, Of fice of Inspection and Enforcement gg United States Nuclear Regulatory Commission N Washington, DC 20555 i
Seabrook Station Construction Field Office . P.O. Box 700 . Seabrook, NH O3874
,Dien3 Curran, Esquire Peter J. Mathsws , Mayo r Harm 2n & Weiss City Enll 2001 S. Street, N.W. Newburyport, MA 01950 Suite 430 U:shington, D.C. 20009 Judith H.,Mizner Silvergat e, Gertner, Bake r ,
Shorwin E. Turk, Esquire Fine, Good & Mizner office of the Executive Legal Director 88 3 road s't.
j i U. S. Nuclear Regulatrry Commission Boston, MA 02110 I
Tenth Floor Unshington, DC 20555 Calvin l A. Canney City Manager Rsbert A. Back'us, Esquire City Hall 116 Lowell Street 126iDaniel Street P. O. Box 516 Portsmouth, NH 03801 M:nchester, NH 03105 St,aphen E. Merrill, Esquire Philip Ahrens, Esquire Accorney General Arciscant Attorney General George Dana Bisbee, Esquire Dspartment of the Attorney General Assis tant Attorney General Sectehouse Station #6 25 Capitol Street Augusta, ME 04333 Concord, NH 03301-6397 Mrs. Sandra Gavutis Mr. J. P.'Nadeau Chairman, Board of Selectmen Selectmen's Office RFD 1 - Box 1154 10 Central Road Konsington, NH 03827 s Rye, NH 03870 Carol S. Sneider, Esquire Mr. Angie Machiros Assistant Attorney General Chairman of the Board of Selectmen Dspartment of the Attorney General Town of Newbury On3 Ashburton Place, 19th Floor Newbury, MA 01950
~
Boston, MA 02108 .
Mr. William S. Lord S:nator Gordon J. Humphrey Board of Selectmen U. S. Senate , Town Hall - Friend Street Wtshington, DC 20510 Ames bury,10L 01913
( ATTN: Tom Burack)
- Richard A. Hampe , Esquire Senator Gordon J. Humphrey Htmpe and McNicholas 1 F,illsbury Street 35 Pleasant S t re'e t Concord, NH 03301 Concord, NH 03301 ( ATTN: ' Herb Boynton)
Thomas F. Powers, III H. Joseph Flynn, Esquire Town Manager Office of General Counsel Town of Exeter Federal Emergency Management Agency 10 Front Street 500 C Street, SW Cxeter, 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 Maplowood Avenue Gary W. Holmes, Esquire P. O. Box 360 Holmes & Ells Portsmouth, NH 03801 47 Winnacun' net Road Hampton, NH 03842 Mr. Ed Thomas Robert Carigg FEMA Region I Town Office 442 John W. McCormack PO & Courthouse Atlantic Avenue ,
Boston, MA 02109 North Hampton, NH 03862 {
J
a ATTACHMENT A Responses to these RAIs were forwarded by SBN-1225, Reference (d):
1 12 24 44 61 2 13 25 45 62 3 14 26 46 63 4 15 28 49 64 5 16 33 50 67 6 17 34 51 68 7 19 35 53 69 8 20 40 55 70 9 21 41 57 71 10 22 42 59 72 11 23 43 60 73 Responses are included in this transmittal for the following RAIs:
18 48 75 27 52 30 54 32 56 36 58 37 65 39 66 47 74 Responses to the following RAIs will be forthcoming in an additional submittal:
29 31 38
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ATTACHMENT B i
RESPONSES TO REQUESTS FOR INFORY.ATION l
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RAI 18
Provide the basis for concluding that the sight glasses in the hatches will not fail under high containment temperature and pressure conditions.
ANSWER 18 The sight glass in the personnel hatch was tested by its supplier, Owen Corning Co. , under the following conditions:
Pressure = 150 psig Temperature = 5500F In addition the pressure was cycled from 0 psig to 150 psig ten times at a constant temperature of 5500F.
The Owens Corning data sheet is attached.
We are currently pursuing discussion with Corning Glass to determine if any testing has been done above these values.
Cocning indicates that a conservative allowable working stress for the
- 7740 tempered glass is 300 psi. At this working allowable, pressures in excess of 200 psi can be accomodated without glass failure.
TEST SAMPLE: TD4PERED SIGHT GLASS, CORNING PYREX NO. 7740. 7 INCHES DIAMETER BY 3/4 INCHES THICK.
W. J. WOOLLEY TESTING:
PERFORMANCE TEST: PRESSURIZED TO 150 PSIG FOR 3 MINUTES.
MECHANICAL CYCLING TEST: PRESSURIZED FROM 0 TO 150 PSIG, MAINTAINED FOR 10 SECONDS, THEN TO O PSIG, A TOTAL OF 100 CYCLES.
RADIATION EXPOSURE: 60 MEGARADS GAMMA. (NO'IE 1)
POST RADIATION PERFORMANCE TEST: PRESSURIZED TO 150 PSIG FOR 3 MINUTES TD(PERATURE : AMBIENT CORNING TECHNICAL SPECIFICATIONS:
MAXIMLH WORKING PRESSUPE: 150 PSIG MAXIMLH WORKING TEMPERATURE: 2900 C, 5540 F NOPJ4AL SERVICE TEMPERATURE: 2600 C, 5000 F ROCM TDiPERATURE M 270 0 C, 5180F HEAT SHOCK TEST:
P'FERENCES:
- 1. CORNING GLASS BULLETIN 15-20, 1971
- 2. CORNING GLASS CUSTOMER PRODUCT SPECIFICATION, JANUARY 15, 1973
- 3. AMERICAN ENVIRONMENTS COMPANY QUALIFICATION TEST REPORT FOR PERSOhHEL AIRLOCK VIEWPORT SIGHT GLASS REPORT NO. STR-62883-2 NCfrE: -
- 1. THE SIGHT GLASS CHANGED COLOR FROM THE GREEN TINTED CLEAR TO A DEEP REDISH PURPLE, DUE TO THE GAMMA RADIATION EXPOSURE. VISIBILITY THROUGH THE SIGHT GLASS WAS REDUCED.
RAI 27
Liscuss the ef fect on risk of hydrogen deflagation/ detonation in the RHR j vault.
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RESPONSE 27
, Transient hydrogen burn analyses were not performed for the RHR vaults.
During a V-sequence hydrogen is expected to be released into the RHR
! vaults via the RHR pump seals. Steam condensation in the water pool f may increase the concentration of hydrogen released at the top of the water pool to a flammable condition. If a hydrogen burn were to occur
]
above the pool in the RHR vault it would almost certainly be a continuous burn at the pool surf ace because of the continuous release through the RHR pump seal prior to vessel melt through. A postulated global hydrogen burn in the RHR vault could cause a pressure induced f ailure of the RHR vault boundary. The RHR vault is part of the enclosure building area ventilation vault boundary and the pressure capacity of this boundary was analyzed
- in Section 6 of Appendix H.1 in the SSPSA. The weak points in the RHR vault pressure boundary are readily apparent. First, the double wing fire doors at grade elevation (+20 feet) communicate into a corridor l
which communicates to the outside through another large double door.
j These doors are located approximately 50 feet above the water pool 4
surface. The second weak point is the RHR vault ventilation system j which communicates with the enclosure building area. -The weak points i in the pressure boundary for the enclosure building were identified as (1) the metal siding in the fire walls between the charging pump cubicle l
and the primary auxiliary building, (2) the enclosure building ventilation ducting in the primary auxiliary building and (3) the HEPA filters in the i enclosure building ventilation exhaust. Table 11.3-4 in the SSPSA shows j that both the ventilation duct work eld the HEPA filters are expected
- to fail at a pressure below 2 psid and the metal siding in the charging
! pump cubicle is expected to f ail at a similar low pressure. The RHR vault structure on the other hand is designed for an internal pressure t
of 3 psid which according to the analysis in SSPSA Appendix H.l. Section 6 has an expected pressure capacity of at least 6 psid. thus the expected f ailure locations due to overpressure in RHR vault are readily identified.
l The analysis of the V-sequence in the RMEPS takes credit for source term
- mitigation, by the deposition of fission products on the structure i surfaces above the pool. The failure mode involving the ventilation duct work or the HEPA filters both would increase the surf aces available for radionuclide deposition before release. The vault failure mode involving
- the two sets of fire doors at grade level would result in radionuclide releases to the environment by a release path that is essentially identical to that modeled in the RMEPS study.
In the WASH-1400 sensitivity study (PLG-0465) credit was only taken for radionuclide scrubbing in the RHR vault pool, but not for any deposition above the pool. The only mechanism which could increase the consequences of a V-sequence in the sensitivity study as a result of a hydrogen burn would be the structural failure of the vault walls at such a low level that the water pool is lost down to the RHR pump level. This would require
failure of the concrete vault walls at an elevation where the walls are 4 feet thick, twice the thickness of the walls at higher elevations. As described above the pressure capacity of the two foot thick concrete walls is at least 6 psid. Failure at these low elevations will thus be prevented by failure first of the metal siding, fire doors, ventilation ducts, or HEPA filters, and secondly by the thinner concrete walls above elevation (-)41. Loss of the water pool will thus not occur and a postulated global hydrogen burn in the RRR vault would not increase the consequences of a V-sequence.
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RAI 30 l The S7W release category isotopic distribution reflects a decontamination l f actor (DF) of 1000, for all isotopes except noble gases, because the release point is submerged in the RRR vault. WASH-1400 source term methodology credited BWR releases with a DF of 100 when they occured through a subcooled suppression pool, but set the DF to I when the pool was at saturation temperature. .
l a.
Discuss the degree of subcooling that would be expected in the RCS water that pools in the RHR vault following blowdown through the RHR system.
- b. Justify the use of a DF-1000 in light of the WASH-1400 methodology and the degree of subcooling expected in the RHR vault water.
1 RESPONSE 30 The subcooling of the water in the RHR vault is a function of the size of the break, the injection rate, and heat losses to structures in the reactor building. Considering the wide variations of accident scenarios that could be postulated, it is reasonable to assume that the bulk subcooling of the pool would be at least 1000 due to concrete heat sinks. In addition, the static head of the water in the vault also represents a significant subcool-ing, since a 10 meter deep pool saturated at the surface would have a subcooling 0
of 100C at the failure location. Due to these subcoolings (10 to 20 C), a considerable amount of deposition of the aerosol fission products would be created by the initial bubble collapse. Analyses that were performed to calculate the degree of subcooling are described below.
The S7W release category is the expected result of an interfacing systems LOCA, or V-sequence initiating event. The f ailure of the RHR isolation valves which initiates the event allows the pressurized primary water mass to blow down into the RHR pump vaults through the RHR pump seals.
The highly pressurized reactor coolant entering through the RHR break will flash upon contact with the low pressure atmosphere in the RHR pump vaults.
The steam generated by this flashing process will enter into the RHR pump vaults and the saturated water generated will fall to the floor of the RHR equipment vault and begin to cool through heat transfer to the concrete.
The transient will begin with highly pressurized water exiting through the break, then will con:inue with a short period of steam flow through the break as the rapid depressurization in the primary system causes some flashing in the remaining reactor coolant. Safety injection will be actuated by the decrease in RCS pressure and will reflood the core with water from the RWST within several minutes af ter the steam flow through the break begins. Water flow will then continue through the break driven by the pressure of the RCS which is maintained at or below the RHR relief valve setpoint pressure until vessel failure occurs.
The RCS pressure is maintained initially by the safety injection pumps until they are flooded by the rising level in the auxiliary building, and then driven by the charging pumps until the RWST is emptied. Once the RWST is emptied, the water level will drop in the core until the hot leg is once more uncovered, and the breakflow will then consist
solely of steam flow for the remainder of the transient. As the core melts and vessel failure occurs, the pressure driving the steam flow will drop off and the steam flow will also decrease.
The ' original analysis consisted of a hand calculation using the model specified above which has since been verified with a confirmatory computer run. The steam flow from the flashing at the break was conservatively _ ignored until the break is covered, since condensation of the steam on the walls and in the air would add inventory to the auxiliary building pool and would cool the water in the pool. Af ter the break is covered by the' rising water in the RHR vault, the water exiting the break was assumed to mix with the water in the pool.
When steam is exiting the break, it is assumed to bubble through the pool without much condensation because if its initial flow velocity.
Heat transfer was only assumed between the water and the surrounding concrete floor and walls, and no heat'tranfer was considered between the' water and metal equipment or the RHR vault atmosphere, which would act to subcool the water in the vault even further. These assumptions were used to verif y the results and conclusions of the original analysis. The results for the RHk vault bulk water temperature versus time for the duration of the transient are presented in graphical form in Figure I with some results in tabular form in Figure 2. Please note that this figure does not predict the return-to saturated conditions near the end of the transient run time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> noted in the Emergency Plenning Study near the top of page 4-36. This return to saturated conditions is due to the steam break flow rate dropping suf ficiently after core melt and vessel f ailure to allow the steam to condense as it flows into the RHR vault water pool. This condition was modeled in the initial hand calculations, but not in the computer results presented here.
The results presented in Figure I show that the RHR vault water pool remains subcooled throughout the period of maximum fission product release through the break (10 to 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> into the transient). Combined with the depth of the RHR water pool, the subcooled conditions of the pool should provide sufficient justification for the use of a DF of 1000 for this release category.
The decontamination f actor is determined by subcooling of the pool, the particle size, and the steam mass fraction being lost from the break location. Typical representations show a decontamination f actor as a function of these elements and the decontamination itself experiences a minimum with the lowest DF being -at a particle diamter of about .25 microns and an equilibrium steam mass fraction in the existing gas flow.
In the attached memo, this is shown to have a DF of about 3,000 with a water depth ~1n the vault of 10 meters. Considering that (1) the water 1 would have subcooling both due to structural heat sinks and the static head of the water pool and (2) that the typical aerosol distribution 7
would have particles in the range of a f ew microns, the average DF of l 1000 used in the study is a conservative assessment for the complex j processes involved in aerosol deposition. For typical V-sequence scenarios l
with flow down a long pipe, the aerosol particles would experience deposition on the pipe wall and re-entrainment by the high velocity gas l
stream before being discharged into the water pool. In this case, the l particles which are entrained by the gas stream would be a few tens of i
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f microns to 100 microns in diameter. Particles4 of this size would have a very high decontamination factor, i.e. > 10 . This is an additional conservatism not accounted for in this analysis.
The above discussion of decontamination factors represents our best state of knowledge about the unique conditions associated with RHR pump vault flooded release paths. In the sensitivity study, it was determined that WASH-1400 methodology did not have an accurate source term for this scenario. The closest match was the use of DF of 100 for subcooled BWR suppression pool releases. It was also known that the IDCOR program assessed BWR suppression pool scrubbing at a DF of 1000 (IDCOR 23.1 for Grand Gulf) and subsequently, based on additional experimental data, a BWR DF of 500 to 700 was supported (IDCOR 85.2). Therefore, the IDCOR work suggests f
that WASH-1400 may have been conservative for BWR suppression pool scrubbing by a f actor of 5 to 10. In view of the greater depth of water I in the Seabrook RHR pump vault, the assumption of a DF = 1000 seemed to be appropriate for performing the sensitivity study. It is clear that the results of the sensitivity study are insensitive to assumptions regarding RHR decontamination f actor. If a DF of 100 had been used, the numerical results for early health risk safety goal and dose vs.
distance would not have changed appreciably. Even if a DF of I had been used, the 200 RE!! dose vs. distance would not have changed because of the low frequency assessed for S7.
FIGURE 1 ANALYSIS FOR SEABROOK V SEQUENCE SUBCOOLED WATER SCRUBBING TEMPERATURE OF WATER IN LOWER VAULT REGION 1
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r FIGURE 2 Tabular Results of Seabrook Auxiliary Building Water Analysis Time (sec) Water Temperature (OF) 192 156.3 268 180.8 613 162.1 1281 200.4 2418 205.1 6087 20 8 .5 10553 209.5 20213 208.9 29997 208.1 40033 206.9 59140 205.0 86401 203.1
Fauske & Associetes,Inc.
DATE: October 31, 1986 TO: R. E. Henry FROM:
P. G. Ellison Ifk
SUBJECT:
Minimum Expected DF in a Pool 10 m Deep with Typical V-Sequence Conditions The decontamination of a gas stream of aerosol is given by kZ DF = e ())
where Z is the pool height and k is given by 2 2h 8h T aas 8 gad k = 81 12 D V u +D 2o V i+ d 1.8 (2)
) 3 VD(3nud)
The slip correction factor s is given by 8 = 1 + Kn (2.514 + 0.8e-0.55/Kn) (3)
Equation (2) has a minimum around a particle diameter d of 0.25 um.
The minimum DF in a pool, assuming that k is not a function of height can be estimated using the minimum particle diameter, equilibrium steam fraction and neglecting entrance effects. For the typical V-sequence conditions given below:
o = 5000 kgm/m3 0 = 5 x 10-3 ,
d = .25 x 10-6 , T gas
= 350*K u = 1.7 x 10-5 kgm/m-sec K = 1.38 x 10-23 joule /*K o = 0.6 kgm/m 3 V = 0.25 m/sec 8 = 1.947 Kn = 0.353 Z = 10 m 16WO70 West 83rd Street
- Burt Ridge, Illinois 60521 * (312) 323 8750
TO: R. E. Henry October 31, 1986 and from Equation (2) '
k = 0.023 + 0.358 + 0.441
, = 0.822 or 10 DF = e0.822 0F % 3000 Thus a minimum DF of the order of 1000 is expected for these conditions.
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! RAI 32 l
l l In your prediction of large deformation behavior of the containment, full bond l was assumed between reinforcement and concrete between two adajacent vertical cracks; assess the effect on containment behavior including penetration capability, if no bond stress is assumed between the reinforcing steel yield point and ultimate strength of steel. Based on our discussions in the meeting, it is our understanding that you will perform this assessment assuming no bond stress.
RESPONSE 32 As was agreed in the review meeting at Brookhaven National Laboratory on October 16 and 17, 1986 we have recalculated the containment pressure versus deflection curve (Figure 4-2 in Appendix H.1 of the SSPSA) under l
the assumption of no bond stress, and we have interpreted the dif ference in the two curves in terms of changes in the containment failure pressure and in the containment failure time. The attached Figure shows both the original pressure displacement curve and the no bond strength curve.
The difference in failure pressure is seen to be 15 psi at the median-failure pressure of 196 psig, 10 psi at the 1 percent deformation strain and 7 psi at the type A containment failure pressure of 166 psig.
The table below compares these pressure changes and the resulting changes in containment failure times. It is concluded that:
- 1. The small changes in the time of containment failure will not have any noticable impact on the source terms or on the consequences for either the RMEPS study (PLG-0432) or the sensitivity study (PLG-0465).
- 2. None of the conclusions in the above referenced studies would be affected.
- 3. The ef fects of assuming no bond stress are well enveloped by the results of the sensitivity study for containment f ailure which was documented in response to RAI 20. In that analysis it was arbitrarily assumed that containment failure occurs at a deformation strain of 1 percent.
i CONTAINMENT FAILURE PRESSURE (PSIG) AND FAILURE TIME (HOURS)
FOR TWO ASSUMPTIONS ON CONCRETE REBAR BOND STRESS CASE CONTAINMENT FAILURE PRESSURE (PSIC) TIME (HOURS) 4 BOND STRESS YES NO YES NO WET, MEDIAN 196 181 54 30 WET, TYPE " A" 166 159 38 36 1
DRY, MEDIAN 172 16 2 89 76 4
DRY, TYPE "A" 157 151 62 56 1 1 I
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with and without Bond Stress i
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RAI 36
Discuss the results of recent EPRI tests to address the potential for strain concentration in the liner at crack locations.
l RESPONSE 36 SMA is presently reviewing the EPRI test reports to determine the applicability of any test results to the assumptions made in the ultimate capacity analysis. The NRC will be notified if the reviews identify any information that has a significant negative impact on the SSPSA or the RMEPS conclusions.
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i RAI 37 Demonstrate that your calculations fully account for the differences in l stress-strain behavior between the reinforcing steel and the lower plate with regards to strain compatibility.
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RESPONSE 37 As discussed in the October 16th and 17th meeting at BNL we have investigated the availability of a stress-strain curve for the liner material in the plastic region. This curve could be utilized to demonstrate that liner strain at ultimate capacity is on the flat part of the plastic region.
j We have not located a pertinent plastic stress-stain curve for normalized A516 Gr 60, however we have located extra liner plate material on site w'.iich is from the same " heat" as installed liner plate. Utilizing actual liner material, we will perform necessary testing to develop the relevant plastic stress-stain curve. This testing and final curves will be completed in December, 1986.
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1 I RAI 39
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Only selected penetrations were analyzed in the calculations; compile a list of all containment penetrations, categorize according to behavior and i
demonstrate that each penetration is adequately covered by the analyses that have been performed.
- RESPONSE 39 i
Containment piping penetrations and their qualification methods are
! summarized in the attached table.
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Penetration Closure Penetration Number Penetration Specifically Qualification Analyzed Method Type Flued Head X-1 to X-8, X-9 to X-15, X-8 Report pages H.1-44 to I.
X-63 to X-66 (18 inch, sch 100 Carbon Steel) H.1-50 Flat Pla teClosure X-25, X-26, X-27 X-26 Pages H.1-39 to H.1-44 II.
Thick Wall - 1.arge (4 inch, sch 160, stainless)
Bore Piping Flat PlattClosure X-16 thru X-24 X-23 Pages H.1-39 to H.1-43 III.
Thin Wall - Large X-28 thru X-34 (12 inch, sch 40, Carbon Steel)
Bore and Small Bore X-39, 41, 42, 50, 60, Piping 61, 67 Flat Plate Closure X-35 thru X-38 X-71 Page H.1-37, H.1-39 l IV.
Thin Wall Piping X-40, X-43, X-47, X-48, Multiple Penetration X-49, X-50, X-52, X-57, X-71 thru X-76 Fuel Transfer Tube X-62 X-62 Page H.1-50 to H.1-55 V.
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RAI 47
Address the risk from creep failure of the steam generator (S/G) tubes due to exposure to high temperatures during core melt sequences in which the reactor coolant system (RCS) remains at high pressure and the secondary sides of the S/Gs are dry. Your discussion should reflect the recent experiments and modeling ef forts that show 3-dimensional convective flows which transfer heat from the overheating core to other places within the RCS, particularly into the upper plenum and f rom the upper plenum along the hot legs into the S/Gs and through the U-tubes.
Also include the influence of pressure driven flows resulting f rom reactor coolant pump (RCP) seal LOCAs, PORV/ safety valve actuations,
" bumping" the RCPs, etc. Localized heating ef fects due to redistribution of fission products in the RCS should be included.
- a. What is the total probability of occurrence for the high RCS pressure core melt sequence with dry S/Gs?
- b. What is the estimated conditional probability that the S/G tubes will fail due to overheating before the pressure is relieved by failure of the RCS elsewhere?
- c. What is the effect of preexisting S/G tube leakage (within technical specifications) on the heating rate and temperature required for failure of the leaking tube (s).
- d. What release category would creep failures of the S/G tubes result in?
RESPONSE 47 The risk from creep f ailure of steam generator (SC, tubes is very small for the following reasons:
o The frequency of high pressure core melt with dry steam generators is very small, o Given the postulated occurance of a high pressure core melt with dry steam generators, creep rupture of the SG tubes is not a credible failure mode.
o A large number of tubes must fail to produce an early large containment bypass.
A simple sequence diagram in Figure i shows that three failures must occur for containment bypass to occur as a result of a high pressure melt seq ue nce. The three failures are (1) failure to recover water to the SG, (2) failure to depressurize primary system and (3) creep failure of SG tubes. Success of any one of these three ensures success (no SG tube creep failure).
The potential for creep failure of the steam generator tubes for core melt sequences in which the steam generators are dry was evaluated using the best tools available including the MAAP 3.0 computer code developed in the IDCOR program. As a result of the technical issue resolution ;
process between IDCOR and the NRC, this code includes models for the )
natural circulation flows between an overheated core, the upper plenum, l the hot legs and the steam generator tubes. The analytical models used I in the MAAP PWR program have been benchmarked with the recent published f experiments carried out at Westinghouse under EPRI aponsorship. In
! addition, these analyses also include the influences of PORV/ safety l valve actuations, as well as models for fission product release , trans port ,
deposition and localized heating.
Several variations of station blackout sequences were considered: all assumed f ailure of main and emergency feedwater such that the secondary sides of the steam generators remained dry. The sequences studied include: (1) no RCP seal LOCA or manual PORV actuations (2) a 50 gpm per pump seal LOCA with no manual RCS depressurization (3) manual RCS depressurization when the core outlet temperature equals 12000F, and (4) several uncertainty analyses on key physical parameters in the system models. These analyses indicated that the steam generator tubes would not be subject to a sufficient temperature increase to result in creep rupture before failure of the reactor coolant system. In particular, since the secondary side temperature is limited to about 7000F due to thermal inertia of the SG shell and heat losses to the containment, the SG tube temperatures only achieve temperatures of about 11000F. This is far less than that required for creep rupture. Based on information provided in Appendix B of the attached draf t report, creep rupture of the SG tubes would not be expected. These conclusions are valid for all of the sequence variations considered as well as all of the uncertainties considered in the analysis which include variations in the core eutectic formation and slumping behavior, flows through the steam generator tubes, etc.
During station blackout sequences, the Seabrook operating procedures require that the operators monitor the functional restoration guidelines. These guidelines would first lead the operator to consider restoration of reactor coolant system inventory control which would be met with the discovery that no electric power is available for this postulated sequence. Subsequently, the restoration guidelines would lead the operator to consider procedures for restoration of a RCS heat sink.
The procedure for loss of heat sink calls for manual depressurization of the primary system using the PORVs. While this procedure is not specifically implemented for station blackouts and is only monitored, if new evidence subsequently turns up to support the view that creep rupture is a credible failure mode, the station blackout procedures could be changed to require some appropriate method of RCS depressurization.
The pref erred method would be to blow down the secondary side of the steam generators using the SG PORVs. Such action would permit introduction of steam generator cooling via the firewater pumps (diesel driven).
An alternative method would be to depressurize the primary system using the pressurizer PORVs. With such an action, the pressure dif ference imposed on the entire primary system pressure boundary is dramatically reduced, thereby eliminating any potential for creep rupture f ailures.
u .
The action of " bumping" the RCPs cannot be accomplished without the return of of f site power. If one postulates of f site power being returned, the operator would first restore ECCS systems. This action alone will recover core cooling and make bumping the RCPs unnessary.
In summary, extensive MAAP 3.0 analyses, which includes models for the pro-cesses in question, were carried out in response to this question. These analyses show that the steam generator tubes would not be threatened by the creep rupture mechanism due to either the natural circulation flows between the core, upper plenum, hot legs and steam generators or the pressure driven flows resulting from RCP seal LOCAs or manual RCS depressurization. Also, the calculations show that localized heating effects due to fission product deposition are not sufficient to increase the temperature of the steam generator tubes to a level where creep rupture would be anticipated. Consequently, if any such mechanism were anticipated, it would occur elsewhere in the primary system long before the steam generator tubes would be threatened.
1
! The mean annual frequency of early high pressure core melt sequences of interest is approximately 4.5 x 10-5 events per reactor-year.
PDS Mean Annual Reference Frequencies 3D 1.5 x 10-5 RMEPS Page 3-58 3FP 8.9 x 10-6 RMEPS Page 3-60 4A 1.4 x 10-5 RMEPS Page 3-61 4C 1.7 x 10-7 SSPSA Page 13.1-28 4D 2.8 x 10-6 SSPS A Page 13.1-28 4E 2.2 x 10-II SSPSA Page 13.1-28 4FP 1.2 x 10-7 SSPS A Page 13.1-28 8A 3.9 x 10-6 RMEPS Page 3-66; sequences 8A-27 through 30 and 8A-34 4.5 x 10-5 through 37 There are many specific accident sequences that comprise the above plant damage states. These sequences broadly include transient and loss of of f site power sequences with f ailure of main and emergency feedwater and failure or inability to feed and bleed and transient without scram sequences. The above results for PDS 8A consist of 8 sequences involving station blackout and emergency feedwater failure in which efforts to recover containment heat removal are successful. In the containment event tree analysis, all the "A" states are assigned a high chance of no containment f ailure, the "C" and "D" states a high chance of long term overpressurization, the "FP" states a high chance of small bypass and the "E" states a high chance of large bypass. Hence, in terms of relative consequences, the "A, C and D" states would experience the greatest increase in consequences if a SG tube failure were assumed to result f rom creep f ailure during high pressure melt sequences.
Early high pr:scura s2quencsa arc of interest bIctu:s s:quences with i
energency feedwater f ailure (i.e. dry S/Gs) are modeled as early melts
! (i.e. plant damage states 3 and 4). The one exception applies to 8 l sequences in RMEPS where containment heat removal recovery was considered.
These particular recovered sequences were assigned to PDS 8A. PDS 3F and 4F are not of interest because a large containment bypass (release category S6) already exists. PDS 3FP and 4FP are not of interest because a release path (S2) already exists.
The above frequency of early high pressure core melt is conservative for three reasons.
- 1. There are still potential recovery actions to ensure wet S/Gs not yet considered. For example, the dominant sequence (3D-1 on RMEPS page 3-58) involves a loss of main feedwater transient and failure of the solid state protection system (mean annual f requency = 8.3 x 10-6). Solid state protection system f ailure is assumed to result in no auto initiation of safety equipment such as emergency feedwater. Emergency feedwater and other safety equipment can be started manually from the Control Room but these actions were conservatively neglected in the PSA.
- 2. There are early high pressure melt sequences with wet S/Gs in some of these plant states (i.e. emergency feedwater available and operating). For example, sequence 3D-4, a non-recovered station blackout and the operators don't depressurize S/Gs. Other such sequences include 4A-2 and 3, and 3A-29. (Exclusion of these sequences would not signif-icantly reduce the total estimated above).
- 3. No credit is taken for operator actions to manually depressurize which reduces pressure (i.e., not high pressure core melt). Also, there is some chance the PORVs may fail to close.
The analyses indicate that the steam generator tube temperatures and the hot legs are both well below the levels where creep rupture would occur. In addition, it is likely that the operator would be instructed to manually depressurize the RCS when the core outlet temperature exceeded 12000F.
This action would decrease the primary system pressure such that the stresses on the steam generator tubes and the hot legs would both be reduced to levels less than that observed under normal operation.
Consequently, the estimated conditional f requency that the plant conditions for the hypothesized creep rupture of steam generator tubes would occur is dependent upon the likelihood that the operator would fail to manually depressurize the primary system. With modified procedures and adequate training, the fre uency of operator f ailure to depressurize could be reduced below 10~ to 10-3 per demand. This would lower the frequency of creep failure potential conditions to the 10-7 to 10-8 range.
Given the occurrence of these conditions, the probability of creep rupture is a matter of our state of knowledge about the laws of physics that govern heat transfer to the SG tubes as opposed to a statement about the relative frequency of a random process (which we make about the likelihood of achieving the necessary plant conditions in terms of
" frequency"). Due to cooling of the steam generator tubes by the secondary side steam, the SG tube temperatures are well below (approximately 200-300 F) that required for creep rupture. Due to the very strong dependence of creep rupture on temperature, this consideration means that creep rupture would not occur, even if high pressures are maintained.
Despite the uncertainties in our models and data, we are quite confident that errors as large as several hundred degress on the low side of the correct valve are not credible. To express our confidence in the models and data, we assign a 99% chance that failure of SG tubes will not occur before reactor vessel melt through or piping nozzle failure.
The ef fect of a pre-existing steam generator tube leakage (within technical specification - 1 gpm per steam generator), would be a small flow rate compared to that created by the 50-gallon per minute flow assumed for the pump seal LOCA case. Since the major influence of an additional flow would be to provide higher temperatures in the steam generators, this is bracketed by the results of the pump seal LOCA case which showed no major influence on the steam generator tube temperatures.
Given the assessment of the primary system response, potential for creep rupture of the steam generator tubes is extremely unlikely. Consequently, no specific calculations were carried out for the actual release categories since it is not risk significant. If specific questions were to arise regarding such releases, these would be analyzed with the HAAP 3.0 PWR code.
If one postulates creep rupture f ailure of steam generator tubes, the pressure inside the previouly dried out and isolated stream generator secondary side would increase until the steam generator PORVs setpoint is reached at which time these valves would lif t and modulate until reator vessel melt through occurs and the RCS depressurizes into the containment. There are no existing release categories / source terms analyzed for Seabrook or other PWR plants that adequately represent this scenario. During the periods of SG PORV opening, there would be a high leak rate bypass condition directly from the RCS to outside the containment. However, after vessel melt through, the leak rate out this path would be low corresponding to any low pressure leakage through the reclosed PORV. This leak path could be enhanced if the SG safety valves also lift and fail to reseat properly; however, it is believed unlikely that the safety valve setpoint would be reached. If a source term were developed for this scenario, it would probably resemble S2 with the addition of an early puff to cover the relief valve opened period of the release. Depending en the sequence, there may or may not be a long term overpressurization component to the source term.
Figure 1 CREEP FRILURE OF STERM GENERATOR TUBES From SSPSR High Pressure Core Melt tulth Dry Steam Generators 1 P Recover Hlater to YES Steam Generator i
NO If Manual RCS YES m /G Depressurization Success d k NO 1 P l Steam Generator YES Tube Integrity Maintained NO II Containment Bypass
RAI 54
The authors concluded on page 3-9 that presence of water in the reactor cavity will decrease (significantly?) the revaporization of fission products f rom RCS c:.d perhaps RHR surf aces. We anticipate that a significant quantity of heat producing radioisotopes will remain in the wreckage of the reactor vessel, and this may be ef f ective in heating what-ever gases or vapor are flowing toward the break. Has this been investi-gated?
RESPONSE 54 Water present in the reactor cavity will essentially eliminate heating of the reactor vessel by the debris. Furthermore, volatile fission products trapped in the primary system and RHR piping will be cooled by the flow of steam f rom the cavity pool to the RRR pump vaults. Calculations indicate that this flow could cool somewhat more than half of the volatile inventory, were it trapped in the RHR line alone. Larger inventories can be cooled as the decay heat drops in long duration sequences or if credit was taken for direct heat losses from the primary system to the containment. Somewhat smaller inventories could be cooled if the fission products were concentrated 1 at one location, although natural processes would tend to distribute the fission products.
7 1
RAI 56
There have bepn a number of indicatida (prior to and including page 3-11) that contai nment spray may be actuated due to RHR relief valve release into containment. What is the justification for this conclusion? Include the ef fect of containment heat sinks and containment cooler operation in the respo nse. '
RESPONSE 56 In the early stage of the V-sequence analysis, it was recognized that there could be some potential for containmene spray (CBS) actuation since the RHR relief valves would be discharging to containment (through the pressurizer relief tank); in the interest of' completeness, this possibility was included in the modeling.- As discussed o;t page 3-11 (top), automatic CBS actuation occurs on a P-sigaal which is generated by a high containment pressure of 18 psig ( 33 psia) ." The V-sequence behavior was analyzed using the MAAP computer program as discussed at the' bottom of page 3-11 and throughout Section 4. MnNJ predicted containment pressure is plotted in Figure 4-13.
MAAP does account for such affects; as heat sinks and containment air coolers, though for this particular case the coolees were assumed to be not operating.
In actuality, 5 of the 6 coolers wou},d typically be running and delay CBS actuation signal even further.
. s ,
It should be noted that CBS pum)[ operation could have a negative af fect on the V-sequence since it directs Ref ueling Water Storage Tank (RWST) inventory to the containment. That is, away ft'om the RCS where it could contribute to core cooling and away f rom the vault wiiere it could contribute to fission product sc rubbing . Thus neglecting the containment coolers is conservative.
The discussion at the bottom of page 3-11 provides a somewhat simplified event timing ' discussion for a particular size of RHR leakage into the vault area. AS discussed throughout section 3, the V-sequence analysis includes event trees modeling of the V-sequence oveh 'the complete range of RHR system overpressure failure modes and leakage rates. The MAAP model simply included both the possibility of CBS pump actuation and the possibility of CBS pump flood-out. In the MAAP model, if actuated, the CBS pumps operate as long as they are not subme rged . (see Table 3-7).
As discussed in response number 52, the event trees include pump failure probabilities as a function of RHR leakage for all pumps in the vault area.
RAI 58
The last paragraph on page 3-11 contains a number of timing of event state-ments. Please provide justification of each. Plots of plant behavior showing suitable parameters and indicating the event points are sufficient for most. Operator response information, in addition to RCS parameter information, is necessary to substantiate the statement that RCPs will be tripped within about 21 seconds of break initiation.
RESPONSE 58 The las: paragraph of page 3-11 discusses the timing of certain events as predicted by the MAAP program for one particular V-sequence event involving the maximum expected RHR pump seal leak area of 1.3 squared inches for each pump. This particular analysis and its results are discussed further throughout section 4 of PLG-0432. Table 4-7 provides further sequence event timing information. Plots of several plant parameters versus time (as pre-dicted by MAAP) are provided in Section 4 including the following:
- Primary System Pressure (Figure 4-11)
- Core Water Temperature ' Figure 9-12)
- Vault Water Level (Figu e 4-20)
- Seal LOCA Flow Rate Into Vault (Figure 4-19)
- Containment Pressure (Figure 4-13)
Enhanced plots of the information provided there are included here as Figures I and 2; they show RCS pressure vs. time with various set points superimposed and subcooling vs. time with Reactor Coolant Pump Trip criteria superimposed.
CBS pump timing in Paragraph 3-11 is based on an approximate hand calculation for one particular break size. Further CBS pump operation is discussed in responses 52 and 56.
Page 3-11 notes that MAAP predicts the RCS to be solid within 30 seconds; this particular event, which is of no significance to the V-sequence, is probably inaccurately predicted by MAAP because the R'IR relief valve discharge to the Pressurizer Relief Tank was actually modeled via the PORV as discussed in Section 4.4.3 Operator Response Information:
The Response to Reactor Trip or Safety Injection (E-0) would be the first procedure utilized by the operators.
The E-0 procedure specifies on the foldout page (see attachment) that the Reactor Coolant Pumps be tripped whenever the following criteria are met.
TRIP ALL RCPs IF ANY CONDITIONS LISTED BELOW OCCUR:
- and -
RCS subcooling - less than 30*F Phase B Containment Isolation (loss of PCCW)
These instructions are valid throughout E-0. For further information on this
, subject, see included excerpt frota Westinghouse ERP Users Guide on RCP Trip section 2.3.2, Evaluation of Alternate RCP Trip Criteria. The-
'ttrip Criteria, 'e RCPS is a function of the present ERP sets which is trained of th ,
upon in both licensed operator training and requalification training'.
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l (E-0) 00 / 05/16/86 !
OPERATOR ACTION
SUMMARY
FOR E-0 SERIES PROCEDURES ;
el CCPs or SI pumps - AT LEAST ONE RUNNING
-AND-RCS Subcooling - LESS THAN 30*F e Phase B containment isolation (loss of PCCW)
- 2. ECCS ACTUATION CRITERIA Actuate SI and go to E-0, REACTOR TRIP OR SAFETY INJECTION , Step 1, if EITHER condition listed below occurs:
e RCS subcooling - LESS THAN 30'F
.1
-OR- ,
o Pressurizer level - CANNOT BE MAINTAINED GREATER THAN '
5% [(35)% FOR ADVERSE CONTAINMENT l
- 4. RED PATH
SUMMARY
- ATTACHMENT F
- 5. KEY CAUTIONS e If offsite power.is lost after SI reset, manual action may be required to restart safeguard equipment.
e RCS pressure should be monitored. If RCS pressure drops below 260 PSIG, RHR pumps must be manually restarted to supply water to RCS.
..q -
The appropriate instrument uncertainties should be added to the RCS pressure value established above. For normal containment conditions, the normal instrument uncertainties should be used, whereas with adverse containment conditions, the instrument uncertainties associated with post-accident containment conditions should be used. The instrument uncertainties should be determined for both the RCS pressure measurement and the steam generator l pressure measurement, and the values should be combined in an appropriate manner to obtain the total uncertainty. The resulting two pressures are the l indicated RCS pressure setpoints at which the operator should trip the RCPs, cepending upon the steam generator pressure and the containment conditions.
To facilitate the use of this parameter, a curve or table can be used which shows the RCS pressure setpoint for RCP trip as a function of steam generator pressure for normal and for adverse containment conditions.
The setpoint for this parameter could also be expressed as a RCS/ steam generator pressure differential. With this method, the RCu team s generator pressure differential setpoint for RCP trip would be equal to the pressure difference from the steam canerator pressure measurement location to the RCS pressure measurement location established above plus the combined RCS and steam generator pressure measurement uncertainties.
2.3.2 Evaluation of Alternate RCP Trio Parameters Analyses have been performed to evaluate the effectiveness of the three alternate RCP trip parameters for small break LOCAs, SGTRs and non-LOCAs. For each of the accidents, a design basis accident was defined and analyses were performed for representative Westinghouse plants. The details of the accident l
analyses are presented in Reference 5.
l l
The objecthe of the small break LOCA analysis was to demonstrate that the alternate parameters would provide an indication of the need for RCP trip prior to the time when trip is,actually required. For acceptability per NRC letters83-10c and 10d, the time available from reaching a setpoint that indicates the need for RCP trip to the time RCP trip is actually required RCP TRIP HP/LP-Rev. 1 0069V:1 17
should be at least 2 minutes. To provide a conservative minimum time period to trip the RCPs for each of the alternate RCP trip parameters, the small break LOCA analysis was based on Appendix K assumptions and the Westinghouse Small Break Evaluation Mocel ' FLASH f /LOCTA-IV Codes. The use of best estimate assumptions and mocels would result in longer time periods than tnose cotained with the Appendix K assumptions and models.
The results of the small break LOCA analysis demonstrate that the three alternate RCP trip parameters (RCS pressure, RCS subcooling and RCS/ steam generator AP) are essentially equivalent in providing an indication to the operator to trip the RCPs during a small break LOCA transient. The results also show that each of the parameters will provide the indication for RCP trip sufficiently early such that more than 2 minutes are available for operator action between the time the RCP trip setpoint is reached and the time when trip is required. This was demonstrated for each of the RCP trip parameters without adding any instrument uncertainty in determining the RCP trip set-points. Thus, each of the alternate RCP trip parameters will satisfactorily indicate the need for RCP trip for a small break LOCA with the instrument uncertainties based on either normal or adverse containment conditions. 5 Because each of the alternate RCP trip parameters are adequate to quickly provide an indication of the need for trip during a small break LOCA, the choice of which one to implement at a given plant may therefore ce based upon the discrimination capability for SGTRs and non-LOCAs and other plant specific instrumentation considerations.
For the SGTR and non-LOCA events, design basis accioents were defined and analyses were performed to determine the behavior of the alternate RCP trip parameters. The cesign basis SGTR was defined as a double ended rupture of one s' team generator tube on the outlet sice of the steam generator. The non-LOCA analyses were performed for credible steamline and feedline breaks since it was determined that these accidents result in the most limiting transients among the non-LOCAs considered. The design basis steamline break was defined as an unisolable break of approximately 4-1/2 inches in diameter in one steamline, which is equivalent to one steam generator PORV failing RCP TRIP HP/LP-Rev. 1 0069V:1 18
RAI 65
Relative water levels in the RHR vaults and the RCS are mentioned on pages 3-35 and 3-36. What are the water volumes in these regions as a function of elevation? (Of particular interest is the level at the top of the core and at the elevation of the hot leg connections to the RHR.)
RESPONSE 65 The volume of the RCS (Reactor Vessel and attached Cold Leg and Hot Leg piping) is 3,332 cubic feet. The center line of the reactor coolant piping lies at (-) 9 feet. The reactor vessel volume at the top of the core is 2,362 cubic feet. The top of the active fuel lies at -14' 2". The connec-tion of the RRR suction nozzle to the loop piping is shown in the attached figure. The equipment elevations, flood elevtions and flood volumes are shown in the attached figure for the containment building spray pumps, the CBS vault sump pump, the RHR pump and the safety injection pump.
, , . - - _y. -
l FICURE I RHR CONNECTIONS TO REACTOR COOLANT LOOP PIPING
- 22 I
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.g
' d a
e Motor Center hne Elevation (q 47' / Flood Elevation 47 6 FT id7 y , y
/ Flood Volume 23634 FT3 I II Ccntainment Building Spray Pump g Motor Center Line Elevat on (-) 58' 5' /
Flood Elevation 58 4 FT ! ! ,/ RH A Pump Flood Volume 4783 FT3 e , o.scharge E'2vation H 56' 6" I #
N Seal Flood Elevation 55 0 FT
\ , y y y Q-*' Seal Flood Volume 10729 FT3 s= .
f m* '
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i CBS vault
/
Sump Pump Flood Elevsten 59 5 FT Flood Volume 2898 FT3 Figure 2 RHR Equipment Vault Volumes and Elevations
RAI 66
What is the justification for the statement on page 3-36 that the water level in the vaults will be approximately the same as that in the RCS?
(We do not agree because of the potential that pressure in the vaults and containment are not the same, and water temperature in the two locations may differ).
RESPONSE 66 Page 3-36 (and preceeding) discusses potential operator actions to recover the V-sequence and prevent core melt. It states that if the LOCA can not be isolated, the water level in the vault will rise until the levels in the vault and in the RCS are approximately equal. This is a general statement made only in evaluating possible operator action; the exact levels are not important to operator action. As the question states, pressures and temperatures may differ in the vault and in the RCS/ containment; however, since these areas are communicating through the RRR leak, the dif ferences would not be significant. Also, the dif ferences might tend to force additional water into the vault area (since RCS/ containment pressure would be greater than vault pressure) providing additional fission product scrubbing. It should be noted that pressure / temperature differences of interest here are between the vault and the RCS not between the vault and containment. Other than RHR relief valve discharges, there is no communication between the containment and the RCS or vault area (at least until vessel melt-through) unless a RIIR relief valve should stick open.
RAI 74
A tacit assumption appears to be Jncorporated into References 1 and 2 that check valves are always closed. In reality, many check valves require a (substantial) reverse flow to force then to close, and they additionally often require a significant reverse pressure to keep them closed. Is this the case for any of the valves of interest here? If so, please discuss the implications. If not, what is the justification for the conclusion?
RESPONSE 74 It was not merely an assumption that resulted in the omission of sequences involving check valve failures to close but rather the result of a considered assessment. First of all, our knowledge of the design characteristics and tech spec surveillance requirements that govern the RHR system during normal power operation give us high confidence that the check valves in the RHR and interfacing systems are initially closed.
The check valves of interest here, those that separate the P.R M system from the CVCS, SI system, CBS, RWST and containment sump are all designed to close without the need to apply reverse pressure. Hence it is necessary to postulate that these valves failed to close the last time they were opened and that the failures remained undetected until the time of the postulated "V-sequence". Secondly, even if one assumes, conservatively th at all these check valves are initially open, our knowledge of check valve f ailure rates for f ailures to close, which have taken into account the San Onofre Unit 1 experience, gives us high confidence that any potential for check valve failed open bypass pathways is greatly over shadowed by the probability already assumed for RHR piping f ailure given RHR pressurization. In the discussion attached we presented our assessment of check valve failure rates in light of San Onofre Unit 1 experience and our conservative bounding assessment of sequences involving assumed failed open check valves. In view of these assessments, the conclusions of the sensitivity study are generally insensitive to assumptions regarding check valve performance.
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, as modeled in the Seabrook EPZ study (Reference 1). The failure modes of concern are disc rupture or gross leakage of a check valve that is initially seated and tested to verify its position 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 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 considerti for the failure mode of interest in the V-sequence analysis, either in terms of mode or cause of failure.
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 i discharge check valves to reseat on demand resulting in i overpressurization of the main feedwater system. A summary of the event, l
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 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
l 1423P091686 l
I i
TABLE 1. CHECK VALVE LEAKAGE EVENT DATA BASE Sheet 1 of 2 NPE Plant Event Description Range Reference (date) (gpal- ,
i
~
V11. A.126 Zion 2 A leak rate of ~0.25 gpa was detected from y - 0.25 (October 1975) the "A" accumulator check valve - wrong size gasket installed. -
V11.A.31 Turkey Point 4 One of the three check valves in the high-head y -~0.33 (May 1973) safety injection lines to the RCS cold legs developed 1/3 gpa leakage with 180 psi of water pressure applied. Two other check valves showed only slight leakage - failure of sof t seats.
V11.A.175 San Onofre 1 A tilting disc check valve located in the LPI y<5 (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 (July 1976) dilution in the "B" accumulator. y < 10 l ;
V11. A.182 Calvert Cliffs 2 The outlet check valves associated with the y < 10 1 (September 1978) safety injection tanks 218 and 228 leakea y < 10 reducing the boron concentration from 1,724 and 1,731 ppe to 1,652 and 1,594 ppe in 1-month period, respectively.
Vll.A.306 McGuire 1 Discharge check valves associated with the cold y < 10 (April 1981) leg injection accumulator A leaked - cause y < 10 unspecified.
Vll.A.343 Point Beach Check valve 1-853C, serving as the first-off y < 10
, (October 1981) check valve from the RCS for the low head safety injection.
V11.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) boren dilution (from about 2,550 down to 1,617 ppm) - cause unknown.
! Vll.A.85 Surry 1 Check valve associated with the 1C accumulator y < 20
( August 1975) failed to seat, resulting in increase in
. accumulator level - cause unspecified.
developed 6 gpa leakage.
V11. A.105 Robinson 2 "8" 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 j 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 i leading to joint binding.
1422P091686 2
TABLE 1 (continued) i Sheet 2 of 2 NPE Plant Leak Rate vent Description Range Reference (date)
(gpa)-
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 betseen disc and seat.
Y.A.273 Davis Besse 1 Gross back leakage through core flood check 20 < y < 50 (October 1980) valve - cause unspecified.
V11.A.384 Calvert Cliffs 1 jlT outlet check valve leaked at the rate of y -~200 (July 1982) 200 gpm ring deteriorated.
e e 1422P091686 3
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.-
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 provided in Table 3. The total number of check valve hours is 1.0 x 108.
1.3 FAILURE RATE ESTIMATE The 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
-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.
Using the data of Table 4, the following values for a and b were derived:
o Parameter a Mean = 0.0976 95th Percentile = 1.0127 Sth Percentile = 0.6915 e
1423P091686 4
TABLE 2.
SUMMARY
OF OPERATING EVENTS Event Percent System Date Power Involved Status Cause Reason for Inapplicability Vermont Yankee 12/12/75 99 LPCI/RUR Open Unknown PWR ECCS/RCS check valves are tested LER 75-24 and verified seated initially. They can not be lef t open undetected.
Cooper 01/21/77 97 HPCI Open Loose Part PWR ECCS/RCS check valves are tested 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-presure after being verified seated initially. These valves are closed and stay closed. They are not cycled; therefore, the failure modes are different.
LaSalle-1 09/14/83 0 LPCI Open Maintenance Errors PWR ECCS/RCS check valves ar'e tested LER 83-105/01T and verified seated before the plant '
goes to operation.
Pilgrim 09/29/83 98 HPCI Open Rusted Linkage on LER 83-48 Air Operator Ha tch-2 10/28/83 90 LPCI Open Maintenance Errors PWR ECCS/:CS 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 08/14/84 100 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.
I 1422P0916%
TABLE 3. CHECK VALVE EXPOSURE DATA Number 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 ARI December 1974 10 20 1.75+6 Crystal River 3 CR3 March 1977 7. 7 5' - 23 1.56+6 Davis-Besse 1 DB1 November 1977 7.08 x 26 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 C1tffs 1 CCI May 1975 9.58 45 3.78+6 Calvert Cliffs 2 CC2 April 1977 7.67 45 3.02+6 Fort Calhoun FCI September 1973 10.25 45 4.04+6 Millstone 2 MI2 December 1975 9 47 3.71+6 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 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 D. C. Cook 2 DC2 July 1978 6.42 34 1.91+6 Haddam Neck HN1 1968 14 27 3.31+6 Indian Point 2 IP2 Januarh74 July 1 10.42 36 3.29+6 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 TR1 May 1976 8.58 22 1.65+6 Turkey Point 3 TU3 December 1972 12 34 3.57+6 Turkey Point 4 TU4 September 1973 11.25 34 3.35+6 Yankee Rowe YR1 June 1961 14 17 2.08+6 Zion 1 ZIl Decen6er 1973 11 50 4.82+6 Zion 2 ZI2 September 1974 10.25 50 4.49+6 Total 1.08+8 NOTE: Exponer tf al notation is indicated in abbreviated form; i.e., 1.75+6 = 1.75 x 106 ,
1422P091686
a TABLE 4. STATISTICAL DATA ON CHECK VALVE LEAKAGE EVENTS IN PWR, ECCS, AND RCS SYSTEMS F
Leak Rate Number of gf 0c u re ce requency of (gpm) Events Exceedance (per hour) i 6 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
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1422P091686 i
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- - - - STATISTICAL 80UNOS \ ~
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AT 90% CONFIDENCE \ -
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BEST FIT
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= - ASSUMEO 95TH AND STH PERCENTILES \ ~
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1x 10'II ' ' ' ' ' ' ' ' ' ' ' ' ' ''' ' ' ' ' ' ' ' '
I to 100 1.000 10.000 CHECK VALVE LEAK RATE (GPM)
FIGURE 1. FREQUENCY OF CHECK VALVE LEAKAGE EVENi5 8
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 s.
with the following bounds:
95th Percentile = -y = 14.2862 + 1.0127 x Sth Percentile = -y = 13.1024 + 0.6915 x 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 F'gure 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 i
failures (failures involving pump discharge check valves) apply to this mode of failure. NPE was reviewed for the period January 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
demands. The corresponding failure frequency estimate (counting four of the five check valve failures) is A
so
= = 3.2 x 10 -3 er demand. '
This value was used together with the generic estimates as well as -
plant-specific data from other plants in a Bayesian updating process described in Reference 4 to develop the failure on demand frequency distribution. '
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 EPRI-81 7.00 x 10-5 in 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 THI-1 12 8,716 e San Onofre Unit 1 Main Feedwater System Check Valves Estimate = 3.23 x 10-5 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.)
- 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.
10 1423P091686
The resulting distribution is shown in Figure 2. Some key characteristics are:
ean Median Perce tile Perc ntile _
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
- 3. REFERENCES
- 1. Fleming, K. N., A. Torri, K. Woodard, and R. K. Deremer, "Seabrook Station Risk Management and Emergency Planning Study," prepared for Public Service Company of New Hampshire and Yankee Atomic Electric Company, confidential, PLG-0432, December 1985.
- 2. Nuclear Power Experience, S. M. Stoller Corporation, updated monthly.
- 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.
1 l
l l
1423P091686 11
,u WASH 1400 NUR EG-1363 EPRI-81 SAN ONOFRE 1 3
G LUMPED DATA FROM 8 PLANTS
! 3 (22 FAILURES IN 56,903 DEMANDS)
O a
b e
m
/
l l l 10-6 10-5 10-4 10-3 10-2 i
FAILURES PER DEMAND FIGURE 2. CHECK VALVE FAILURE ON DEMAND FREQUENCY DISTRIBUTION i
.m.
ATTACHMENT A t
(
I
- . PAGE I NPE C4ECK VALVE LISTING - 693 ARTICLES TOTAL THERE ARE 73 ARTICLE (5) To BE DISPLAYED: 7 gwR - CHECK VALVE FAILURE 3.- 19p'0-1975 3.04.8.0010 : VARIDUS CONTROL SLADE, DRIVE PR03LEMS 3RESDEN 1 - 1960-12 *>
8.04.8.3036 : DISK AND SEAT SURFACES DIRTY - CRD C4ECK WALVE LEAKAGE *
--- DRESDEN 2 - MAY 75 - REFJELING SHUTD0dN '
$.05.E.0002 : CHECK---
VALVE LEAKAGE DRESDEN 2 - SPRING 72 II 8.05.E.0003 : FW CHECK VALVE LEAKAGE
--- DRESDEN 2 - SPRING 71 J 9.05.E.0014 : FW CHECK VALVE LEAKASE - CHANGED T3 $1LICON SEAL RING
--- DRESDEN 3 - SPRING 1974 t 9.06.E.0018 : 3-RINGS ADDED TO FW CHECK VALVES
--- DRESDEN 3 - IST 1/2 0F 1773 J 9.06.E.0022 : CHECK VALVE
- LUG LEAK ~
--- OYSTER CREEK - JULY 73 J t
8.06.E.0023 : SILICONE
--* 0-RINGS UNACCEPTA8LE - VALVES LEAKED J DRESDEN 2 - MAR S MAY 74 B.06.E.0024 : FW--- CHECK DISC DID NOT .1 ATE WITH SEAT U MONTICELLO - SPRING 74 8.05.E.0029 : LEAKY GASKET IN FW CHECK WALVE ()
ARNOLD - JULY 74 (POJER ESCALATI3N TESTING) j 8.06.E.0034 : ,
C)
LEAKAGE BETWEEN CHECK VALWE SEAT RIN3 AND VALVE SODY
--- 2UAD-CITIES 1 - APR 74 (REFUELING SHUID3WN) 3.05.E.0040 : FLAW---
IN FW CHECK VALVE .I QUAD-CITIES 2 - FEB 75 - REFUELING SHUTDOWN 8.06.E.0042 : CHECK VALVE SEAL RING LEAKAGE J auAD-CITIES 2 - DEC 74 & 75 - REFUELING SHUTDOWN 9.06.E.0043 : SEAT LEAKAGE AND CRACKED SELLOWS IN FW CHECKS J
--- PEACH BOTTOM 2 - MAY E JUNE 75 - SHUTDOWN AND 655 P3WER
' 8.06.E.0048 : ERODED CONTR3L VALVES
, TARAPUR 1 8 2 - 1970 t "1
8.06.E.0058 : FW --- C4ECK VALWE 0-RINGS C098AESSED - EXCESSIVE LEAKAGE DRESDEN 3 - APR S MAY 75 - REFUELING SHUTDOW4 g 3.05.E.0059 : FW---CHECK VALVE 0-RINGS DISASSOCIATED DRESDEN 2 - N3V 74 - REFJELING SHUTDOWN 8.05.E.0075 : PIN---
HOLE LEA ($ IN FW CHEC( VALVE SEAL PLATE, CRACKED C01V3 LUTE PEACH 90TTOM 3 - AUG 75 - STARTING UP
--n , -.
e 5>f
o
, PAGE 2 8.06.E.0081 : CRACCED FW VALVE SEAL PLATE 3ELLOWS
--- PEACH 80TTOM 2 - JUNE 75 - 740 MdE j 3.05.G.0002 : MISC. TURSINE CYCLE SYSTEi$ PROBLE15
--- 9UHLE8ERS - 1972-14 ,y 8.07.A.0015 : CHECC VALVE DISC-HINGE JOINT FRACTURED
--- 2UAD-CITIES 1 - APR 74 (SHUTD0dN FOR REFUELING) ()
5.07.4.0020 : RCIC---TURBINE EXHAUST CHECC ASSEMBLED INCORRECTLY MONTICELLO - JAN 75 - SHUTDOWN ()
t.07.A.0031 : CHECC VALVE 3 LOWING STEAM, R3D CONTR3LS N3T CHECKED FITIPAT91CK - JULY 75 - (POWER ESCALATION TESTINS) -
J SHUTDodN 5.07.A.0032 : EXHAUST ---
LINE CHECC VALVE CLAPPER DISCONNECTED #
3RUNSWICC 2 - APRIL 75 - (POST CRITICAL TESTING) 85 P3WER S.07.A.0034 : EXPAUST CHECC VALdE FLAPPER WAS J41MED J
--- BRUNSWICC 2 - OCT 75 - 635 POWER B.07.A.0040 : RCIC ---
TUR8INE DISCHARGE CHECK VALVE FLAPPER DISC STUD 3R3EEN #
1 3RUNSdICC 2 - FE4 75 - H3T STAND 3Y 8.07.A.0043 : TORUS DISCHARGE VALVES HAD DIRTY SEATS AND MISSI4G FASTENERS J
--- COOPER - OCT 75 - SHUTD0dN I
8.07.C.0014 : SCALE ON VALdE SEATS St
--- MONTICELLO - SPRING 74 4
8.07.C.0030 : SCALE ON CHECC VALVE SEATS 10NTICELLO - JAN 75 - SHUTDOWN -
s,07.C.0042 : 11Pa0PERLY A3 JUSTED CHECK VALVES l ---
BRUNSWICC 2 - MARCH 75 (PDST CRITICAL TESTING)J 55 PCJER '
8.07.C.0043 : CHECC VALVE SEAL LEACED d TARAPUR 1 & 2 - 1969 04 F0 3.07.C.0044 : STUCC CHECC d4LVES "'
TARAPUR 1 & 2 - PRIOR TO 1973 8.0T.C.0047 : SCALE ON SEATING SURF 4CES OF CORE SPRAY CNECK VALVES -'
MONTICELLO - SEPT 75 8.07.C.0061 : CORE SPRAY C1ECK VALVES SEATED IMPROPERLY - DISCHARGE HEADER DEPRESSURIIE3
--- DRESDEN 2 - JAN 75 - REFUELING $NUTD0dN --- DRESDEN 3 - AUG
- 74 i ~# I 1
a.07.D.0016 : WIRE AND G1!1 DER ARBOR FOUND IN RNR SYSTE1 2UAD-CITIES 2 - JAN 13 g i
S.07.D.0035 : SCALE ON VALdE SEAT
--- MONTICELLO - SPRING 74 ()
i
- 8.07.0.0061 : SCALE ON CHECC VALVE SEAT -
MONTICELLO - JAN 75 - SHUTDOWN C
o PAGE 3 8.07.0.3076 : RHR ---
AND HPCI VALVE PACKING LEAKS 3RUNSWICK 2
- SEPT 75 - 52% P0JER #
3.07.0.0082 : VALVE LEAKAGE, LPCI LOOP 3VERPRESSURIZED - HEAT EXCHANGER GASKET FAILED gj
--- WERMONT TANKEE - DEC 75 - 99X PodER 3.07.D.3106 : TORUS SPRAY VALVE OPERATOR MOUNTINS PLATE WELD FAILED, CRACKED Y3KE ARM - 115UFFICIENT JELD PENETRATION, UNDERSIZED MOUNTINS .
80LTS
--- 3ROWNS FERRY 1 - MAY L SEPT 76 - COLD SHUTD0dN -
C S.07.E.0007 : STEAM VALVE DISC PIN FAILURE - AUPTURE DISCS RUPTURE - TEMP ^
SdITCH DAMAGE j 10NTICELLO - JULY 72 i
' 3.07.E.0005 's STEAM VALVE LEAKAGE '
10NTICELLO - JULY 72 3.07.E.0011 : LODSE RUST - CHECC VALVE LEAKED TORUS WATER BACK T3 WARDS 1PCI TURBINE < ~
PILGRIM - JULY 72 (P0 DER ESCALATION TESTING)
J 8.07.E.0016 : VACUJM IN TUR8INE EXHAUST - d&TER NAMMER WERMONT fANKEE.- 1971 (PRECPERATIONAL TESTING)
.)
8.07.E.0025 : WATER HAMMER OF TURBINE EEHAUST - INSTALLED CONDENSING SPARGER BROWNS FERef 1 - OCT T2 (PREOP TESTING)
[ gj 8.07.E.0068 : GLAND CONDENSER GASKET FAILED - P3SSIBLE INJECTION VALVE LEAKAGE
, 3ROWNS FERRY 2 - NOV T4 - 42X POWER O
B.07.E.0053 : SURFACE FLAWS IN VALVE DISC AND SEAT
--- PEACH BOTTOM 2 - FE8 75 - SHJTD0JN #
, S) 8.07.E.0054 : SENT HINGE PIN AND BAD DISC 3N TUR811E EXHAUST CHECK VALVE
, --- 10NTICELLO - JAN 75 - SHUTDOWN 8.07.E.0058 : MISSING WELD IN SWING CHECK, INSUFFICIENT TACK WELDS IP STOP
~ CHECK, VALVE DISSASSEMBLED WITH C30LANT, GREATER TNA1 212 DEGREES F '
ARNOLD - APR 75 l* 8.07.E.0061 : STEA9 VALVE LEAKED QUAD-CITIES 2 - DEC 76 - REFUELI1G SHUTDOWN a.07.E.0077 : SCRATCHED SEATING SURFACES - TURBINE EXHAUST VALVE LEAKAGE "#
MONTICELLO - SEPT 75 8.07.E.0130 : HPCI TURBINE EXHAUST CHECK VALVES LEAKED - WORN VALVE SEATS ' #
--- 2UAD-CITIES 2 - DEC 76 & SEP TS - REF'JELING --- 30AD CITIES
, 1 - JAN 76 - REFUELING b
8.07.F.3007 : VARI 3US PR08LEMS WITH VALVE LIMITORQUE OPERATORS
, --- DRESDEN 2 & 3 - DEC 70 - JAN T1 8.08.C.0055 : DIRT IN SERVICE WATER CHECK VALVE
, --- DYSTER CREEK 1 - MAY 15 - 390 MWE gg
s PAGE 6 a
8.08.C.3064 : CRo$$--- THRE4DED PIPE S LOOSE VALVE $0LTS - LEAKS
- 3RUNSdICC 2 - JUL 75 - FE POWER I
9.07.F.0006 : CHECC---
VALVE HAD MISSING DISK AND P3P*ET 2UAD-CITIES 1 - JUL 76 - SHUT 03WN S) 8.09.F.0007 : INSTRUMENT CHECK VALVES HAD dEAK SPRINGS, 'ITTED POPPETS AND 840 SEAT auAD-CITIES 2 - APR 75 - COLD SH'J T00W4 9.09.G.3147 : DAMASED ---
SEAT ON INSTRUMCNT N2 VALVE PEACH BOTTOM 2 - M&T 75 - 1001 P3WER O 3.11.A.0057 : STUCC ---
OIESEL FUEL 8ALL CHECK VALVE VERMONT TANKEE - JUL 74 j 8.11.A.0103 : STUCC DIESEL AIR CHECK VALVE
--- 3RUNSWICK 2 - DEC 75 - 360 MWE B.14.8.0017 : SECONDARY C04TAIN1ENT VIOLATIONS DRESDEN 2 & 3 - OCT 71 - MAR 72 #
3.16.8.0037 : CHECC VALVES LEAKED DRESDEN 3 - JAN 13 9.14.8.0056 : CONTAINMENT CHECK VALVE MISALIGN9ENT MONTICELLO - NOV 73 8.14.8.0059 : LEAKING --- VACUJM BREAKER PE4ETRATIONS - INADVERTENT RELEASE PEACH BOTTOM 2 - DEC 73 (P0 DER ESCALATICN TESTIN3) ,
gj 8.14.8.3072 : VACU3M BREAKER LEAKA3E
--- 3YSTER C4EEK 1 - APR 74 - 642 9WE 3.14.8.3081 : DEFORMATION 3F CHECK WALVE RUSSER SEAL #
--- PEACH 80TTOM 3 - MAR 75 - 1005 P3WER -
gg
=
8.14.8 3088 : REVERSE ---
FLOW CHECKS REMOVED FROM SGTS - M3 TOR OVERLOAD COOPER - 9AT 75 - COLD SHUT 00d4 J 3.16.8.3115 : CHECK VALVE PISTON FOUND MUNG UP
--* PEACH 80TTOM 3 - DCT 75 - 553 PodER 8.15.A.0007 : THIN WALLED WALVES - BWRS IN GENERAL - 1970 ((P.15.3))
8.15.A.0037 : NITR3 ---
GEN ACCJ4ULATOR VALVE LEAKED BRUNSWICt 2 - DEC 75 - 85X POWER
.b 3.16.C.0001 : QUESTION ON SUITABILITY OF CERTAIN PIPING AND VALVES
--- 3YSTER CREEK 1 - JUL 59
\ 0 8.16.C.0036 :
PROCEDURAL PROBLE9 - LPCI VALVE MOTOR BUR 1ED OUT DURIMG TEST.
{ --- VERMONT TANKEE - MAY 73 g
8.15.C.3099 : PROCEDURAL PROBLEM - EXCESS FLOW CHECK VALVE'S OfPASS VALtES LEFT OPEN PEACH 10TTOM 2 & 3 - NOV 74 - 103% & SOE POWER
............. ALL 00NE, PRESS < RETURN > (EY TO G3 T3 MENJ .................?
()
r - -
p ~ - -
s ' ,
SAGE 5 THERE ARE 123 ARTICLE (S) TO 8E DISPLAYED: 19-4UG-1986 13:37:37 BWR CHECK VALVES, 1976 - 1980 8.04.8.3047 : CRD RETURN LINE ISOL4 TION VALVE H43 WORN SEAT, RJST 04 PISTON '
--- 3RESDEM 3 - OCT 16 - REFUELING SHUTDOWN S.04.8.0114 : CONTRCL
--- ROD DID NOT LATCH - DIRECTIONAL VALVE FAILED #
BROWNS FERRY Z - JUN 80 - 40X PodER 5.04.8.3117 : WEST ---
SDV HEADERS DID NOT DRAIN AFTER M4NU4L SCRA9 3RESDEN 3 - J UL SD - SHUTDOWN d.04.8.3145 : CONTROL R00 3 RIVE SYSTEM 14LFUNCTIONS C
- 3RUNSWICK 1 - AUG 80 - SJBCRITIC4L (STARTUP) --- 3RJNSWICK 2 FES 81 - 1.5% P3WER --- OYSTER CREEK -N3v 80 - 63F PodER 8.05.C.0257 : THREADED LOCKING DEVICES 3N VALVES, PUMPS, VALVE OPERAT3R$
FAILED ~
3WR'S IN GENERAL - MAR 83 ((P.083.293)) '
S.05.C.3261 : ADS LIR SUPPLY ACCUMULAT3R CHECK W4LWES LEAKED
--- HATCH 2 APR 33 - COLD SHUTD0d4 ~#
3.05.C.3266 : ADSA ---
AIR SUPPLY ACCU 1ULAT3R CHECK WALVES LEAKED COOPER - APR 80 " REFUELING j a.06.E.0054 : RUST ON FW C4ECK VALVE SEATS PEACH SOTTOM 2 - MAR T6 - SHUTDOWN >
3.06.E.0055 : Fd CHECK VALWE LEAKAGE - EXCESSIVE DISC-73-SEAT CLEARANCE DRESDEN 2 - MAR 76 - MEFUELING $4UTD0dN gj B.06.E.0060 : Fd CHECK VALWES LEAKED - WITON 0-RINGS DETERIORATED .
3UAD-CITIES 1 - JAN 75 - REFUELI13 SHUTDOWN g) 8.06.E.0064 : DIRT IN FW CHECK VALVES, IMPROPER DISC-TO-SEAT CLEARANCE, '
EXCESSIVE LE4KAGE - TESTING METHODS dERE INADEQUATE gg DRESDEN 3 - SEPT & OCT 76 - REFUELING SHUTDOWN S.06.E.0065 : FW S4MPLE PR3BES LODGED I4 HPCI,& FW CHECK VALVES, FLLNSE #
JAN ?? - APPROXIMATELY 76% POWER; BRUNSWICK 2 - FE8 76 - SHUTDOW1 a.06.E.3076 : WORN FW VALVE SEAT / DISC ASSEMBLY PINS, DEFORMED SEAT RI4GS DRESDEN 2 - OCT FT - REFUELING $4UTD0dN "
8.06.E.0086 : FW CHECK VALWES SEATS MACHINED. SEAT SEAL 0-RINGS CH4MGED To KALREI J
--- 3UAD-CITIES 2 - SEP) 76 - REFUELING 8.06.E.0093 : W3RN---
SEAT / DISC ASSEMBLY PINS IN FW CHECKS I J
DRESDEN 3 - MAR 78 - REFUELING ,
s 9.06.E.0097 : WORN---
SEATS AMD RINGS IN Fd CHECK WALVES U MILLSTONE 1 - MAR 78 - REFUELING S.06.E.0107 : DIRT---
ON FW C4ECK V4LVE SEAT '
QUAD-CITIES 1 - JAN 77 - REFUELI1G 8.0$.E.0109 : WORN SUSHINGS - FW CHECK WALWES LE4KED
i
. PAGE 6 s
3RUNSWICC 1 - APR 79
- REFUELI1G 9.05.E.0110 : WORN ---
FW CHECC VALVE SU$HINGS '
3RUNSWICC 2 - MAY 79 - REFUELING s.06.E.3112 : FW CHECK V4LdES LEAKED - SEATS REPAIRED, SEAL REPLACED '
--- MILLSTONE 1 - JUNE 79 - REFUELING B.06.E.0116 : Fd CHECK ---
VALVE KING PIN C3VER LEAK = HIGH CONTAINMENT TEMP 3RUNSWICK 2 - AUG 79 - 945 P3 DER S.06.E.0127 : FW --- CHECK VALWE LEAK - WOR 4 SEAT / DISC ASSE18LT PINS DRESDEN 3 - Ftt 80 - REFUELIN3 8.06.E.0131 : FW LEAKE0 ---
INTO TORUS UNDETECTED - CHECK VALVE B04 NET SEAL FAILED J VERMONT fANKEE - JUNE 80 - 355 P3WER S.06.E.3134 : FW --- CHECK VALdE SEAT / DISC ASSEMBLY PI4S WORN DRESDEN 3 - MAR SD - REFUELING 8.06.E.0141 : dORN SEAT, DIRT IN FW CHECK VALVES OUAD-CITIES 1 - SEPT 80 - REFUELING SHUTDOWN 3.0T.A.0058 : SENT DISC WASHER, IMPROPER SEATING I1 RCIC EXHAUST VALVE 90NTICELLO - OCT 77 - REFUELIN3 SHUTD3WN 9.07.A.0059 : ROUGH SEATINS SURFACES IN RCIC STEAM VALVES J
--- COOPER - OCT 77 - REFJELING SHUTDOWN 9.07.A.3062 : RCIC TURSINE STE41 EXHAUST CHECK WALWE LEAKED - FLAPPER Bt0KE I 3UAD-CITIES 2 - SEPT T6 - REFUELING 9.37.A.0071 : ROUGH AND CRACKED EXHAUST VALVE SEATS
--- COOPER - APR 78 - REFUELING
/
a.07.A.0075 : DIRT ON TUR8INE EXHAUST VALVE SEATS
--- MONTICELLO - OCT 78 - REFUELI13 ,
S.07.A.3076 : RCIC VALVE PIN SHEARED #
--- BROWNS FERRY 1 - DEC T8 - REFUELING S.07.A.0080 : DIRTV VALVE INTERNALS 3UAD-CITIES 1 - JAN 77 - REFUELING 8.07.A.0083 : LOOSE VALVE DISC BLOCKED RCIC TURBINE EXHAUST LINE - BLOW 1' RUPTJRE DISC HATCH 2 - JUNE 79 - 35 P3WER B.0T.A.0085 : DEFECTIVE TORQUE SWITCH, ROUSH SEATS - STEAM SUPPLT AND EEHAU$t VALVES LEAKED
--- COOPER - APR 79 - REWUELING ' #
s.07.A.0094 : RCIC TUR8INE FAILED TO START - BOLT LODGED IN STEAM EEHAUST CHECK VALVE g SRUNSWICC 1 - NOV 79 - SHUTDod1 3.07.4.3095 : RCIC TRIP - CHECK VALVE LO K .8ROKE1, VIBRATED CLOSED
--- HATCH 1 - DEC 79 - 80C P3WER B.0F.A.0096 : RCIC DRAIN P3T LEVEL VALVE PACKING RJPTURED - NORMAL DEAR i
e
'PAGE T
=== AANOLD -
OCT 79 - 935 P0 DER 9.07.A.3103 : RCIC -**
BYPASS VALVE TORQUE SWITCH C31 TACTS DIRTT 3ROWNS FIRRY 3 - JULY 80 - 495 P3WER 8.07.A.0121 : RCIC --- TURSINE TRIPPED - STEAM EXHAUST CHECC VALVE FAILED
- 3RUNSWICC 2 - NOV 80 - 573 P0 DER 9.07.8.0018 : SSLC PUMP CHECK VALVE LEA (S
--- QUAD-CITIES 2 - MAR 76 - SHUTDOWN 9.07.8.0028 : CHEMICAL ---
DEP35!TS ON SBLC VALVE SEAT DRESDEN 3 - APR T8 - REFUELING SHUTD0dN 9.07.C.3070 : SCALE ---
ON VALVE SEATS 10NTICELL3 - SEPT 77 - REFUELING SHUTDOWN '
8.07.C.0083 : CORE SPRAY CHECK VALVE REMOVED FOR MAINTE4ANCE - CONTAINMENT VIOLATION 1 PILGRIM 1 - JJNE 79 - 103! P0 DER J
- 8.07.C.0090 : DATER HAMMER DAMA3ED PIPE SUPPORTS, VALVE 8GLTS LO3SE 1ILLSTONE 1 - FEa 80 - 100% P3 DER I
B.07.D.0103 : LPCI CHECK VALVE PACKING LEAC
--- MILLSTONE 1 - SEPT 76 - 72% P3 DER 1
i 3.07.D.0107 : NUT HOLDIN3 4HR CHECK VALVE CLAPPER SHAFT WAS MISSING
--- FITZPATRICK - OCT 76 - 28% POWER 8.07.0.3108 : DIRT AND SCRATCHES 04 LPCI CHECK VALVE SEAT
()
DRESDEN 3 - OCT 76 - REFUELING 34UTD0dN -
()
3.07.0.0115 : DIRT ON REACT 3R HEAD COOLING ISOLATI3N VALVE SEAT AND DISC P!VOT PIN /
DRESDEN 3 - NOV T6 - REFUELING SHUTD0dN s
9.07.0.0116 : SHEARED WC3DRUFF KEY IN R4R VALVE OPERATOR
--- COOPER - FEB TT - 964 POWER a.07.0.0119 : WORN LPCI VALVE COMPONENTS, SENT M373R SHAFT - VALVE SINDING
--- 3R0dNS FERRY 1 - MAR 77 - 851 PodER s.07.0.3153 : STRIPPE3 LPCI VALVE LOCKNUT, DAMAGED YOKE AND 8EARIN35 - VALVE SINDING 3R0dNS FERRY 3 - FEB T8 - 1003 P3WER 8.07.D.0169 : LOW ---
DP, RNR VALVES FAILED TO SEAT FULLY - HIGH TORUS LEVE.
SRUNSWICC 2 - NOV Td - H37 SHUTD3dN 8.07.D.0171 : LOOSE DISC RETAINING NUT, CHECK VALVES STUCK OPEN I '
FITZPATRICK - MAR E DEC 78 - DE POWER 9.07.D.0180 : CRAC(ED LPCI CHECK VALVE DRAIN LINE b
--- DRESDEN 2 - MAR 19 - REFUELING 8.07.0.0185 : BROKEN RNR CHECK VALVE DISC STEM NUT PIN ()
--- FITZPATRICK - JUNE 19 - COLD SHUTDOWN 8.07.D.0220 : TDRUS SUCTI01 VALVES FAILED DUE TO STEAR LEAK FR3M A CHECC V4LVE
8 a
PACE 8 HATCH 2 - MAY 80 - 99% P3WER 9.07.0.0236 : RHR ---
CHECK VALVE DID NOT SEAT WERMONT YANKEE - SEPT 80 - 89X P3WER 8.07.0.0238 : INADEQUATE SJPPORT - LPCI DRAIN LINE WELD LEAK *'
--- DRESDEN 2 .94Y $3 - SHUTDOWN 3.07.D.3259 : CHECC VALVE STUCK - PLANT MODIFICATI3N INSTALLED '
3RUNSWICC 2 - Nov 80 - 65E POWER 4
3.07.E.0076 : HPCI---
STEAM C1ECK VALVE GASKET FAILED QUAD-CITIES 1 - JAN 76 - REFUELING SNJTDOWN S.07.E.0081 : IMPR3PER MOU1 TING - CORROSION AND PITTING ON HPCI VALVE SEAT
- DRESDEN 2 - MAR 76 - SHUTDOWN B.07.E.3090 : SINDING IN H8CI TURSINE EXHAUST VALVES - RUPTURE DISC BL0dN #
--- 3ROWNS FERRY 3 - AUG 76 - 3% P3WER 8.07.E.0095 : LOOSE RETAINING MUTS, BROCEN RETAI411G STUDS DISLODGED DISC IN '
HPCI TURSINE EXHAUST CHECC VALVES
--- BRUNSWICC 1 & 2 - OCT 76 - COLD SHUTD3W1 o
a.07.E.0113 : HPCI TURBINE EXHAUST VALVE DISC STUD FAILED, MISSING PARTS
--- COOPER - APR ?? - COLD SHUTD0d1 S.07.E.3114 : HPCI FLOW OSCILLATIONS AT Lri SPEED - CHECK VALVE DISC HI1GE FRACTURED 3RUNSWICC 1 - MAY 77 - C3LD S H U '. 0 0 W N 8.07.E.3121 : DIRTY EXHAUST VALVE SEATS 10NTICELLO - SEP 77 - REFUELING SHUTDOWN O
, 3.07.E.3122 : ROUGH AND SCRATCHED WALVE SEATS #
j COOPER - DCT ?? - REFJELING SMUTDOWN t s 4)
)
4 a.07.E.3136 : STICKING HPCI CHECK VALVES, DATER HA9MER - SNUB 8ER SHAFT BROCEN SRUNSWICC 2 - MAR 78 - SHUTD0d1 i
8.07.E.0139 : MECHANICAL I1TERFERENCE IN CHECK VALVE i
--- 3ROWNS FERRY 2 - MAR 78 - REFUELING 8.07.E.0152 : GLAND SEAL C4ECK VALVE LEAKED j --- PEACH 80TTOM 2 - OCT 78 - $15 P0 DER '
' B.07.E.3155 : DIRT ON TURBINE EXHAUST VALVE SEATS j ---
10NTIC EL'.0 - 3CT 78 - REFUELING 8.07.E.0161 : CHECC VALVE SASKETED SEAT FAILED DUAD-CITIES 1 - JAN 77 - REFUELING ' d 9.07.E.0209 : TORUS SUCTIOM VALVES FAILED DUE TO STEAM LEAK FROM A CHECC VALVE HATCH 2 .9AY 80 - 995 POWER U 3 2.07.E.3213 : FOREIGN MATERIAL BETdEEN CHECK VALVE BODY & SEAT
--- SROWNS FERRY 2 - SEP BD - REFUELING SHUTDOWN )
~
3.07.7.0037 : UNION SETWEE4 ISOLATION FLOW CHECC VALVE AND STEAM LINE F.0W SENS3R WAS CROS$* THREADED $
s i
PAGE 9
--- DRESDE1 3 - N3V 76 - 5% 80WER 8.07.8.0050 : TILTING
--- DISC CHECK VALVE INSTALLED I4 WR01G ATTITUDE
- ALL BWRS - JUL ?$ ((P.07.A.1753)
B.07.F.3065 : PE88LE IN FIRE PUMP CHECK VALVE
--- FITIPATRICK - AUG 79 - C3LD SHUTDOWN 8.07.F.0066 : FIRE DRAIN C1ECK VALVE HAD OFF CEMTER DISK
--- BRUNSWICC 2 - JUL 79 - 8SI PDdER 3.07.F.0070 : VALVE SLOW T3 CLOSE - SWITCH NEEDED ADJUSTMENT CI
--- PILGRIM - OCT 79 - 83I 9.07.F.0071 : VARI 3US VALVES LEAKED - DIRTY SEATS *REVE4TED CLOSURE
--- PEACH BOTTOM 3 - NOV 79 - REFUELING 8.07.F.0072 : VARI 3US CONTAINMENT ISQLATIO1 VALVES' LEAK RATES UNACCEPTABLE
--- HATCH 1 - APR, MAf, DEC 79 - REFJELIN3 AND 972 P3 DER --- '
HATC4 2 - 94f 79 - SHUTD3dN 8.07.F.0078 : VARI 3US PRIMARY CONTAINMENT !$0LATION VALVE LEAKS J
--- 3UAD-CITIES 2 - NOV-DEC 79 - REFJELING FE8 83
- e 8.07.F.0051 : VARI 3US C01TAINMENT ISOLATI01 VALVES FAILED LLRT
--- 10NTICELLO - FE8 & MAR 83 - REFUELING a.07.F.3033 : VARI 3US VALVES LEAKED IN EXCESS OF TECH SPECS
--- HATCH 2 - MAR S APR 83 - COLD SHuTDOW1 8.07.F.0089 : $/
PRIMARY
--- CONTAINMENT COOPER - PENETRATION LEAKS EXCEEDED TECH $*ECS MAY 80 - REFUELING '
5.07.F.3131 : VARI 3US VALVE LEAKS O MILLSTONE 1 - OCT 80 - REFUELI4G '
8.08.C.0086 : ,
63 EXCESSIVE ARM PL4f BETWEEN SERVICE WATER CHECK VALVE DISC AND DISC VERMONT fANKEE - AUG 76 - 95I PodER #
B.09.C.3121 : LEAKING DRAI4 VALVE, COCKED SPRI4G I4 RECIRC PUMP SEA'. PRES $JRE 4
VALVE
< s l
--- COOPER - OCT 77 - REFUELING SHJTDOWN 1
' 3.09.C.3171 : CORR ---
3DED SERWICE WATER VAULT DRAIN VALVE SPRINGS #
3UAD-CITIES 1 - FEB 77 - REFUELI1G 8.08.C.0183 : CHECK
--- VALVES DAMAGED BY REVERSE PRESSURE - VALVE SPRING I4 PUMP #
SRUNSWICK 2 - AU3 79 - 33I P0 DER S.03.C.3199 : SW P'MP J LOCK COLLET 3ISENSAGED FR31 IMPELLER AND SHAFT
--- HATCH 1 - JAN 80 - 941 P3WER J 8.08.C.0201 : RHR --- SW SEDPLATE DRAIN VALVE FAILED - OPERATOR SPRI4G CORR 3DED OI l 3UAD-CITIES 2 - DEC 77 - REFUELI4G 8.03.C.0210 : RHR SW CHECK VALVE LEAKED
- --- BROWNS FERRY Z - AUG 80 - 91I P0JER I) 8.07.A.0045
- DIRT ON TIP PURGE CHECK VALVE 8ALL OR SEAT
t PAGE 10
--- DRESDEN 3 - OCT 76 - REFUELING SHUTD0dN i 8.09.8.3085 : LOOSE LEVEL INSTRUMENT SENSI1G LINE CHECK VALVE NUT
--- COOPER - DEC 17 - 75% P0JER 8.09.E.3353 : DIRT IN JET PUMP RISER DP INSTRUMENT LINE VALVE '
--- QUAD-CITIES 2 - MAR 78 - REFUELING SHUTDOWN 3.07.E.3616 : NO ---INTERNALS IN INSTRUMENT LINE CHECC VALVE O 3ROWNS FERRY 2 - NOV SD - SHUT 30JN
+ 3.09.F.0016 : CRUD ---
ON INSTRUMENT LINE CdECK VALVE POPPET COOPER - APR 78 - REFUELING 4 9.07.F.0025 : RCP DP INSTRJ9ENT LINE CHECK VALVE LEAKED - POSSIBLE FOREIGN MATERIAL ON SEAT ,
J HATCH 1 - JUN 78 - 23 POWER J
8.09.F.0026 : DIRT IN INSTRUMENT LINE CHECC VALVE
--- QUAD-CITIES 1 - FE8 77 - REFUELI4G b
-l 3.07.G.3103 : SERVICE AND INSTRUMENT AIR STSTEM DEPRES$URIIAi!3NS, STUCC CHECC VALVE, REACTOR TRIP
--- 3RUNSWICC 2 - MAY 76 - 18% P3 DER, JUL 76
- j B.07.G.0121 : 0-RINGS LEAKED IN CONTAIN1ENT OXYGEN SAMPLING VALVES
--- NINE MILE PT 1 - APR 77 - OMSNUTDOWN
- 9.09.G.3151
- STUCK
--- SALL F'04T CHECC VALVE - WATER IN RADIATION 1011T3R FITIPATRICK - JAN 78 - 99E P3 DER gj 3.07.G.0171 : INADEQUATE DRAINING, LEACING FITTINGS, CONDENSATION .M3ISTURE -
AND AIR LEAKAGE INTO CAC 90NITORS g)
BRUNSWICC 1 - MAR 78 - 55%-1005 POWER --- 8RUNSWICK 2 - MAR 78 - APPROX 11ATELY 97% P3 DER f
. NN 8.07.G.3178 : LEAKING FITTINGS, STUCK VALVES, CONDENSATION - RADIATIO 1 AND HYDR 3 GEN - OEYGEN MONITORS FAILED 3RUNSWICC 1 - MAR & APR 78 - 130E POWER --- 3RUNSd!CC 2 - J APR 78 - 32% POWER 8.09.G.3179 : CRACCED PUMP SHAFT, WATER IN SAMPLE LINE, STUCK VALVES LOW FLOWS IN CAC MONITORS 3RUNSWICC 1 - APR & MAY 78 - 69X TO 92E POWER --- BRU1SWICC l.
i 2 - 1AY 78 - 100% P0 DER '
8.07.G.3234 : INSTR.
CABINET EXHAUST CHECK STUCC FROM C3RRQSION SRUNSWICC 1 - JUN 79 - 130% PodER 'j a.09.G.0265 : ADS HARD-SEATED CHECK VALVE REPLACED WITH SOFT-SEATED DESIGN, SUPP3RTS ADDED g
! SWRS IN GENERAL --- PEACN 80TT3M 2 - JAN 80 - STARTUP (41X i POWER) --* PEACH 30TT3M 3 - JAN 30 - SHUTDOWN 3.11.A.3134 : DG AIR RECEIVER CHECC VALWE RUSTED CLOSED 6
--- 3RUNSdICC 2 - NOV 76 - C3LD SHUT 30dN I
9.11.4.3343 : DEFECTIVE FUEL CHECK VALVE O
--- 3RUNSWICC 2 - OCT 80 - 975 P0JER Gd
PAGE 11 9.11.8.3189 : DIRTf LINKAGE, 8INDING SREAKER INTERLOCK - RHR VALVE FAILED TO OPEN NATCH 1 - JUL 78 - 69E P3WER 3.11.8.3198 : 8 LOW 1 LPCI C3NTROL POWER FUSE MONTICELLO - NOV 78 - BOE POWER 0#
8.12.A.3060 : DIRT IN SE4WICE AIR AND PROCESS VALVES - R/A IN AIR SYSTE1
=== PEACH 50TTOM 2 & 3 - SEP 78 - $98 POWER (UNIT 2), 1303 P3WER g)
I (UNIT 3) 9.12.A.0066 : R/A IN BREAT4ING AIR SYSTEM C
--- PEACH 90TTOM 2 - SEP 78 --- SWR $ IN GENERAL ((
8.16.8.3188 : SINDING VACUJ1 BREAKERS ~
--- DRESDEN 2 - DEC 77 - 22X POWER, FEB 78 - $HUTDOW1 '
l 5.16.8.0190 : GROWING OF BJSHINGS - VACJUM BREACER SINDING
--- 10NTICELLO - FE8 78 - 103E POWER j' 3.14.8.3226 : VACUJM 3REAKER DESIGN DEFICIENCY a 3RUNSWICC 1 & 2 - NOV 78- APPROXIMATELY 100X POWER
( 3.16.8.3342 : EXCESSIVE CD1TAINMENT LEA (AGE THRU AIR COMPRESSOR DISCHAR3E #
1 CHECC VALVES j --- VERMONT YANKEE - AUG 78 - SHUTD0dN
- 8.16.8.0612
- VACUJM RELIEF SYSTEM POTENTIALLY 110PERABLE IN SWR $ d!TM 1 ARC I 3 E II CONTAIN1ENTS - BUTTERFLY VALVE CONTR3L SEPARATI31, SAFEff CLASSIFICATI3N gj
--- 3hR$ IN 3ENERAL - APR 85 --- PE ACH 80TTOM 2 5 3 - 1783 S.15.A.0051 : SNUB 3ERS FAILED T3 LOCC UP, 3-RING DETERI3 RATED
--- VERMONT YANKEE - JUL 76 - REFUELINC SMUTDOWN
/
8.15.A.0089 : DETERIORATED VALVE SEATS AND 0-RINGS, PACKING LEAKS, ETC. ' 60 2 DAD-CITIES 2 - JAN 78 - REFUELING 3.15.4.0244 : REACTOR SCRA1MED WHEN AIR C01 PRESS 3R FAILED d 3ROWNS FERRY 1, 2 5 3 - AUG 78 --- MONTICELLO - FE8 81 -
1303 POWER 3.16.C.0646 : THREE VALVES CLOSED - RCIC AND HPCI INOPERABLE 3RUNSWICC 2 - SEP 80 - SHGTD0d1 J
............. ALL DONE, PRESS.<r.ETURN) (EY TO G3 T3 MENJ .................?
( THERE ARE 138 ARTICLE (5) TO BE DISPLAYED: 19-AUG-1986 13:48:17 BWR - CHECK VALfES - 1951 - 1986 J 8.06.8.0169 : CRD HCU ACCU 1ULAT3R$ LOST PRESSURE - LEAKING VALVES
--- SUS 3UEHANNA 1 - AUG 82 - PREOPERATIONAL I 1
1 8.06.8.0213 : MANUAL $HUTD3dN - CR0 ACCJMULATOR IN3PERA8LE - CR0 PU1P C1ECE VALVE FAILED TO CLOSE - CRUD BUILDUP OI
- --- LASALLE 1 - JJN 85 - 72% POWER 8.05.A.0047
- REACTOR SCRA1 ON EXCESSIVE RCS LEACAGE - RECIRC PUMP SEAL C3 DETERIORATIO1, Lod WATER LEVEL - DEAR, SLOW RESPONSE OF FLOW
, VALVE POSITIJNERS GRAND GULF 1 - FE8 86 - 62X PDdE4 II
a PAGE 12
/
9.05.C.3452 : MSIVS & FW CHECK VALVES F4ILED LLRT - SEAT AND DISC WEAR
--- 4RNOLD - FE8 SS - REFUELING p 1 3.05.C.3455 : MSIV CHECK V4LVES SEATED IMPROPERLY 3N GRADUALLY DECREASING AIR PRESSURE - C1ECK WALVES REPL4CEO
--- g; SWR $ IN SENERAL - APR 85 ((P.36.D.3481) 3.05.C.3477 : 405- 3PERASILITY SRVS DEFECTIVE P3TENTIALLY COMPR31ISED - EXCES$1VE AIR LE4K4GE POWER SWR $ IN SENER4L - JUN 86 --- GRA1D GULF 1- MAR $6 - 80E
()
3.05.0.0056 : EXCESC FLOW CHECK VALVE 110PER48LE - ELECTRICAL AND MECH 41IC4L COMP 3NENT FAILURES -
--- NATCM 1 - JA1 $3 - REFUELIdG
- 8.05.D.0065 : UNIDENTIFIED LEAK 4GE INT 3 DRYWELL FL30R DRAIN PU1P - RECIRC '
PUMP SYPASS VALVE, VENT V4LVE, Fd CHECC VALVE LE4KAGE
--- ARNOLD - APR 84 - 103E P3WER l J 3.06.D.3039 : 1AIN---
STEAM LEAKASE TO AUX BOILER SYSTEM - RELEASE 3ROWNS FERRY 1 - DCT S1 - 97% P0 DER *
)
8.06.D.0047 : REACTOR TRIPS - DEH SYSTE1 PRESSURE TRANSIENTS - MS SfP4SS VALVE SEATINS IMPROPER, FLUID PRES $URE L0d
--- WNP May S4 - 18, 20I POWER 5.06.D.0052 : TUR91NE, REACTOR TRIPS - HIGH MOISTURE SEPARATOR DRAIN T41K LEVEL - TWO PMASE FLOW SURGE IN TUR$1NE CROSS-ARQUND PIPING gj
--- SUSQUEH41NA 1 - OCT 85 - 64% P3WER 9.05.E.0149 : FW--- CHECK VALVES LEAKED
- HATCH 2 - JAN 81 - REFUELING
/
j 3.06.E.3163 : Fd --- CHECK VALdES LEAKED - dQR1 SEAL RINGS b 1ILLSTONE 1 - OCT 82 - REFUELING -
l j
3.06.E.3165 : FW --- CHECK VALVE SEAT LEAKAGE - RESILIENT SEAL MFG DEFECT '
VERMONT f4NKEE - MAR 33 - REFUELING l 3.06.E.3166 : FW--- CHECK VALdE LEAKED - F41 LED 8041ET SEAL RING DRESDEN 2 - MAY 83 - 71% POWER 8.0$.E.0169 : FW CHECK VALdES LEAKED - SAPS ON PERIMETER OF DISC SEAL MATERIAL, ALIGNMENT PR08LEMS
--- LASALLE 1 - N3V 83 - SHUTDOWN 8.06.E.0170 : FW CHECK VALWE LEAKED - SEAL RING SE4 TING AREA CUT
--- DRESDEN 2 - N3V S3 - 65% POWER l J 8.06.E.0172 : FAILURE OF'A1CHOR DARLING SOFT SEAT VALVE SEALS - IMPROPER VULCAN! ZING
--- 3WR'S IN GENERAL - FE3 84! L454LLE 1 - AUG, NOW 851 #
INDIWIDUAL PLANTS AS LISTED
- 9.0$.E.0174
- F W, COMBUST!3LE GAS CONTR3L $ DRfWELL EQUIPMENT ; RAIN V4LdES E CEEDED LLRT LIMITS - MISALIGNMENT, PITTING, WE4R
--- LASALLE 1 -
FE8 $4 - SHUTDOWN
__._ f\
0 PAGE 13 8.06.E.01TF FW SPEED INC4 EASE PU1P DA1 AGED - DISCHARGE CNECK VALVE FAILED T3 CLOSE, SEARIN5 CVERHEAT
--- 11NE MILE PT. 1 - NOW 83
- 761 P3WER #
9.3$.E.0178 : INCREASED GENERATOR LOAD 1 APRM READINGS - STH STAGE FW HEATER FLOODED gj MATCH 1 - J AN 84
- 958 P3WER a.06.E.3185 : ADDITION TO WI.E.169 RE: FW CHECK VALVE SEAL LEACASE - C}
CORRECTIVE ACTIONS LASALLE 1 & 2 - Nov $3 - $NUTD3W1 O
8.06.E.3190 : Fd CHECK 8 RCIC STEAM HIG4 LEAKAGE RATES - BUSHI1G WEDGED VALVE 3 PEN, DISC 410 SEAT MISALIGN1ENT, DA1 AGE
--- LASALLE 1 - OCT 86 - COL 3 $HUTDOJN 9.06.E.3197 : REACTOR, LOW TUR81NE VACUUM TRIPS - HEATER SHELL RELIEF VALVE DISCHARGE RUPTURE DISCS SLEW - RELIEF VALWES LIFTED AFTER LP FW >
HEATER HIGH DATER LEVEL TRIP - POWER FLUCTUATIONS JNo NOV S4 - 95% POWER b
9.05.E.0199 : FW---C4ECK VALWE FAILED LLRT - SEAT DETERIORATED LASALLE 1 - MAR 85 - HOT SHUTD3W1 -
8.06.E.0200 : FW CHECK VALVE LEAKAGE - SEAL RING W3RN
--- DRE5 DEN 3 - APR 85 - SHUTDOWN 8.06.E.3213 : STOP CHECK VALVES IN AUX FW PUMP STEAM $UPPLY FAILED 3N L3W FL0d - DISC AS$E13LY DEGRADED - VI8 RATION
--- SWRS IN 3ENERAL - FEa 86 ((P.06.E.6303) IJ 3.06.E.0218 : REACTOR SCRA1, CONTAINMENT ISOLAT!3NS - Fd TRANSIENT DutI16 PUMP SWAP - TRANSIENT-INDJ'CE3 WATER HAMMER CAUSED DRAIN LINE g)
BREAK PEACH 90TTOM 2 - DEC 55 - 44I PodER '
, C) 8.07.A.3139 : RCIC TUR81NE EXHAUST CHECC VALVE 11TERNALS DAMAGED - TUR9INE
, EXHAUST BLOCCED, EXHAUST DIAPHRAG1 RJPTURED
--- HATCH 2 - NOV 81 - 965 POWER #
9.07.A.0149 : RCIC TUR8tNE EXHA'JST CHECC VALVES DA1 AGED DURING TESTINE 3N AUX STEAM - L0d 5fEAM FLOW #
$HOREHAM 1 - FE8 $2 - STARTUP TESTING 8.07.A.0150 : RCIC & HPCI TUR8INE EXHAUST CHECK VALVES DAMAGED - L3W STEAM FLOW SUSQUEHA1NA 1 & 2 - APR 81 - PRE-OP TESTING S.07.A.3152 : $ WING CHECK WALVE PROBLEMS SWR'S IN GENERAL - Jug 82
> J a.07.A.3153 : RCIC TURBINE EXHAUST CHECC VALVE DAMAGE EXPERIENCE REVIEW!D -
HPCI TURBINE EXHAU$T CHECC VALVES SUSPECT SWR'S IN GENERAL - JUL 82 00 9.07.A.3162 : RCIC TESTASLE CHECK WALVE, EQUALIIING VALVE FAILURES TO CLOSE -
INDICATION PROSLEMS, SPRI4G TENSID1 ADJUSTMENTS NEEDE) II LASALLE 1 - JUL $2 - 2X POWER (PRE-0P), AUG 82 - 3, 95 P3WEk
, (PRE-OP)I SEPT 82 - 2E P0 DER (PRE-3P) gg
S
, PAGE 14 s
S.07.8.0048 : 58LC CHECC V4LVE LEACED - SEAT REW3R(ED MILLSTONE 1 - OCT 82 - REFUELING J
3.07.C.3100 : CONTACTOR FAILED TO ACTUATE CSS VALVE OPE 1 CIRCUITRY 3ROWNS FERRY 2 - MAR 51 - 991 P0 DER SJ a.07.C.0117 : CORE SPRAY AND RdCU WELD CRACKS FOUN3 - IESCC
--- 3ROWNS FERRY l - AUG S3 - REFUELING
()
3.07.C.0118 : HPCS CHECK VALVE FAILED - SYPAS$ VALVE FAILED - 4CTU4T04 $PRING L335E
--- LASALLE - OCT 82 & JU1 83 - 23 & 421 POWER -
8.07.C.3132 : HPCS CHECK v4LVES F AILED TO PREVENT REVERSE FLOW FROM JP' 4* -
SCORED SEATING SURFACES - FM * '
] -- .ASALLE 1 - JJL S4 - PTE POWER ,
! 8.07.D.3291 : $UPPRESSION CHAM 3ER LEVEL HI - RNR P'JMP DISCHARGE CHECK V4LVE I
i FAILED 70 $E4T, OPERATOR FAILED T3 REALIIE A8 NORMAL C3NDITION
--- 3RUNSdICK 1 - DEC 81 - 03 POWER (STARTUP) 2 i
s.07.D.0297 : RHR PUMP DISCHARGE CHECK WALWE DISC LOOSE - DESIGN PR38LE1
, SUSPECTED *
--- FITZPATRICK - MAR 82 - 12% P0 DER (STARTUP) j 5.07.D.3300 : LPCI CHECK V4LVE DRAIN / EQUALIZING LINE WELD CRACK - IMADE3UATE PIPE SUPPORT "
--- DRESDEN 3 - APR 82 - REFUELING 9.07.D.3315 : RELE4SE TO I17AKE STRUCTURE - UNDETECTED LEAK PATH THRO'JG1 HP SW SYSTEM FR31 RHR HX PEACH 30TTOM 3 - Nov 52 - 89% PodER
' () '
3.07.D.3321 : 4HR PUMP DISCHARGE CHECK WALVE FAILED - DISC HANGER RET 4IMING
} CAP SCREWS MISSING, SHEARED - POSSIBLE VIBR4 TION 04M43E /
--- FITIPATRICK - FES 83 - 1303 P3 DER ,
O
, 9.07.D.3343 : RHR PUMP DISCH4RGE CHECK WALWE WOULD NOT SEAT PROPERLT - 3ISC
- MI$ ALIGNED, DISC HINGE PI4 BENT "#
SUSQUEHANN4 1 - AUG 83 a.07.0.0368 : RHR PUMP DISCHARGE CHECK WALVE FAILED To SEAT PROPERLY - SISC MISALIGNED, DISC HINGE PI4 BENT i
SUSQUEHAMMA 1 - NOV 53 - 1003 P0 DER l 8.07.0.0372 : RHR CHECK VALVE LEAKED EXCESSIVELY - IMPR3PER SE4 TINS -
}
MI5 ALIGNED HINGE ARM 1
SROWNS FERRY 3 - OCT $4 - REFUELING l
8.07.0.3375 : LPCI CHECK V4LVE F4ILED T3 H3LD PRESSURE - TIGHT P4CCIN3, SH4FT CORR 3$ ION '
] --- LASALLE 1 - OCT 84 - COLD $HUTD0dN 3.07.D.3406 : RCS NOT Is0L48LE FROM LP COOLANT SYSTEMS - PCIV LEAK 4SE
--- BWR5 IN 3ENERAL - JUN 86 --- 4R10LD - MAR 86 - SHUTD3WN ---
PILG41M - APR 86 - 93% P0 DER O
s.07.E.0235 : HPCI STEAM EtHAUST CHECK WALWE DISC PIN F4! LURE - DISC L33GED I , IN PIPING j --- MONTICELL3 - MAR 81 -
REFUELIN3 gg
is
. PAGE 15 J
8.37.E.3237 : HPCI TUR8INE EXHAUST CHECK VALVE LEACED P4ST SEATING RI4G TpAE4DS ,
- --- 10NTICELLO 1 - APR 81 - REFUELIN3 8.07.E.0252
- ADDITION SLOW 4 TO v!!.E.245 RE: HPCI GLAND SEAL CONDENSER NEAD 34SKET '
--- 3ROWNS FERRY 2 - SEP 51 - SHUTD0dN (F3LLOWIN3 SCRAM) g)
9.37.E.0275 : HP ANGLE I1JECTION STOP CHECK V4LVES DISC STUCK SHUT - INC3RRECT SEAT 3WR$ IN 3ENERAL - AUG 82 ((P.07.4.3753) ()
8.07.E.3284 : HPCI TURSINE EXH4JST FAN CHECK VALVE LEAKED - WEAR - DEFECTIVE VALVE SEAT #
--- 3RESDEN 3 - FE8 82 - REFUELING 5.07.E.3286 : HPCI PUMP TRIPPED DURING SCRAM REC 3VERY - CONTROL CIRCUITS
- WETTED - GLA1D SEAL C3NDE4SER GASKET F4ILED - OVERPRES$URE PILGRIM - AUG 82 - STARTUP 8.07.E.0311 : ALAR 15 DURIN3 HPCI SURVEILLANCE - VALVE PROBLEMS, C011U1ICATION ERRORS
--- PILGRIM - SEP 83 - 963 P3WER ~
8.07.E.0335 : HPCI TUR8INE EXHAUST RUPTJRE DISCS F4ILED - OVERPRESSJRE FR01 JATER IN EXH4UST LINE, SLUGGISH CHECC VALWES '
--- PEACH S3tTOM 3 - OCT 53, JAN 84 - 17, 100X POWER 3.07.E.3354 : HPCI ISOLAT!3N - EXHAUST LINE DIAPHR4GM RUPTURED - F1 BLOCKED I#
DRAIN LINE - PLANT SHUTD0dN - CHECK WALVE FAILURE - DEFECTIVE SEAL, DISC OJT OF ADJUSTMENT, LODSE MUTS, NICKED SEAT
--- HATCH 1 - JA1 85 - 933 P3WER g) 3.07.E.3360 : HPCI, RCIC TJRBINE VALVES NOT LOCAL LEAK RATE TESTED PR3PERLY - /
LEAKING SONNETS - PERSONNEL, PROCEDURAL ERR 3RS g)
--- 3ROWNS FERRY 1, 2, 3 - M4R 85 - 702 P3WER, REFUELINS, COLD SHUTDOWN
)
9.07.E.0377 : HPCI IN0PERA3LE - TESTABLE CHECK I1JECTION VALVE DISC P381 TION INDICATING SJITCH QUT OF A D J 'J S T M E N T COOPER - NOV 85 - 7I POWER (STARTUP) J
, 9.07.E.0385 : HPCI CHECK V4LVE INTERNALS D4MAGED, MISSI1G - CAP SCREWS 10T -
LOCKED IN PL4CE
--- SHOREHAM - NOW 85 - C3LD SHUTD3W1 4
8.07.1.uset : PRIMARY CONT 4INMENT VALVE LE4KS - 10RMAL DEAR s
--- COOPER - MAY 81 - REFUELING 8.07.F.0121 : VALVE LEAKS - FOREIG1 MATERIAL /0PERATOR ADJUSTMENT / FAILED SEAL / 1 J CORR 3SION NINE MILE PT 1 - APR $ MAY 81 - tEFUELING G
5.07.F.3136 : DRAG 3N EXCESS FL0d CHECK WALVES FAILED 70 FUNCTION IF LINE SAEAC WAS NOT NEAR THE VALVE j --- LASALLE 1 & 2 - SEP $1 - CONSTRUCTION PHASE ()
{ 5.07.F.0152 : VARI 3US V4LVE LEAKS - WEAR, FOREIGN 14TERIAL i --- FITIPATRICK - NQv 81 - REFutLI4G G3
e PAGE 16 o
3.0F.F.3169 : v44I3US ISOLATION VALVES FOU13 LEAKI1G DURING LLET - IMPR3PER SEATING DUE T3 DA1 AGED C39P01ENTS, CRU3 BJILDUP ,
--- PEACH 90ff0M 2 - JUN $2
- REFJELING S.07.F.0193 : VALVE LEAKAGE FOUND DURING LLRT "#
--- HATCH 2 - FEB 82 - REFUELING 3.07.F.0196 : EXCESS FLOW CHECK VALVE FAILED TO SEAT - CRUD
--- 3RUNSWICC 1 - OCT 82 - 54U70041 2.37.F.3210 : LLRT LEAK 43E - N3494L WEAR
--- FITIPATRICK - JUN 83 - REFUELING S.07.F.0211 : CIVS LEAKED EXCESSIVELY - VARIOUS CA3SES
--- LASALLE 1
- AUG S3 - JI POWER ($10TD0dN)
- 3.07.F.3215 : OFP RAN BACKJARDS - ENGINE DAMAGED - CHUNC 0F WQ3D L3DGED IN DISCHARGE CHECK VALVE
--- LASALLE 1 8 2 - DEC $3 - DI P3 DER (STARTUP PRE 3PERAT!3NAL) a.07.F.0227 : VARIJUS VALVES LEAKED EXCESSIVELY j
--- 90NTICELLO - SEP S OCT 82 - REFUELING ~
< o 8.07.F.0243 : LLRT LEAKAGE IN EACESS OF TECH SPEC - NOR1AL WEAR - VALVE MODIFICATIONS
--- QUAD-CITIES 1 - 1AR 86 - COLD SHJTDOW1 #
3.07.F.3256 : RHR, MSL, TIP, HPCI, LPCI, CS 4 OTHER CIVS FAILED LEAC RATE TESTING - NORMAL JEAR g,
--- HATCH 1
- OCT-DEC 84 - COLD SHJT30WN 9.07.F.3257 : EXCESSIVE CId LEAKAGE RATES
--- PEACN SOTTOM 2 - JAN SS - REFUELING 9.07.F.3266 : PCIVS FAILED LEAK RATE TESTS - VARIOJS C4JSES l PILGRIM - DEC 83 - REFUELING ,
8.07.F.3271 : CORROSION FAILURES OF 410 SS VALVE STEMS - IMPROPER HEAT l TREATMENT SWR $ IN 3ENERAL - JUL 85 --- SRU1SWICC 2 - AUG 82 --- BR3dNS FERRY 3 - FE3 84 3.07.F.3291 : VARI 3US CIVS FAILED LLRT - N3RMAL WEAR, INTERNALS MISSIMG FROM 4 SYPASS VALVE
--- dATCH 2 - APR 85 - REFUELING 8.07.F.0304 : ESF ACTUATIO15, CHARGING JATER SPILLAGE - ORYWELL PRES $3RE SWITCHES SHORTED - MOISTURE IN JUNCTION 83XES FR3M EARLIER INADVERTENT FIRE PROTECTION ACTUAT!01
--- 3R0dNS FERRY f - APR-MAT 86 - C3LD SHUTDOWN I #
l 3.07.F.3305#: VALVES EXCEE3ED LLRT LIMITS - NORM 4L WEAR 3UT, TORQUE SWITCH ADJUSTMENTS 1 FAILURE, VALVE DESIG1 00 PEACH 80FTOM 3 - SEP $5 - SHUTD0dN 3.DT.e.JS06e: SSGT FILTER TRAIN IN3PERA3LE - CHARC3AL FILTER BED WET - II LEAKING DELU3E VALVE 9 ASKED Bf LEAKING DRIP CHECK VALVE
--- HATCH 1 1 2 - NOV 85-14R 86 - REFUELING, 851 P3WER
. . . . .--- .-- _ - - - - - - - =-
I
o PAGE IT
)
3.05.C.3236 : FAILEDSW PJMPSMAFT PACKING AND SEAL WATER SUPPLY CHECK VALVE FOR RHR MONTICELLO - 1AR 81 - 943 POWER o S.31.C.3245 : RMR --- SW PUMP 3ISCMARGE- CHECK WALVE FAILED
- WEAR MATCM 1 - JUL 81 93E P3WEA ss S.38.C.3281 : RHR SW PUMP 3!$ CHARGE CHECK WALVE ST3CK OPEN - DISC NJT L30$E, C3TTER PIN MISSING
--- FITIPATRICK - MAY 82 - 1303 P3 DER (3 8.09.C.3285 : RHR SW PUMP 4EAD LOW, DISCHARGE CHECC VALWE LEAKAGE - WEAR, SET SCREd ERODED ARNOLD - MAY 82 - 90% P0 DER, AUG 82 - 51% P0 DER 3.05.C.3288 : RMR SW PU9P 1EAD AND FLOW L0d - CHECC VALVE STUCC OPE 4 IN DISCHARGE FL3W PATH
--- HATCH 2 - AUG 82 - 965 P3WER J
3.05.C.3303 : ESW CHECK VALWE INTERNALS WORN, MIS $!NG - ACCELERATED R3N TIME
--- SUSQUEHA4NA 1 - APR S3 - COLD SHJTDOW4 (STARTUP TESTItG) ,
8.31.C.3311 : POTENTIAL DA1 AGE TO REDUNDANT SAFETY EQUIPMENT FROM 3AC(FLOW THROJGH EQUIP 9ENT AND FL33R DRAIN SYSTEMS 3We5 IN SEMEAAL - JUL 83 ((P.39.C.1353) 9.08.C.3331 : RMR$d PUMPDEAR STUD THREA3 DISCHARGE CHECC VALVE FAILED T3 SEAT PROPERLY - DISC '
--- MONTICELLO - DEC 83 - 795 POWER j
8.05.C.3334 : SW---PJMP VALVE FAILURES 4>
HATCH 2 1 1 - OCT 82, QCT 83 - 55, 70E POWER 3.0S.C.3335 : ADDITION
--- TO WIII.C.338 RE: LOST PACIFIC CHECK VALVE MUT o SUSQUEHAMMA 1 - APR $1 - COLD SHJTD0d4 (STARTUP TEST!1G) '
a.09.C.3338 : RHR O
--- SW VAULT SUMP DISCHARGE CHECK WALVE LEAKAGE - FM 8UIL3U*
! 3U40-CITIES 2 - DEC S3 - REFUELING
] 3.05.C.0339 : HPSd7 DISCHARGE CHECC VALWE STUCC 3 PEN - VALVE DISC PIN, ARM J i INTERNAL WEAR t
PEACH BOTTOM 2 - DEC 83 - 97X PodER '
3.05.C.3346 : POTE1TIAL LOSS OF ESW - FAILURE OF OTPASS 04 CHECK VALVES -
FLOW DETECTOR INSTALLED SUSQUENA1NA 1 - OCT 82 - SHUTD3W1 (PRE 0P TESTING)
S.08.C.0358 : PLANT SHUTD0JN - DRYWELL FLO3R DRAIN SUMP PUMPS FAILED, SJMPS J
DVERFLOWED - OROKEN DRIVE COUPLINGS, FAULTY FW CHECK WALVE HINGE PIN HATCH 2 - DEC 84 - 995 P3WER I J
8.09.C.0134 : RCS INSTRUMENTATION LINE EFCWS FAILED TO CHECK FLOW I4ITIALLY
--- MATCH 2 - MAY 83 - REFUELING b
I I 8.09.E.3634 : WORN INTERNALS ON INSTRU1ENT LINE EXCESS FLOW CHECC WALVES i
--- 3ROWNS FERRY 3 - JAN 51 - REFUELING $NUTDOWN f)
- j. S.09.E.0681 : INSTRUMENT LINE EXCESS FL3W CHECK WALVES FAILED aROWNS rERRr 1 - SEP si - REruELING GP
e PA3E 18 a.07.E.3866 : EXCESS FLod CHECC VALVE 3*ER481LITT 10T VERIFIABLE WITH30T REM 0 DING VESSEL HEAD - DESIG1 DEFICIENCY SUS 3UEH4444 1 - AUG SS - DE P3 DER (PREOP TESTING) 8.09.E.3900 : MSL INSTRUME1T LINE VALVES INDPER43LE - INSUFFICIENT FL3W, FAULTY POSITION Sd!TCHES 4RNCLD - FES $3 - SHUTD0dN 3.09.E.1033 : RPS, ATWS TRIP SIGNALS - INCORRECT EXCESS FLOW CHECK VALVE O OPENED DURIN3 TEST - PER$3NNEL ERR 3R 3UAD-CITIES 1 - 94R $$ - REFUELING ()
5.09.G.3376 : RCLD4SS CHECC VALVES STUCC $1UT - FOREIGN MATERIAL
--- PILGRIM - JUN 81 - 133% POWER 8.39.G.0391 : INSTRUMENT LINE EXCESS FL3W CHECK VALVES FAILED - VALVES REPL4CED
--- NINE MILE PT 1 - JUN S1 - REFUELING 8.09.G.3551 : ADS INSTRUMENT PRESSURE L3W - AIR DRVER UNIT SOLENDID'V4LdES STUCE, h!TR03EN SUPPLY CHECK VALVE LEAKIN3 - MODIFICATI3NS RE3u! RED 1
--- LASALLE 1 - DEC 82 - 19E POWER ESTARTUP TESTING) J 9.11.A.3412 : DGS TRIPPED 34 HIGH C00LI4G d4TER TEMP - CAUSE UNKN0 dis C30 LING W4TER PUMP DISCHARGE CHECC V4LVE SUSSEQUEMTLY REPLACED 3RESDEN 3 - OCT 4 NOV 81 -
66 & $0% P3WER l
3.11.A.3430 : ADDITION T3 XI.A.412 RE: DGS TRIPPED ON HIGH C00LIMG d4TER TEMP gj
--- DRESDEN 2 8 3 - DEC $1 --- 8dR$ IN GENERAL ((
3.11. A .04 7 8 : EMER3ENCY 05 STARTED SLod - FAILED FUEL OIL DAY TANK RETURN I LINE CHECK V4LVE
--- WERMONT f4NKEE - OCT S2 - 1005 P3WER ,
4)
S.11.4.3490 : CHECC VALVE FAILURES IN 03 R4W W4TER COOLING SYSTE15 - IERNAL DISASSEMBLY - ABRASION, C3RR3SION L 2 ((
SWR $ IN 3ENERAL - MAR 83 --- DRESDEN 2 8 3 --- QJ40-CITIES 1 '#
5.11.A.0521 : DG ---MECHANICAL PR33LE15 REVIEdED
)
SWR $ IN GENER AL - AUG 83 --- BRU4SWICC 1 --- CLI1T04 1.4 2 --
- DRESDEN 3 --- FERMI 2 --- JUAD-CITIES 2 --- SH3REH41 **-
SUSQJEHANNA 1 83 --- OTHERS AS NOTED IN '
i a.11.4.3566 : DG PROBLEMS - START FAILURE, VOLTA 3E COULD NOT BE MAN 3 ALLY ADJUSTED - FJEL OIL RETUR1 LINE CHECC VALdE LEAKING, LOCA*. M30E J
' SELECT SWITC1 CONTACTS DIRTY HATCH 2 - OCT 83 - 1031 POWER I
J 8.11.A.3641 : DG M4LFUNCTI31 - PIN BROCE JFF AIR START CHECK VALVE S L33GED i
IN A1 AIR INLET VALVE 3 RAND 30.F 1 - MAR 85 - SHUTD3dN d) 9.11.8.0526 : LOSP -
TRANSFORMER F4ILURE - CABLE F4 ULT - INSULAT101 DEGRADED -
4GE, ENVIRO 11 ENTAL DAMA3E I)
--- 3WRS I4 3ENERAL - JUN 86 (CP.11.3.861))
3.12.A.0052 : RADWASTE SERd!CE AIR PIPING CLOGGED -
VENT LINE SLOC(ED -
i, 67f
I PAGE 19 UNM0gITORED RELEASE DVSTER CREEK - JUN 82 - 84$ P0 DER 9.12.A.3108 : UNM04!TORED CST SJMP DISCHARGE - RADIOCHEMISTRY NOT 11F3R1ED 8EFORE MAINTENANCE BEGAN
--- SHOREMAM - APR 86 - C3LD SHUTD3d1 3.14.A.3161 : EXCES$1VE PE1ETRATION AND VALVE LEAK 4GE - VALVES WOR 1, D414GED -
WARICUS CAUSES g)
QUAD-CITIES 2 - MAR 85 - REFUELING S.14.8.3347 : ISOLATION CHECK VALVE LEA (ED WORN 3-RING NINE MILE PT 1 - MAR 81 - REFUELING S.14.8.3458 : REACTOR BUIL3ING PRESSURE Lod - VE1T SYSTEM SUPPLY F41 F4ILED -
BLADES BRO <E 3FF LASALLE 1 - OCT 82 - 32% POWER (STARTUP TESTING)
)
8.14.8.3572 : EXCEST FL0d CHECC VALVE FAILJRES - VALVES INSTALLED 11 dR3NG LINES - INCORRECT MFR NAME TAGS FDLLOWING MODIFICATI04 JNP FEB $4 - PREOP TESTIN3 '
3.15.4.3246 : CHECC VALVE FAILURES - MF3 DEFECTS
--- 3WRS I4 3ENERAL - DEC 81 ((P.15.373))
8.15.A.3258 : ADDITION TO Rd.244 RE: AIR C3MPRES$0R FAILED, REACTOR SCR4MMED 90r.'ICfLLO - FEB 81 - 103% P3 DER 5.15.4.3300 : POTENTIAL FOR PIPING SYSTEM $ PIPE SUPPORT DAMAGE - PIPE BEhDING - TE1P DIFFERENCES INDUCED BY STR4TIFIED FL0d gj 3ha$ IN 3ENERAL - DEC 84 --- 41P AUG 84 - 11 P0dEt (STARTUP) ((
()
8.15.A.0302 : CONTAMINATION OF SREATHING AIR SYSTE13 - AIR COMPRES$3R$ FAILED, RADIDACTIWE GAS S FM IN LIR SURGE AT SYSTEM STARTUP /
3ROWNS FERRY 1, 2 & 3 - SEP 84 - VARIOUS POWER LEVELS --- ,
b0 SdR$ IN GENE 4AL - JAN 85 ((P.15.455))
3.15.A.3310 : LIMITORQUE ACTUATOR d3R1 SHAFT GEAR FAILURES - IMP 4CT L34)!NG #
OF CLUTCH MECHANISM DURIN3 M3DE CN4NGES AT HIGH SPEED 3dRS IN 3ENERAL - AUG 85 --- 3THER UNITS AS LISTED IN ARTICLE (( J S.19.C.3724 : VENT ACCUMUL4 TOR CHECC VALVES LEAKED - IN4DEQUATE PM
--- 10NTICELLO - MAY $1 - REFUELIN3 2 9.16.C.3748 10CFR$0.44 AMD TMI LESSO1S LEARNED R!QUIREMENTS CONFUSED # REP 3RTING REQu!REME4TS OVERLOOKED - MANAGEME1T PROBLEM >
--- PILGRIM - JUN 81 - 1004 P0 DER 8.16.C.3757 : RdCu SYSTEM RETUR1ED TO SERVICE WITH ISCL4 TION VALVE IN3P!R48LE I J TECH SPEC aEQUIREMENTS TO 3E CHA1GED MONTICELLO - JUN 81 - 103% POWER S
, 8.15.C.3534 : SURVEILLANCE MISSED - ADMIN AND PR3CEDURAL CONTROL BREACD3dM i ,
3fSTER C4EEK - MAY B1 To FEB B2
- VARIOUS
! 3.16.C.3935 : NONIDENTICAL REPLACEMENT PARFS - RECURRING PROBLEM
, --- SdR$ IN *ENERAL - FEB 83 ((P.15.C.184S3) gg
i PAGE 23 3.16.C.1073 : CECW ---
CHECK V4LVE INSTALLED 84CKWARDS 3R3m%5 FERRY I - 3CT S3 - REFUELING o
5.16.C.1109 : REACTOR WATE4 LEWEL DECRE45ED - LPCI CHECK VALVE STUCC 3 PEN, PACKING GLAN3 TIGHT - IMPROPER M41NTENANCE
--- LASALLE 1 - SEP 83 - COLD $HUTDOJN gj 5.15.C.1241 : EXCESS FLod CNECK VALVES NO TESTED - ADMI1, PER$0NNEL: ERR 3R
--- 3YSTER CREEK - JUN 84 - REFUELING ()
3.15.C.1271 : CORE $ PRAY PIPING OVERPRE55U4IIED, P4 INT SA14GED - W3RKERS CONT 4MINATED - CPERATOR TESTING, 141NTENANCE ERR 3RS ()
--- 3R0=NS FERRY 1 - AUG S4 - 1035 P3 DER 9.15.C.1319 : INADWERTENT 5AFEff SYSTEM STARTS DURING TESTING - INSTRJMINTS #
NOT ISOLATED
--- 3ROWhS FERRY 3 - SEP S4 - REFUTLING 3.16.C.1437 : EXCE55IVE DRfWELL LEACAGE - TEMPOR4Rf NOSE SLIPPED OFF FITTING -
WALWE SOCKET WELD CRACK FOU1D LATER - VI5 RATION
--- 3ROWNS FERRY 1 - JAN SS - 01 P3WER (STARTUP) 8.15.C.1461 : P3fENTIAL VI3LATI3N 3F PRIMA 4Y CONTAINMENT - OPEN FL0d CNECK VALVE BYPAS$ VALVE - PER$3NNEL ERR 3R JNP FEB 85 - 1001 PodER 3.16.C.1506 : EXCESS FL0d CHECK VALWES INOPERABLE DURIN3 STARTUP - SYPASS #
VALVES LEFT 3 PEN LIMERICK 1 - APR 85 - HOT SHUTD0dN D
8.16.C.1545 : REACTOR, TUR3INE TRIP DN LOW CONDE15ER VACUUM - SREACER INADWERTENTLY TRIPPED, LU3E d4TER PU1P CHECK VALVE LE4KED 3 RAND GULF 1 - MAY 85 - 741 P3 DER p) 5.16.C.1580 : REACTOR HP SCRAM $1GNAL - CRD FL3d 11CRE4 SED WITH 403T W4'VE . #
CLO5ED - CHECK VALVE LEAK 4GE gg PEACH SOTTOM 2 - MAY 85 - COLD $1UTD0dN
............. ALL DONE, PRESS < RETURN) (ET TO GD T3 MENU .................?: #
- JTHERE ARE 81 ARTICLE (5) 70 3E DISPLATED
- 19-AUG-1986 13:55 23 PWR - CHECC V4LVES - 1950-1978 P.05.F.3004 : PR09LEMS WITd VALVE OPERATIN3 SYSTEM SHIPPINGPORT - 1750's P.05.F.0016 : CHECC VALVE INTERNALS REPLACED WITH 10RE DURA 8LE M4TERI4LS
--- YANKEE R3dE - AUG 74 - REFUELING $HUTD0dN 3
P.06.A.0014 : CONTROL & ST3P VALVE PR38LEM5
--- CONN YANCEE CHADD4M NECK) - 1967, 68, 69 i ,e P.06.D.3025 : VALVE STEM C3RROS!0N & BI1DI1G
, --- CONN. YANKEE (HADDAM NECC) - 1767, 68 g
P.05.E.0022 : CHEC(VALVE G45KET FAILURE - JATER HAMMER, 8 LOWDOWN, 54FEff INJECTICN, PIPE SUPP3RT 04 MAGE
--- SURRY - 3CT T2 (P3WER ESCALATION TESTING)
P.05.E.3329 : CASTING DEFECT IN CHECE V4LVE
--- SuRRY Z - MAY T3 f)
.n.. -
)
PAGE 21 J
P.06.E.0055 : NIGH --- ENERGY LINE MODIFICATIONS NEGATED AUE. FW REDUNDANCY SURRY 1 1 2 - AUG 74 I
' P.05.E.0064 : RELEa5E THROJGN FEED PUMP $EALS .
PT. SEAC1 - MAR 75 - COLD $NUTD0dN P.03.E.3106 : Fd CHECK V ALWE GASCET LEACED
--- PALISADES - PRIOR TO 1974 I)
P.06.E.0119 : EXCE551VE Fd FLod CAUSED d4TER HA41ER - $1USOERS & PIPE
$UPP3RTS DAMAGED COOK 1 - JAN 76 & MAR 77 - $NJTD3WN ()
P.06.E.3128 : CHECC VALVE FAILED - BURA ON HINGE - PUMP CA$1NG GASEET RUPTURED - '
REACT 3R TRIPPED ROSINSON 2 - AUG 77 - 103E POWER P.36.E.0142 : FAULTY VALVE TORDUE SdITCH REPLACED
)
DAVIS-8E5SE STAND 8Y 1 - DCT 77 (POWER ESCALATION TESTING) - N3T P.06.E.0153 : FW REGULATIN3 VALVE FAILED OPEN - TUSING FITTING BROCE .RCP SEALS DAMAGED FT. CALH3ON 1 - APR 74 - 90% P3 DER P.35.E.3155 : FW---CHECK VALWE PIN BR3CE SAN ONOFRE 1 - APR 76 - STARTUP P.0$.E.0162 : LOOSE 8'ONNET BOLTS, SCRATCHED GASCET - AUX FW VALVES LEACED
--- DAVIS-3 ESSE - JUL 77 - $NUTD0d1 - APR 78 - 70% P3 DER P.0$.E.0171 : CONDENSATE STORAGE TANK LEVEL LOW - VALVING ERROR, CHECC WALVE STuCC
--- TROJAN - JUL 78
- HOT STAND 8Y #
s 4)
P.36.E.31S$ 2 Aux Fd CHEtt VALVE ORIENTED dRONG - PLATE NINGE LUGS MIS $1NG -
PLATE RECOVERED FROM SG
--- FT. CALN30N 1 - N3V 75 - REFUELING P.05.E.0353 : AUX---FW PU9P PROBLEMS - VALVE LEACS DEFECTIVE FUSE N0LDER ROBINSON 2 - DEC 77, JUN S1 - 103% P0 DER, STARTING JP ,
P.06.F.3069 : WORN, CORRODED PU9P5 - EXTENSIVE REPAIRS - VALVES INSTALLED IN LUBE WATER PIPING ST. LUCIE 1 - 1977 P.07.A.0013 : CHECC VALVE ASSE99 LED INC3RRECTLY PALISADES - MAT 72 '
P.07.4.0020 : INDICATIONS IN VALVES '
J 94INE YAMEEE - SUMMER 1972 (AFTER HOT FUNCTI3NAL TESTING)
P.37.4.0025 : WELD $ LAG UNDER VALVE SEAT - SI TA1K CONCENTRATION L3d OI
--- MAINE YAMEEE - DEC 72 P.07.A.0032 : CNECC VALVE SOFT SEAT FAILURES
--- TURKET PT 4 - MAT 73
(
P.07.4.0060 : CRACC INDICATIONS IN VELAN CNECK VALVES
8
. Pagg 22 CALVERT CLIFFS 1 - SEP 73 (FUNCTIONAL TESTING)
P.07.4.0063 : CHEC( VALVE LEAKAGE - ACCJMULATOR DILUTIQ4 3 INN 4 - SEP 76 P.07.A.3055 : CHECK WALVES DID NOT SEAT, PROCEDURAL PROSLEM - 80404 CONCENTRATIO4 LOW
--- SURay 1 - AUG 75 - HOT STAND 8V P.07.4.3105 : LEAKING CHECC VALVE -
ACCUMULATOR 80RON L3W
--- 40SINSON 2 - JAN 76 - 103% P3WER gg P.07.A.3114 : LEAKING CHECC VALVES CAUSED SORON DILUTIO4 IN SI ACCd4ULATOR SURRY 1 - JUL 76 - 133% POWER A
P.07.4.3120 : 80401 DILUTI3N IN SI 4CCU1ULATOR C4USED 8f LEA' KING CHECK JALVES SURRY 2 - AUG 76 - 133% POWER P.07.A.0122 : SACK LEAKAGE IN ACCU 13Li'3R DISCHARGE LINE CHECK V4LVES ZION 1 - JUN 76 - 5Gs PodER D
P.07.A.3126 : WRON5 SIZE G4SKET INSTALLED IN ACCUM'JLATOR CHECK VALVE
--- ZION 2 - OCT 75 - 30% PodER I A
P.07.A.3143 : IMPR3PER SEATING OF SIT DUTLET VALVES CAUSED DILUTION OF 54T'S ZION 2 - MAR 77 - 503 P0 DER, APR 77 - 48% PodER 2
P.3 7. A. 3172 80LTS & BRACKET FOR CHECK VALVE DISC MIS $1NG
=
INDIAN PT. 2 - MAY 78 (L3d P3 DER PHYSICS TESTS) gg P.37.A.3175 : TILTING DISC CHECK VALVE INSTALLED I4 WRO1G ATTITUDE SAN CN3FRE 1 -
MAY 78 - APPR3XIM4TELY 55% P0 DER gg P.37.A.0152 90R04 CONCfMTR4 TION LOW - SI TANK CHECK V4LVES LEAKED . l CAL %ini CLIFFS 2 - SEP 73 - 985 POWER '
s @;
P.07.A.3187 : STOP CHECK DISC SPRING JAMMED - SPRING RETAINER ADDED ARKANSAS 3NE 2 - JUL 78 - PREOP TESTING
~
P.07.A.0225 t SI T4NK 80R31 CONCENTRATI3N LOW - CHECK VALVE LE4KED .
MILLSTONE 2 - APR 77 - 971 POWER P.07.8.0012 : CRAC(ED ROCKIH4FTS AN3 CRACKED 80Df:IN CHECK VALVES SURRY 1 1 2 - NOW 74 l 1
P.07.8.0053 : CTS 10Z2LES LEAKED - VALVES NOT $ NUT TIGHTLY - CHECK VALVES
~
INSTALLED 3ACKWARDS C00K 2 - Nov 18 - SHUTD0dN "I
~ P.07.E.3020 : HYDR 3 GEN DILJTION BL0 DER CHECK VALVES. STUCK q i
--- DAVIS-8 ESSE 1 - APR TS - 58E POWER dI l
P.07.E.0026 : CONTAINMENT #ACUU1 RELIEF VALVE H4D 3EFECTIVE WELD, L30SE FLAN 3E BOLT - DIRTY VALVE SEAT gl l
--- DAVIS-GESSE 1 - MAY 75 - SHUTDOW1 P.07.E.3028 : CHECK VALVES LEAKED - DISCS S SE4TS CLEANED, LAPPED C DAVIS- BESSE 1 - JUN 78 - REFUELING P.07.E.0058 : ADD!t!ON YO ITEM WII.E.25 RE: LEAKIN3 CONTAINMENT VACUU1 RELIEF h@ jl
u PAGE 23 WALVE 3AV!5-8 ESSE 1 - MAY 8 JU1 78 - REFUELING
]
P.37.E.0237 : CORR 3SION FAILURES OF 413 $$ VALVE STEMS - IMPROPER HEAT TREATMENT FEB PWR$ IN 3ENERAL - JUL 85 --- DC01EE 1 - DEC 71 --- FARLET 1 -
84 ((5.37.F.2713) 0#
P.03.4.3035 : KER0---
TEST V4LWE5 INASEQUATE IN SAMPLI1G SYSTEM TURKEY PT. 3 & 4 - IST 1/2 0F 1973 P.05.A.3153 50DIu9 $ULFATE ACID SYSTE1 C3MPONENTS CORRODED, PUMP SE4L WORN C>
--- THREE MILE IS. 1 - 1915 P.05.A.3230 : CHECC VALVE LEAKED - RELIEF VALVES, FLOW TRANSMITTERS 0414GED -
CHECK VALVE REWORCED
' --- DAVIS-3E55E 1 - JUL 77 (AFTER INITIAL FUEL LOADI45) - H0T STAND 8Y P.38.A.3262 : 14KEUP $Y$fE1 VALVES LEATED R/4 G45 DURING VALVE INSTALLATI01 DN h!TROGE1 1EADER - RELEASE J
--- CRYSTAL RIVER 3 - FE8 78 - HOT STANDST j
P.05.A.3504 : VCT VALVES L!AKED TO ATM3 SPHERE - RELEASE #
--- FT. CALH3UN 1
- LPR 77 - 931 P3WERa SEPT 78 - 98E P3d!R l
P.38.A.3325 : CHECC VALVE SPRI1G PRES $URE EXCES$1VE - $PRING RETAINER MODIFIED - DE4VER VALLEY 1 - OCT TS - SHUTDOWN P.05.8.3C25 : SCALE HELD C4ECK VALVE OPEN E#
--- CONN YANCEE (HADD4M NECK) - APR T4 P.05.8.0025 : VELA---
1 STOP CNECKS STUCK CLOSED - SEATE REDESIGNED RANCHO SECO - 1974 (H37 FUNCTI3N4L TESTING) i P.3S.S.0043 : DEBRIS IN CHECK V4LVE CAUSED PUMP TRIPS CRACKED PU1P SEAL - bI 4081N50N 2 - JAN 75 - 103E PodER P.05.8.0079 : UNPLANNED RELEASE - RELIEF V4LVE LIFTED DUE TO CHECK VALVE LEAK THREE MILE IS. 1 - SEPT 75 - SuSCRITICAL P.04.8.3082 : DIRT---
ACCUMut4 TION - CHECC VALVE FAILED TO SEAT PROPERLY PT. BEAC4 2 - MAR 76 - REFUELING SHUTDOWN P.08.8.0090 : DECAT HEAT PJ9P M3 TOR SHAFT SHEARED
--- THREE MILE IS. 1 - M4Y & JUN F5 - 20E POWER I
P.05.8.0095 : SERVICE WATER BOOSTER Pu1PS TRIPPE3 DURIN3 AQUTINE SHIFT!1G '
l --- 4081N50N 2 - 3CT 75 - 103% POWER
' P.05.8.3118 : SERVICE WATER PUM* DISCHARGE CHECC V4LVE STUCK - LU8RIC4 TION ' #
PALISADES - FE9 77 - 100E POWER P.05.8.3127 : PUMP DISCHAR3E CHECK VALVE JAM 1ED - TRAVEL STOP READJUSTED - 00 REDUNDANT PU1P STARTED 01 WR3NG 4CV 305 1
, --- CALVERT CLIFFS 1 - APR 77 - 103E POWER O
I P.05.8.3150 : PIPING DESIG1 ERROR, RNR PUMP NPSN M4RGIN4L - VALVES THRDTTLEDs
, ORIFICES INSTALLED
--- FARLEY 1 - JUL FT (AFTER INITIAL FUEL LDADING) gg
e
- 4GE 24 4
P.08.8.3151 : CMECC ===
VALVE STUCK - FLAPPER MINGE *!1 CORRODED PALISASES - SEPT 77 - 103% POWER j P.39.8.3132 : $ERWICE WATER SYSTEM CHECC VALVE RUSTED DAVIS-SES$E 1 - MAY 75 - REFUELI1G gj P.05.8.3183 : SERWICE WATER CHECK VALVE HI4GE PIM CORRODED
--- PALISADES - APR 78 - 70% POWER ()
P.35.8.3187 : DECAT HEAT FLOW RATE REDUCED BELod TECH $PEC LIMIT T3 REPAIR VALVE SONNET LEAK ()
--- DAVIS-8 ESSE 1 - JUN 75 - REFUELI1G P.08.8.3195 : LEAKS FOUND 11 $ PENT FUEL POOL C03 LING SYSTEM
--- ARKANSAS ONE 1 - MAY & N3V 77 - 70% P3WER & COLD SHJTDOWN P.05.8.3293 : THREADED LOCCING DEVICES 3N VALVES, PUMPS, VALVE OPERAT3R$
FAILED TURKEY PT. 3 - OCT 79 --- DAVIS-SESSE 1 - OCT 78 TO OCT 79 --
- $ALEM 1 - 3CT 79 --- ARCANSAS ONE 2 - FE8 80 --- Pdt$ I1 ',
GENERAL P.09.8.3346 : CRANE CHECK WALVE DISC PIWOT PINS 40RKED LODSE
--- 43RTM AN1A 1 - OCT 77 (C3NSTRUCTION)
P.05.C.3304 : H3SE SLIPPED JFF TEMPORARf FITTING - RELEASE
--- $URRY 1 - AuG 74 - 753 MWE P.07.M.3044 : CHECC VALVE FAILED, 40RN *EALS IN SETECTOR - RELEASE j 1AINE T44KEE - APR 75 - 70E POJER P.07.M.0117 : INSTRUMENT AIR C09 PRES $0R DISCHARGE CHECK VALVE STUCK, 835$18LE RCP SEAL DAMAGE - VALWES REPLACiD, AJTOMATIC 8ACKUP ADDED ST. LUCIE 1 - APR 77 - 1331 P3 DER ',
gg P.11.A.3122 : DIESEL GENERATOR, OIL $UPPLY CHECK VALVES IMPROPERLY INSTALLED IN DRAIN LINE
--- INDIAN PT. 3 - OCT 75 #
1 P.11.A.3209 : COMPRESSOR HEAD GASKET 8LEW - TANK CHECK VALVES LEAKED IN DG AIR START SYSTEM 1 ---
KEWAUNEE - OCT 77 - 100% POWER l
)
P.11.A.0236 : DG AIR STARTING CHECK VALWE 8ROKE - LINE GASKETS BLED OJT
--- COOK 2 - MAR 78 - 0% #0WER (AFTER INITIAL CRITICALITTI P.11.4.3268 : FUEL INJECTIJN PU9PS, INJECT 3R FAILED - PISTON BOLT L30$E J i
COOK 2 - SEPT 78 - SH3fD3dN & 975 POWER P.11.A.3269 : DG AIR START RESEAVOIR PRES $URE L3d - AIR COMPRES$0R RELIEFS #
CHECC VALVES LEAKED
--- FARLET 1 - SEPT 78 - 1004 POWER 9
P.11.8.0213 : THER9AL OVERLDADS ADJUSTED TO PROPER SETTING - CHECK dALVE LEAKED R08tN$0N 2 - JUL 77 - 103% P3 DER P.12.A.0064 : LEAKING CHECC VALdES. IMPROPERLY R3UTED LINES - UNREP3 RTE)
Tatitum RELEASES gg
. . .._ -_ .-- x - - - --- -
PACE 25 IION 1 & 2 - 3CT 76 P.16.4.3358 : CONT 4INMENT ISCLATIO1 WALWES & AIRLOCK LE4KED - VALVE SEATS '
I REPLACED, F04EIG4 MATERIAL FLUSHED, LOOSE STEM SEAL PLATE TIGHTENED
--- PRAIRI! 15. 1 - MAR 77 - REFUELING P.14.8.3078 : WALVES LEACED EXCESSIVELY - SERVICE DATER CHECK VALVES DIRTY, '
PJRGE ISCLATION VALWE OPERAT3RS ADJUSTED g)
--- COOK 1 - FES ?? - REFJELING P.15.A.0373 : CHEC( ---
VALVE FAILURES - MF3 DEFECTS THREE MILE 15. - FEB B0 --- CRYSTAL RIVER 3 - APR 83 ---
2 - 3CT181 SURRf - JA1
& APR 83 --- FT. CAL 10JN - 13V 78 --- ARC 4NSAS ONE DAVIS-9 ESSE 1 - OCT 83 --- PWR'S IN GENERAL
.! P.16.C.3769 : ADDITION --- T3 XWI.C.639 RE: PEMETRATIO1 A004 FLOOR DRAIN REMOVAL I DAVIS-SESSE 1 - JUL 75 - STARTUP (0% POWER) #
j ............. ALL DONE, PRESS < RETURN > (Er TO G3 T3 MENJ .................T y~
THERE ARE 183 ARTICLE (S) TO BE DISPLAYED: 19-4U3-1985 14:30:54 PWR CHECK WALWES - 1979 - 1982 P.36.8.3389 : CRD ---
3UIDE TU3E SUPPORT PIM F4ILED - SGS D4MAGED d!THI1 43JR$
dESTINGH3USE PWR'S IN GENERAL - JUL 82 P.05.8.0318 : RC MAKEUP LI1E CRACKS, THERM 4L SLEEVE DAM 4GE - FATIGUE FAILURE SUSPECTED
--- CRYSTAL RIVER 3 - JA1 82 --- 3C01EE 1, 2 & 3 --- PWR'S 11 GENERAL
! I#
- P.05.8.0020
- RCS CHECK VALVES LEAKED
--- 1CGUIRE 1 - FEB 82 - HOT SHUTD3W1 I L P.05.C.0067 : PORV STUCK 07EN - SCLENOID OPERATE 3 CONTR3L VALVE LOCKED IN
ENER3!!ED STATE St VENT RESTRICTI31 g IINN4 - JAN S2 - RAPID C30LD0d1 P.05.C.0050 : PORWS INOPER4BLE - LEAKIN3 DIAPHRAGM, AIR SUPPLY CHECK WALVE SURRY 1 - OCT 82 - SNJTD3dM P.06.8.0033 : GENERATOR EXCITER REQ'J IRED REPAIRS - REACTOR TRIPPED, FJ 1AMMER '
DAMASE PT. 8EACH 1 - MAR 81 - P3WER 3PERATIO1 P.05.0.0222 : MSCV DISC STJD LOCKING DEWICE FAILED, NUT FOUND IN TURBINE C3NTROL VALVE TURKET P3 INT 4 - MAY SO - SHUTD0dN '
P.05.D.3223 : AUX FW PU9P STEA9 $UPPLY CHECK VALVE LEAKED
--- NORTH AN14 2
- JUN 83 (PRIOR T3 INITI4L CRITICALITY) - H3T ' #
STAND 8Y P.06.D.3224 : MAIN STEAM NRW GUIDE TO DISC WELD FAILED b
--- 9AINE TA1CEE - JUN 83 - STARTUP (ISE POWER)
P.36.0.3236 : ADDITION TO WI.D.207 & 222 RE: MSCW DISC STUD NOT LOCKING DEWICE II TURKEY PT. 3 & 4 - NOW 83 - SHUTDOWN #
P.05.0.0266 : MS C1ECK VALWE ROCKSH4FT DISPLACED ACIALLY - SET SCREd5 L30SENED
e PAGE 26 o
4081N10N 2 - JUN 82 - REFUELIN3 P.05.0.0271 : ADDITION To d!.D.264 RE: 15 CHECK VALVE ROCK $ HAFT DISPL4CED '
ARIALLY - SET SCREW L305E1ED 4081N531 2 - JUN 82 - REFUELIN3 el P.06.D.0274 : CHECC ---
VALVE 313C/$fE1 SEPARATION - CfCLIC FATIGUE FAILURE SEQuGYAH 1 - NOV 82 - REFUELIN3 0
P.0$.D.3294 : ADDITION ---
T3 WI.D.274 RE: CHECC VALWE STUD BROKEN - F4TIGUE.
SEQUOY4H 1 - NOV 82 - REFUELIN3 O
P.0$.E.0199 : LOOSE ---
VALVE 1UT & WASHER - STEAM SUPPLY V4LVf IN3PER43LE SEAVER VALLEY 1 - MAR 79 - STR47UP (DE POWER)
)
P.0$.E.0244 : VALVE STEAM LEAKS 91548 LED PRVS
--- DCONEE 1 - SEP 79 - SFX POWER 2
P.0$.E.3257 : ADDITION TO WI.E.227 RE: FW PIPIN3 CRACKS
--- ZION 1 --- PT. BEACH 2 --- KEdAU1EE - JUN TO SEP 19 2
P.06.E.3259 : ADDITION TO dI.E.250 RE: FW PIPE CRACKS
--- SALEM 1 - JUN 79 - REFUELING 8
P.0$.E.0272 : VALVE UN304LIFIED SEISMICALLY - PIPE BRACE ADDED
--- SALEM 1 - JAN 80 - 984 P3 DER 1
P.35.E.0278 : CHECK VALVE 4tNGE PIN SET SCREW N3T FULLY INSERTED
--- CRYSTAL 41VER 3 - APR 80 - COL 3 SHUTD3WM (A
) P.05.E.328o VALVE LEAKS RAISED PENETR4TI3N R031 HUMIDITY A80VE DESI51 LIMIT 3CONEE 2 - Jug 83 - 6St POWER O
P.OS.E.3301 : PENETPATION R33M HUMIDITT HIGH - FW VALVES LEAKED 3CONEE 2 - OCT 83 - 130% POWER /
s Q
8.06.E.3315 : SHUTDOWN ---
RECJIRED TO DEBURA CHECK WALVE PIV0T PIM BUSHI13 NORTH AN1A 2 - AUG 83 (L3W POWER PNVSICS TESTING) - DE POWE4 j P.06.E.3319 : LEAKING CHECC VALVE avERP4ESSURIZED 4UE FW PUMP SU'CTION STRAINER - CASE CaACKED #
--- C00K 2 - JAN 81 - 723 PodER P.06.E.3325 : DIRTY CHECK VALVE STUCK DPEN COOK 1 - FEB 81 - 1031 P3WER P.0$.E.0331 : FW --- CNECK VALdE SEAT / DISC REQUIRED LAPPING #
SEQUOYAH 1 - MAR 81 - COLD SHUTDOWN l
P.0$.E.0340 : .tAIN FW CHECC VALVE DISC STUD NUT MISSING ' '#
--- TURKEY P31NT 3 - APR 51 - REFUELING P.0$.E.3362 : AUX FW PU1P CHECK VALVES LEAKED
) ---
COOK 2 - JUL B1 - 90% PoJER j
~
P.0$.E.3368 : TDAF4P ---
CUCTI3N PIPING DVERPRES$URIIE3 - RELIEF V4LVES 11$fALLED II MCGUIRE 1 - AJG 81 (PRIOR TO INITIAL CRITICALITY) - STARTUP P.36.E.3370 : Fw PJMP DISCgnaGE CHECK VALVE FAILED - UNIT TRIPPE3 b
l e
PAGE 2T PT. SEAC4 1 - NOV 80 - GREATER THAN FPI POWER P.06.E.0373 : EFW PUMP TJR3!NE STEAM SUPPLT CHECC VALVES IMPROPERLF ASSEM8 LED
--- ARKA15AS ONE 2 - OCT 51
- SHUTD0dN P.0$.E.3378 : FW---CMECK VALVE DISC RETAI11N3 NUT COTTER PINS FAILED "'
SURRY 1 - JAN & APR 83 - SHUTDOW1 P.06.E.3385 : STRIPPER FEED HEATER TUBES LEAKED - RELEASE SURRY 2 - NOV 81 - REFUELING P.06.E.3386 : STEA1 SUPPLY CHECK VALVE PARTS LODGED IN AUX FW PUMP TRIP VALVE
- DISC NUT RETAINING PIN INSTALLAT101 SUSPECT 10RTH AN1A 1 & 2 - DEC 81 - 99 81003 POWER P.05.E.0388 : ADDITION TO VI.E.373 RE: FW PUMP STEAM CHECK VALVE ARKANSAS 3NE 2 - DCT S1'- COLD SHUTDodN P.06.E.0398 : AUX FW PUMP LEAKOFF LINE CHECK VALVE FAIL!D Coot 1 - APR S2 - 103E P3WER 9
P.36.E.3400 : ABFP EQUALIIING LINE CHECC VALVE STUCK DA1 AGING PUMP THRUST BEARING - VALVE INTERNALS RE10VED INDIAN PT. 2 - Mar 82 - 40T SHJTDOWN P.05.E.3402 : MAIN FW CHECC VALVE STUD Muf FASTE1ER FAILED, MODIFIED TURKEY PT. 4 - JUN 82
- SHUTD0dN
] P.0$.E.3412 : EFW PUMP TURRINE STEAM SUPPLT CHECK VALVE FAILED - DESI3N ERROR ARKANSAS ONE 2 - OCT 82 - REFUELING g, .
P.05.E.3413 : STEA1 SUPPLY VALVE INOPERABLE - NUT 5 WASHER LODGED I4 V46VE 3EAVER VALLEY 1 - SEP 82 - SHUTD3WN () .
P.05.E.0415 : SG Fd ISCLATI3N VALVE LOJ NITROGEN ACTUAT3R PRESSURE - LEAKY '
PLUG ON NITR3 GEN ACCU 1ULATOR E.
--- SAN ONOFRE 2 - SEP 82 (PRE 0PERATIONAL)
P.05.E.0416 : AUX STEAM TU43INE PRES $URE C3NT40L VALVE FAILED TO CL35E - >
) FOREIGN MATERIAL UNDER SEAT - UPSTREAM CHECK VALVES SJSPECTED SOURCE
--- TURKEY P31NT 3 - SEP S2 - 1005 P3WER P.0$.E.3417 : AUX FW PUMP STEA1 SUPPLY CHECK VALVE DISC STUD 8ROCEN, MUT A1D C3TTER PIN MISSI13 SALEM 2 - OCT 82 - 82t POWER P.36.E.3463 : STE41 BINDIN3 IN AFW PUMPS - LEAKA3E FROM MAIN FJ SYSTE1 -
PWR$ I4 SENERAL - JAN 84 --- C00( 2 - JUL & QCT S1 - 70I POWER R H07 STANDSY --- CRYSTAL RIVER 3 - OCT 8 DEC $2, N3V S3 -
) NOT STAND 87 ' */
P.0$.E.3542 : AUX FW PUMP INOPERABLE - STEAM LINE CHECK VALVE DISC NOT IN PLACE - DISC-TO-MANGER STJD SHEARED OFF b f' *--
SALEM 2 - AUG 82 - 82% P3WER P.0$.E.0623 : ADDITION TO VI.E.417 5 542 RE: AUX FJ PUMP STEAM SUPPLY C1ECC II VALVES DAMAGED - PRESSURE OSCILLATIOMS - VALVES REPLACED I --- SALEM 2 - AUG, OCT 82 - $25 P3WER gg
e 8 AGE 28 P.07.A.3200 : CHECK ---
VALVE LEAK!D - VALVE 83DY OWER$1IED 4RKAN545 ONE 2 - JAN 79 - 193 P0 DER (P0 DER ESC 4LAT!3N TESTING)
- P.07.A.0204 : INCORRECT VALVE WEIGHTS USED IN PIPE STRESS CALC 3LATI3NS -
HANGERS M3DIFIED gg
--- NORTH AN4A 1 - MAR 79 - 923 P3 DER P.07.4.3229 : 847 30RON C04 CENTRATION L3W - VALVE LE4KED - LIMIT SdITCHES READJUSTED I!ON 2 - OCT 79 - 94X POWER P.07.A.0230 O
$ WINS CHECK dALVE SEARING CAP 0-RING FAILED
--- 1AINE YA1(EE - OCT 79 - 97I P3 DER P.07.A.3240 : HPI CHECK VALVE SEAT HOLD *D0dNS F3343 LOOSE 4
1
--- THREE MILE !$ 1 - FE8 80 - SHJTD3WN P.07.A.3256 : CONT 4INMENT PENETRATION CNECK VALVE SEATS LEAKED
--- CONN YANCEE CHADDAM NECK) - M4Y 53 - REFUELI1G s
1 P.07.4.3260 : HPI VALVES LEAKED - PIPINS HEATED EXCESSIWELY
--- 3CONEE 1 - JU1 83 - 73% POWER
- l P.07.A.3262 : FAILED CHECK VALVE, 3 PEN ISCLATIO4 VALVE - 20 GAL RELEASE CRYSTAL RIVER 3 - JUL SO
- SHJTD3WN r
P.37.A.3266 : SIT---RECIRC LINE LEAKED - dELD CRACCED - VALVE CONTROLLER FAILED MILLSTONE 2 - JUL 80 - 1303 P3JER
! It P.07.A.3270 : CHECC VALVE STUCK CPEN - DISC WELD /V4LVE 30DY INTERFERE 1CI SEQUOYAH 1 - SEP 50 - HOT STA108Y
- O P.37.A.3271 : $1T---PRESSURE LOW - NITROGEN CONTROL VALVE FAILED l 4RKANS45 3NE 2 - OCT SD - 103% P3WER
gg P.07.A.3273 : CFT 3VERPRES$URIIED - CHECK VALVE IMPROPERLY ASSEM3 LED
--- DAVIS- BESSE 1 - OCT $0 - HOT ST4ND8Y j
P.07.A.0275 : HIGH HEAD SI PUMP CHECK V4LVE ANTI-R3TATI3N DEVICE BINDINS DUE TO FLOd OR PRESSURE SURGES 1
3EAVER V4LLEY 1 - OCT 80 - SHJTD3WN t
i P.07.4.0278 : ADDITION TO WII.A.273 RE: CHECK VALVE j
--- DAVIS-SESSE 1 - DCT 83 - HOT STA408Y P.07.A.0280 : SI TANK LEVEL LOW - 50LEN3ID VALVE LE4KED
--- CALVERT CLIFFS 2 - N3d 83 - 1001 POWER P.07.4.3285 : SI SYSTEM CHECK VALVE STUCK 3 PEN
--- SALEM 1 - DEC 83 - SHJTD3WN !
- o P.07.A.3291 : SI ACCUMULAT3R 83RON DILUTED - CHECK VALVE LEAKED
--- SURRY 2 - JAN 31 - 103% POWER .
gp P.07.4.3294 : LPI CHECK VA*VE . DISC COCKED BY DEP35tT BUILDUP
--- 3CONEE 1 - FE3 81 - 31 PJWER (NE4 TUP)
P.07.A.3302 : ADDITION T3 d!!.A.294 RE: CHECK VALVE 3C0%EE 3 - MAR 81 - 54uT33dM gg I
~
I DAGE 29 P.07.A.3306 : ACCU 1ULATOR DISCHARGE CHECK WALVES LEAKED STA40ST1CGUIRE 1 = APR 81 (PRIOR TO IMITIAL CRITICALITV) - H3T D.07.4.3307 :
COLD LEG INJECTION ACCUMJLAT3R DISCHARGE CHECK VALVES LEACED STAN3av 1CGUIRE 1 - APR $1 (PRIOR TO INITIAL CRITICALITY) - H3T P.0F.A.3310 : LEAK h
--- TEST DAMAGED UN! ACCUMULATOR WALVES - SEAT TYPE CHAN3ED SnuTD0=M 1C3UIRE 1 - MAY $1 (PRIOR TO IMITIAL CRITICALITY) - C3LD g'
P.07.4.3311 : ACCU C3LD1JLATOR
$HuTD0d1CHECK VALVE FAILED - 83T4 RHR TRAINS IN0PERAB*.E IN SHUTDOWN MCGUIRE 1 - 94V 81 (PRIOR TO I1!TIAL CRITICALITY) - C3LD '
D.3T.A.0315 : SI ---
CHECK VALdES STUCK QPE4 PT SEACH 1 - JUL 81 - SHJTD0dN D.07.4.3316 : ADDITION TO WII.A.243 RE: TILTING DISC CHECK VALWES THREE MILE IS 1 - JUN 81 - COLD SHUTD3WN .
P.07.4.3339 : CHECC VALVE 31$C MIS $1NG = C345TRUCTION ERR 3R
--- COOK 2 - 3CT 81 * $NJTD0dN D.07.A.3341 : SI ACCU 9ULAT3R BORON CONCENTRATION HIGH - VALVES LEACED -
PROCEDURES REVISED
--- NORTH ANIA 1
- N3W 81 - 995 P0 DER #
D.07.A.0363 : LEAKING $1 Sf5 TEM CHECK VALVES SEATED FOLLOWING FLUSHING
--- *T 8EACH 1 - OCT 81 - REFUELIN3 g)
P.37.A.3345 : SI ACCUMULAT3R BORON CONCENTRATI04 HIGH - VALVES LEAKED, FAULTT '
PROCEDURE
--- 40RTH AN1A 2 - OCT 81 - 97X PodER PoDT.A.3346 : ADDITION To dII.A.331,335 8 337 RE: ACCUMULATOR BORON J CONCENTRAT!315 40RTH AN14 1 & 2 - SE* 81 - 133X POWER J
Do07.A.3352 : 817---
INLET VALVE FAILED - IHAFT PACKIMG T03 TIGHT suRRY 1 - DEC 31 - C3.0 5HUTD3dN D.07.A.3363 : LPSI $ WING C4ECK WALVE WEAR, DAMAGE - DESIGN PALISADES - SEP 81 - REFJELING P.07.A.3370 $ DING CHECK WALVE PR08LEMS
--- PWR5 I4 3ENERAL - JUN 82 ((8.7.A.152))
t y P.07.A.0372 : HP I1JECTION STOP CHECK VALVES STUCK - UP TO 580 PsID REGJIRED T3 ESTABLISH FLDJ
--- DAVIS-8 ESSE - JUN 82 - REFUELING U D.07.A.3375 : ADDITION TO WII.A.372 RE: HP INJECTI3N ST3P CHECK VALWE STUCK -
INCORRECT SEAT ANSLE II
--- DAVIS-SESSE - JUN 82 - REFUELING --- PWRS IN GENERAL = AUG
, 82 9
t 040E 30 o
P.07.4.3384 : $1T 3UTLET C1ECE VALVE LE4EED 4tNGS DETERIORATED
--- CALVERT CLIFFS 1 - JUL 82 - ST4RTuP P.07.4.33S6 : COLD LEG INJECTION CHECK v4LWE LEACE3 - C4RBON STEEL CL3SJRE
$TuD$ CORRos!D NORTH 4N14 1 - SEP 82 - REFuELIN3 P.0F.4.0390 : 4DDITION 70 dII.A.275 RE: VELAN CHECC VALWE SINDING PR3BLEMS -
DI$Cs REPLACED g) 3E4WER v4LLEY 1 - OCT 82 - $PECILL TESTING P.37.4.3392 : $1 CgECK v4LdES STUCC - DISC STUD PR3TRUDED A80VE nut, DISC MIS 4LIGNED
--- 4RK4N545 DNE 2 - OCT S2 - REFUELING P.07.4.3396 : MINot LEAK 4GE INTO SIT C01POJNDED BY $1T LEVEL' INDICATING '
J
$fSTEM FAILURE C4JSED TEC4 $PEC LI4ITS 70 8E EXCEEDED - W4LVE LEAC4GE, TEMS EFFECTS ON REFERENCE LEG
--- PALIS40ES
- SEP-DEC 82 - 1003 P0 DER P.07.4.3397 : $1 T4NC LEWEL Lod, 83RON CONCENTR4 TION L0d - VALVE LE4K4GE '
P4LISADES - N3V S2 - 1001 POWER P.37.4.3400 : ADDITION T3 dII.A.397 RE: CHECK V4LVE DISC MISSING C00C 2 - DEC S2 - REFJELING I
P.07.4.3403 : ACCUMULAT04
--- 30R04 CONCENTRATION L34 - CHECK VALVE LE4(ED
$URRY 2 - $EP 82 - 133% POWER P.07.4.3422 3 4DDITION 70 d!!.4.354 RE: SIT OUTLET CHECC VALVE LEACED #
RING $ DETERI3 RATED - MATERIAL CHANSED
--- CALVERT CLIFFS 1 & 2 - JUL 82 - STARTuP '
P.37.8.3063 : $PR4f PUMP M370R $TATOR C31LS/PM4$E LEAD CH4 FED, $NORTED O
--- CALVERT CLIFFS 1 - SEP 77 - 133% PodER '
I P.07.8.3080 : CMECC VALVE DI$C AND RELATED PARTS N3T IN$TALLED O
--- SE0uGYAH 1 - APR 80 - PREOPERAT!3NAL 2
P.07.8.0094 : SdST LEVEL L3d -
48$ $YSTEM WALVE FAILED TO RESEAT, 4LARM FAILED
--- 3CONEE 3 - APR 83 - ME4TJP P.37.8.3123 : CONT 4!NMENT SPRAY ADDITIVE T4NK CHECC VALWES LEA (EB RANCMD SECO - APR 82 - SMUTD0d4 P.0F.8.0131 : CONT 4IN9ENT $ PRAY & 440H 4DDITIVE SYSTEM CHECK V4LVE FAILED -
SEAT & DISC 14CHINED
--- GINNA 1 - SEP 82 - 133% PodER >
P.07.8.0134 : $PRaf PUMP CdECK v4LVE PIN /$ DING 4RM WORN
--- GINN4 - JuN 82 - 100% PodER ' #
P.37.8.3142 : C3%T4INMENT $ PRAY Pu1P DISCH4RGE CHECK VALVE FAULTT
- PIN #$ WING 4EM JORN GINN4 - Jul 82 - 100% POWER
- b P.07.D.3034 : Stutt VALVE $ FREED BY CYCLIN 3 - CHECC VALVE INSTALLED 84C(WARDS II
- PR3CEDURE ERROR
--- COOr 1 - $EP 751 PL te 79 TO N3d 83 - 133% POWER --- C00C 2 - JJN 50 -
gg
- ~ .. ,e=- .
-m-.-~ - -..- . - -
e CAGE 31 i
i P.0T.D.0047 : C3NTAIN9ENT STEA9 HEAT SUPPLT CI CHECK VALVE FLANGES LE4 KID -
SOLTING RELAtED YANKEE R3dE - APR 82 - 1303 P3 DER P.07.0.3053 : CONTAIN9ENT ISOLATIOM VALVES F4ILED LLRT - COMPONENT FAILJRE, *#
DES!3N/ FABRICATION ERRORS 34VIS-SESSE - Jus 82 - $4UTDod1 O
P.07.0.3059 : C3NT41N9ENT ISOLATIO4 VALVE LEAK R4TE GRE4TER THAN R41GE 3F TEST EQUIP 9E1T - NEd TESTER PURCHASED SALEM 1 - APR 82 - REFUELING (CO.O SHJTD0dN)
P.07.E.007s : HYDR 3 GEN PUR3E FAN TRIPPE3 - CORR 33ES CHECK VALVE
--- 4REANSAS ONE 1 - MAR SD - 100E P3WER P.0T.E.0084 : Sd LEAKED 3Y IDLE VENT FA4 SEALS, *IPE F0JLING - HTDR3GE1 PURGE FLOW Low #
--- 4RKANSAS ONE 1 - JUN $ J JL 80 - 100% PodER P.07.E.0090 : HYDR 3 GEN PUR3E FAN TRIPPED - SE4L WATER LEAK CAUSED CHECK VALVE #
i STICKING AND FILTER RESTRICT!31
--- AaKANS43 04E 1 - SEP $0 - 1005 P3WER P.07.E.0129 : FIRE PROTECTION SPRINKLER SYSTEM ALARM CHECK VALVE LEAKED -
DISC SEAL INJERTED CONN YAN(EE (HADDAM NECE) - M47 82 - 1001 POWER P.07.E.0151 : HYDR 3 GEN DILJTIO4 SL3 DER CHECK VALVE STUCC - MISAPPLICATI3N, LIFT CHECK REPLACED WITH SWIVS CHECK g,
--- DAVIS-SESSE - N0d 82 - 973 POWER P.07.E.3174 : FIRE Pa0TECTION SPRINKLER SYSTEM CHECK VALVE GASCET LE4CE) )
- --- YANKEE'R3dE - SEP 82 - 82% P3 DER i
P.07.E.3175 : SPECIAL HAIARDS HEADER AL4RM CHECK V4LVES FAILED TO 314T s NI
--- YANKEE R3dE - DEC 82 $ FEB 83 - REFUELING & 103% PCdER P.0$.A.3341 : LOOP FILL CHECK VALVE LEACED - SEAT!1G SURFACES LAPPED J CONN. TAMEEE (HADDAM NECE) - FEB 79 - REFUELING P.05.A.0352 : STORAGE TANK BORON lod - CHECK VALVE LEAK!D '#
GINN4 1 -
APR 79 923 POWER l P.03.A.0401 : CHAP 3tNG PUM* DISCMARGE C4ECC VALVE SEAT LEAKED
--- PT. SE4C4 2 - DEC 79 - 983 POWER P.05.A.3404 : LEAKING CHECK VALdE DILUTED 50RIC 4CID ST3 RAGE TANKS GINNA 1 - DEC 79 - 33 POJER P.0$.A.0414 : ADDITION TO WIII.A.434 RE: B4ST VALVE LEAK ' #
3 INN 4 1 - DEC 79 - 3E P0 DER l P.05.A.3435 : M4KEJP SYSTE1 CHECK VALVES LEAKED - SUILDUP ON SEATS 00 j --- 34WIS-8 ESSE 1 - gav 83 - REFUELING i
P.0$.A.04S9 : WRON3 INTERN 4LS USED IN VALVE II
--- 10RTH AN14 2 - SEPT 83 (PRIOR TO POWER OPERATION) - C3LD
, SHUTDOWN f
gg
0 PAGE 32 P.05.4.0417 : MPI sump BEARING DVERNEATED - C00LIN3 $YSTEN DEFICIENT, FILTER CLOG 3ED, CHECK VALVE STUC(, PER$341EL.ER434 3CONEE 2
- h3W 83 - 130% PCWER P.08.A.0508 31T, BAST 8043N DILUTED - CHECK VALVE LEAKED
--- 10RTH AN1A 2 - DEC 80 & FE8 81 - 1003 POWER g' P.05.A.352T : VALVE SONNET GASKET LEAKED - Post *1AINTENANCE TESTIN3 01ITTED
--- SEQUOYAH 1 - DEC 80 - 103E P0 DER G P.08.A.3553 : CHAR 3ING PuM* FAILED TWICE ARK ANS AS ONE 2 - J UL, AUG 81 - 72, 40E POWER gg P.08.4.0554 : CHAR 3!NG LINE CHECK VALVE STUDS CORR 3DED
--- C00K 2 - OCT 81 - SHJTD0dN ;
P.QS.A.0566 : VCT 345 $UPPLY VALVE LEAKED - RELIEF VALVE SETPOINT L3W -
RELEASE CALVERT CLIFFS 2 - N3W 81 - NOT STAND 37 P.08.A.3575 : CONTAINMENT ISQLATION VALWES REQUIRED MAINTENANCE '
SEAVER VALLET 1 - APR 80 - SMJTD3WN P.33.A.3580 : DI HATER HEA)ER C3NTAMINATED - CHECK VALVE LEAKED AREAN$AS CNE 1 - JAN 52 - 85% P0 DER P.35.A.3618 : CHAR 31NG PUM* OIL PUMP CHECK VALVE STUCK - GUIDE FINGER SENT
--- COOK 2 - APR 52 - 103% P3 DER P.05.8.3228 : SERVICE WATER VALVE DISC dQR1 AT NINGE COOK 2 - JAN 79
- 1005 P3WER P.08.8.3233 : $4LTdATER PU1P DISCHARGE CHECK VALWE FAILED TO OPEN - F3REIG4 DEBRIS SUSPECTED CALVERT CLIFFS 1 - FE3 79 - 945 POWER #
G P.05.9.0234 : $ERVICE WATER CHECK VALVE DISC HINGE POINTS WORN - REPL ACED WITN $$ DISCS COOK 1 - FES 79 - 1005 PDWER J P.03.B.3249 : ADDITION TO WIII.3.234 RE: CHECK WALWE HINGE WEAR C00K 2 - APR 79 - 103% P3WER #
P.03.8.0255 : C3NTAINMENT ISOLATION VALWES LEAKED - NEW TYPE CHECK WALVES INSTALLED
--- COOK 2 - JUN 79 - REFUELING P.05.8.0265 : WELD---
CRACKS FOUND FOUND 14 B3 RATED WATER PIPING ~"
SAN ON3FRE 1 - SEPT 79 - 103% P0JER P.05.8.0290 : NESW VALVES LEAKED - SAND DEPOSITS, PITTED SEATS ' #
--- COOK 2 - DEC T9 - REF3ELING P.05.8.0310 : RIVER WATER CHECK VALVES DETERIORATED
--- SEAVER VALLEY 1 - APR 80 - SMJTD3dN
- P.09.8.0319 : SERVICE WATER PUMP DISCHARGE CHECC VALVE DETERIORATED
--- PALISADES - AUG 79 - 501 POWER P.0S.S.3364 : ADDITION TO WIII.3.319 RE: C4ECK WALVES - SEAT RING C3RR25!04 $d -
h 8
PAGE 33 PALISADES - AJG 79 - BOE POWER P. 1 3.3380 : CHAR 31NG PUMP $W PUMP $ PLASHED WITH DATER - MOTOR WI4DI1GS SMORTED SURRY 2 - APR 81 - 1333 POWER t)
P.05.8.3382 : INADVERTENT C3NT41NMENT SPRAf - DES!3N/ VALVE ERR 3R ~~
--- CRYSTAL RIVER 3 - Jul 80 --- SEQUDYAH 1 - FES 81 - C3*D .
SMUTDOWN --- PWRS IN GENERAL - MAR 81 gy P.05.8.3394 : $d PJMP DISC 1ARGE PIPING, VALVE SC47 RING THREADS CORR 03E3 INDIAN PT. 2 - M47 81 - COLD $1UTD0d1 ()
P.03.B.3411 : LEAKING NESW CHECK VALVE 5 TO SE REPL4CED d!TH DI4PHR4GM TfPE VALVE COOK 1- -
JUL 81 - REF'JELING P.03.8.0423 : CHAR ---
31NG PUM8 SW CHECK VALVE FAILED TO SE4T - DISC WORN /8INDING #
SURRY 2 - AUG 81 - 133% POWER P.05.B.3461 : CHARGING PUM8 SW PUMP CHECK VALVE dORN, STUCK #
$URRY 2 - JAN 82 - 8$t P3WEa P.0S.'8.3493 : SW $YSTEM $ WING CHECC VALVE DISC SE!!ED FULLY CPEN - CORR 3SI3M DEPOSITS, DISC STJD DETERIORATEb
--- DAVIS-SESSE - JUN S2 - C3LD SHUTDOWN P.05.8.3521 : CCW---
SUPPLY IS3LATION VALVE LEAKED - CLAPPER AND SEAT LAPPED PT. BEAC4 1 - NOV 82 - REFUELI4G D
P.33.8.3566 : RBCU SW CHECC VALWE STUCC CL3 SED S U M.1E R - OCT 82 - OX POWER (ST4RTUP)
O P.35.8.3571 : $d PJMP STRAINER CLOGGED - SJ 8AY FL30DED, VALVE G45CET FLILED SALEM 2 - DEC $2 $ JU1 81 - $$t POWER & SHUTD0dN #
s O P.05.8.3610 : ADDITION TO VIII.B.378 RE: MISSING FLANGED 8EARI1G ON VALdE ACTUATOR F0013 TO BE CAUSE OF FAILURE
--- DAVIS-5 ESSE - APR 81, NOV 82 - 93X P3 DER ~
8 P.09.D.0226 : AUX FW PUMP $ PEED L0d - OWER$ PEED CHECK VALVE LE4KED, GDVERN3R MI5 ADJUSTED SALEM 2 - MAR 82 - 103E POWER P.09.F.3120 : VUCDT FLOW T3TALIIER ERRATIC - CHECK VALVE LEAKED '
1CGUIRE 1 - AUG S SEPT 82 - 504 POWER P.09.H.0581 : R/A GAS MINGT3R CHECC VALWE LEAKED - POSSIBLY DIRT FR3M PJMP J VANE WEAP SINNA - 3EC 81 - 100I P0 DER f J P.09.H.0637 : CONT 4!NMENT ISOLATION CHECK VALVE LE4KED, VALVE INTER 1ALS DIRTY
- REPORVING PROCEDURES REVISED PT. BEAC4 2 - APR 80 & JUN 82 - REFUELING & 103% P0 jet ~
O(i P.07.H.0640 : RADIATION MOVITOR SA9PLE RETJRN LINE CHECC VALVE LEAC - F3 REIGN MATERIAL, SQLYED 30N1ET TfPE VALVE INSTALLED I)
--- GINN4 - APR 82 - REFUELI1G P.09.H.3689 : R/A MONITOR CHECK VALVE LEAKED = F3 REIGN MATERIAL II
etGE 34 s
5thMA -
JUN & SEPT 82 - 1005 P3WER
?.13.4.0060 : CONTAMINATIO1 0F PRI94RY 1AKEUP T41E RESULTED IN ERR 31E3US D 8340N ANALY$!$
SAN ONOF4E 1 - MAY 83 - REFUELINS 99 P.11.A.3408 : CHECC VALVE LEAK - DG $L3d T3 REACH RATED SPEED
--- FAALEY 1 - MAY 83 - 1001 POWER C)'
P.11.A.3450 : DG ---
$ TART TIMES EXCESSIVE - FUTL SUPPLY CHECC VALVE LE4KED
' FARLEY 1 - M4R TO SEPT 83 - 133% POWER
()
P.11.A.0682 : ADDITION T3 XI.A.(66 RE: DG AIR START SYSTEM
--- (EW4UNEE - JAN 81 - 1303 POWER J
P.11.A.3525 : bG PROBLEMS FARLEY 1 & 2 - A84 8 1Af 81 - 103E P0 DER E SHUTD3dM s
P.11.A.3541 : DG C30 LING W4TER VALVE SH4FT/ DISC MI5ALIG1ED - V4LVE 3PER4TI3N REVERSED
--- PALISADES - JUN 81 - $6% POWER 4
P.11.A.3570 : DG FJEL DIL CONNECTION LO3SE, CHECC VALVE SEATED IMPR3PER.Y FARLEY 2 - SEPT 81 - 1005 POWER #
P.11.A.3589 : DG FJEL DIL LINES REQUIRED MANUAL PRIMING
--- 408IN$3N 2 - NOV S1 - 5OE POWER P.11.4.0614 : DG C30 LING W4TER PUMP DISCHARGE CHECC VALVE DISC SEP4RATI24 -
ENGINE TRIPPED 01 HIGH TE4P gj
--- *WR$ IN 3ENER4L - DEC 81 ((8 11.4.430))
P.11.A.3656 : EMER3ENCY DG PROSLEMS
--- COOK 1 8 2 - JUL 82 - REFUELIN3 8 100E POWER
/
P.11.A.0717 : DG MECHANICAL PR38LEMS REWIEJED -
--- PWR'S IN GENERAL - AU3 83 --- CALVERT CLIFFS 1 8 2 - LPR 33 -
RANCHO SECO - 14Y 83 --- SAN ON3FRE 1, J
P.11.A.3721 : DG STARTING FAILURES - 843EEN INLET 4IR CHECK VALVE DISC, MISSING COUNTERWEIGHTING - FUEL SUPPLY LINE REQUIRED PRI11NG
--- CALVERT CLIFFS 1 & 2 - JUL 82 1 4PR 83 - 100t PodER t P.11.8.0519 : CIRCUIT BREACER CONTACTS 1I34LIGNED (EW4UMEE - APR 81 - 1002 POWER
- P.13.4.0166 : W45TE GAS RELEASES - RUPTURED COMPRESSOR RUPTURE DISC $, F4ILED PRES $URE CONTROL WALVE - 4LARM RESP 015E DELAYED #
SAN ONOFRE 2 & 3 - SEPT $2, M4F 56 - 1001 POWER, HDT STA108Y t .)
P.16.A.0323 : CONT 4INMENT PENETRATION LEAK 4GE - VARIOUS CAUSES
--- PALISADES - N3V 80, SEPT 81, AUG & OCT 83 - DI P3 DER (COLD SHUfDOWN 8 REFUELING)
P.16.e.0172 : RUST SCALE L3DGED IN H2 PJRGE SYSTEM CHECC VALVES
--- ARKANSAS ONE 1 - JUN SO - 1005 P3WER II P.14.8.3228 : CONTAINMENT PU4GE VALVE AIR SUPPLIES LEAKED
- ALISADES - SEPT 81 - REFUELIN3 hE
_- =
o PAGE 35 P.16.8.0237 : $W PJMP ROOM DRAI15 UNISOLA8LE - DESIGN ERROR
--- CALVERT CLIFFS 1 S 2 - N3V 81
- 1003 POWER #'
- .15.A.3376 : C3NTAINPENT ISOLATION SERWICE AIR CHECK V4LVE LEAKED - DISC DIRTY PT. 8E4C4 1 - OCT 81 - REFUELING P.15.A.3379 : SERVICE AIR CHECK VALVE LEAKED - 11TERNALS DIRTT '
mi. BEAC4 1 - OCT 81 - REFUELI1G
' P.15.A.3334 : AUX---FW SYSTE1 CHECK VALVES L3CKED SEISMIC PROTECTION SALEM 1 5 2 - FEB 82 - REFUELING & 103I POWER P.15.A.0391 : CHECK VALVES LEAKED - SEAT C3RRJS!3N '
PT. 8E4C4 1 - MAR 82 - 931 PodER P.16.C.0823 : SURVEILLANCE PROCEDURE DEFICIENT - PLANT 3RDERED $ NUT D3W1 #
4
--- 4RCANSAS ONE 1 - JUN 19 - STARTUP ((
~
' P.16.C.0854 : REQUIREMENTS FOR MODIFYINS TEST PR3CEDURES DEFICIENT #
DAVIS-3 ESSE 1 - JUN 77 - STARTUP P.16.C.0922 : ADDITION TO XVI.C.823 RE: PROCEDURE VIOLATION 4REANSAS ONE 1 - JUN 79 - STARTUP P.16.C.1135 : RIVER WATER STSTEM CHECK VALVES INST 4LLED SACKWARDS SEAVER V4LLEY 1 - SEPT 83 - REFUELING P.16.C.1139 : MAINTENANCE FAILED TO DOCJME1T REM 3 VEL OF CHECK VALVE 11TERNALS I#
SURRY 2 - SEPT 83 - 1303 POWER P.16.C.1297 : CHECK VALVE CLOSURE dEIGHT MISSING (
NORTH AN1A 1 - MAR 81 - SHUTDOJN P.1$.C.1561 : VALVE MOTOR FOUND DISCONNECTED - M3DIFICATION WIRING ERROR 43 SEQUOYAH 1 - NOV 81 - 103% POWER '
P.16.C.1776 : BAAT PUMP DISCHARGE VALVE F0JND CLOSED - PROCEDURE ERROR >
DAVIS-8 ESSE 1 - MAY 82 - COLD SHJTDOW1 P.1$.C.1848 : NON!3ENTICAL REPLACE 9ENT PARTS - RECJRRINS PROBLEM J SEAVER V4LLEY 1 - OCT 80 --- 8ELLEFONTE 1 S 2 - DEC B2 ---
P6R'S
)
P.1$.C.1909 : EDG STARTING AIR 3ANK PRESSURE LOJ - OPER4 TOR ERROR
, --- SURRY 1 - SEPT 82 - 130% POWER
............. ALL DONE, PRESS < RETURN) (EY TO G3 T3 MENU ................ 73 THERE ARE 95 ARTICLE ($3 TO BE DISPLAYED: 19-4UG-1986 14:37:31 PdR - CHECK VALVES - 1953 - 1986 . >
P.05.C.0059 : RCS LEAKAGE - PORV 8 LOCK VALVE, SI C1ECK VALVE, SG Tutt 1ILLSTONE 2 - MAR 83 - 1303 P3 DER b P.05.E.0063 : PORV LEAKAGE PROBLEMS, C01TR3L AIR !$0 LATED, PCS VENT PAT 4 L3ST RA100M MAI1TENANCE PR03LEM - PORV LEAKAGE PR08LEM U1RES3LVED ()
SALEM 2 - JAN 83 - C03LD3dN, S H L' T D O W N P.05.F.0043 : PRIM 4RY C00L4NT STSTEM LE4KASE - CHECK VALVES FLUSHED T3 NI
f PA E 36 IMPR3VE SEATING PALIS4 DEI - JUL $4 - HOT STANDSY o
P.35.F.3052 : PLANT SMUTD0d4 - EXCESSIVE RCS LEACA3E - WALVE PACKI43, RCP FLAN 3E LEAKED CATAWBA 1 - OCT B5 - 1035 PodER #
P.36.A.3096 : REACTOR TRIP DN LOW SG LEVEL, POWER TRANSFERRED FROM 1AI4 T0 RESERVE - TU451NE ST3P VALVE FAILED TO CLOSE - DISC SEP4R4TED, SEAT D4MAGED
()
1AINE YA1CEE
- N3W 84 - SOE P3 DER .
O P.06.A.3101 : TURBINE /RE4CTOR TRIP - HIGH TUR814E WIBRATION DURING C04T40L VALVE TESTINS - MAIN L AUE FEED PU4P PROBLEMS - SPEED SETPOINT HIGH, 0IL PU1P DISCHARGE CHECK VALVE STUCC, MOTOR F4ILED j
--- DAVIS-8 ESSE - JUN 85 - 85% POWER P.05.A.3104 : REACTOR, TUR31NE TRIP - F1 ACCUMUL4TED IN MAIN TURSINE HP '
CONTROL DIL SYSTEM - DIRT ENTERED STRAINER $ AND 3RIFICES SURING MAINTENANCE INDIAN PT. 3 - OCT 55 - 30I P3WER P.06.D.3290 94IN STEAM N34 RETURN CHECC V4LVES STUCE OPEN - PACKI45 INDUCED FRICTION - C3MM04 MODE FAILURE PWR$ IN 3ENERAL - AUG 83 --- TROJAN - MAR 83 P.05.3.3305 : ADDITION TO WI.D.290 RE: 94IN STE41 MONRETURN CHECC VALVES STUCC UPON - PACCING INDUCED FRICT!04
---~ TROJAN - MAY, SEP S 3CT - COLD SHUTDOWN, POWER OPERATION
(#
P.05.D.3319 : REACTOR / TURBINE TRIPS - G4001D IN FW PUMP TRIP CIRCUITRY - MSCV PROBLEMS, FR4YED INSULATI3N -
VIBR4 TION TROJAN - MAR, APR 84 - 130, 781 POWER ()
P.06.D.3328 : ADDITION To v1.D.319 RE: REACTOR /TUR3INE TRIPS - MSCW PR33LE15 - #
HIGH FRICTI3N F3RCES FOUND -
gg
--- TROJAN - MAR, APR 84 - 130, 785 POWER P.06.0.3333 : MS!v FAILURES - ACTU4 TOR ROD SEAL PISTON LEAKAGE, FAULTf 44ND #
SJITCH C01 TACTS, REDUCED ACTJATOR PRES $URE
--- CALVERT CLIFFS 1 - JUL-DEC 84 - WARIOUS POWER LEVELS P.06.D.0347 : MSIVS FAILED TO CLOSE - AIR SUPPLY CHECK #ALVES SE4TED IMPR3PERLY - INSUFFICIENT SOFT-SE4TED POPPET GUIDING, R3'J3H 5007 SEAT FINISH SYRON 1 - MAR 85 - HOT STAND 8Y P.05.D.3348 : ADD 1FION TO d!.D.347 RE: 151V CHECC WALVES SEATED IM*R3*ERLY ON GRADJALLY DECREASING AIR PRESSURE - CHECK VALVES REPL4CED PWRS I4 3ENERAL - APR 85 --- BfR3N 1 - MAR 85 - H3T STAND 8Y -
-- OTHER UNITS AS LISTED IN ARTICLE i P.05.D.3353 : REACTOR TRIP - MSIV CLOSED - PRESSURE TRA4SDUCER FAILED, CHECK VALVE LEAKED g
dATERFOR3 3 - MAY 85 - 17% P3 DER P.36.D.3367 : MSIV FAILED TO CLOSE FULLY - GAS SUS 3LE I4 HYDRAULIC FLJI), HP WVDR4ULIC SU8 PRES $3R 9 LADDER PR03LEM3, ACCU 1ULAT3R C4P EN) CHECK
', A L V E FAILURE
--- CALVERT CLIFFS 2 - JUL-AJG 85
- HOT STAND 8Y gg
-. .. . l
, PAGE 3T D.05.E.0444 : FJ AIR --- ACCUMJL470R CHECK WALWES LE4K!D - DATER IN INSTRUMENT AIR CRYSTAL RIVER 3 - MAR 83 - REFJELING P.06.E.0454 : SD4Fd PUMP 113PER48LE - P3TE17tAL STEAM BINDING, 84CCLE4K4GE THROJGH DISC 4ARGE VALVES - STSTEM REVIEW UNDERTACE1 ROSINSON 2
- JUL $3 - T9I POWER p.06.E.0462 : AUX FW PUMPS IN0PERA9LE, WAPOR BOU1D - CHECE VALWE, LUBE 31L COOLER LEAK 43E
--- SURRY 2 - NOV E DEC $3 - 100% PodER O
P.06.E.3468 : FAILURE OF A1CHOR DARLING SOFT SEAL WALVE SEALS - IMPROPER VULC4NI!!N3 PWR$ 11 3ENERAL - FEB 84 --- INDIVIDU4L PLANTS AS LISTED
((B.6.E.1723)
- P.05.E.3481 : MDAFJP DISCH4RGE CHECC VALVE LEAKING - HI4GE PIN BUSNIN35 MIS $1NG, WOA1 '
FARLEY 1 - DEC 83 - 100% POWER e
D.0$.E.3483 : REACTOR TRIP - RCP TRIP - WATER IN M3 TOR LE4D PENETR4TIONS -
MAIN FW RESULATING V4LVE LEA (AGE, FAJLTY R0D POSITION I4DICATOR SURRY 2 - MAR 84 - 133% POWER .
P.05.E.3489 : REACTOR TRIPS - SG Fd LEVEL CONTR3L SYSTEM PROBLEMS -
COMP 3NENTS D4 MAGE 3 - DATER H4MMER #
--- SALEM 2 - APR 84 - 22, 6, 51 P3WER D.36.E.3495 : REACTOR, TUR3INE, GENERAT3R TRIPS - 141N FW TRANSIENT - STUCC ' #
PILOT VALV!, CHECC V4LVE FAILURE 1 1CGUIRE 2 - M4T 54 - 100 8 23% P3 DER O
P.06.E.3505 : REACTOR TRIP 04 L3W SE LEVEL - CIV F4ILED CLOSED - 50LEN310 VALVE FAILURE 1CGUIRE 1 - JJL S4 - 1005 POWER 5)
P.06.E.3521 : REVERSED TDAFP ROTATION CLUSED SUCTI3N PIPING OVERPRESSURIIATION - SUCTION INSTRUMENTATION DAM 4GED - CR4CKED #
CHECC VALVE dELD MCGu!RE 2 - AUS 54 - 100% POWER 4
P.06.E.0524 : MANU4L REACTJR TRIP -
FW HEATER TU3E LEAK - SG LEVELS DECREASED
- BINDING CHECK VALVE STE1 J
MILLSTONE 2 - NOW 84 - 62% PodER '
1 D.05.E.0555 : LP Fd HEATER CASCADE DRAIN CLOSED, HEATER SHELL SIDE J4 DER VACUJM -
CHECK VALVE INSTELLED BACKW4RDS J SYRON 1 - MAR 85 - 23% P3WER P.05.E.0574 : ADDITION TO WI.E.447 & 4$4 RE: AUE Fd PUMPS TRIPPED, IN3PERASLE 9 #
DISCHARGE d4LVES LE4KED R081NS3N 2 - APR, JUL $3 - SHUTD3WN, 19E POWER P.06.E.3588 : REACTOR TRIP, ESF ACTUATI3NS - L0d SG LEVELS - SG FW CHECC L
VALVE LEAKAGE, PR3CEDURAL ERROR
--- JCLF CREEC - JUN 85 - HOT STANDBT (PRE 0PERATIONAL) f)
P.05.E.0594 : REACTOR TRIP DN MAIN FW PJMP TRIP - FW PU1P RECIRC VALVE FAILED 70 HANDLE FL3d TRANSIENT
- C3NTROLLER DESIGN PR05LEM - $3 PORVS O
0 P&GE 38 FAILED TO OPEN - OUT 3F CALI5 RAT!31 CATAWBA 1 -
JUN 85 - 64% P0 DER 9
P.06.E.0615 : REACTOR, TUR31NE TAIPS - FW PUMPS TRIPPED - FW ISOLATION #ALVE, IRM CHANNEL AND Fd FLOW TRANSMITTER FAILURES
$U1MER - SEP B5 - 93% P0 DER gj P.06.E.0624 : REACTOR TRIP DN L355 3F VITAL GUS - LOSS 3F FW STSTE1 I1TEGRITY
- WATER HAMMER - 1AIN FW CHECK VALVES FAILED
()
9 POWER PWR$ IN 3ENERAL - JAN 86 --- SAN ONOFRE 1 - NOV $5 - 601 O
P.06.E.3630 : STOP CHECK VALVES IN AUX FW PUMP STEA1 $UPPLY FAILED 34 L3W d
FLOW - DISC ASSE19LY DEGRADED - VI844TIO1
- WRS IN 3ENERAL - FE3 86 --- TJR(EY PT. 3 & 4 - 43V 85-JAN 86 - VARIOUS power LEVELS ((
P.06.E.3634 : 53 BLOJDOWN 13NITOR IN0PERA8LE - FW IN 8 LOWDOWN LINE - CHECK VALVE BACKLEA(AGE
--- YANKEE R3dE - DEC 85 - DE POWER (STARTUP)
J P.06.E.3644 : AUX FW STEA9 $UPPLY !$0LATION VALVES INOPERABLE - GUIDE ETUDS SRCKEN - HIG4 CYCLE FATIGJE
- TURKEY PT. 3&4 - JA1 86 - 103, 1004 POWER ~,
P.06.E.3646 : CHECC VALVES OMITTED FROM EFJ PUMP STEAM SUPPLY - INA3EQULTE DESIGN REVIEJ j
--- ARKANSAS ONE 1 - JAN 56 - 90E P0 DER P.05.E.0648 : P3fE1TIAL FAILURE OF ANCH3R DARLINE CHECK VALVES - 14TER4AL TACK WELDS MISSIN3 - MFG ERR 3R SUMMER - FEB 86 - 103% P3WER O
P.06.E.0673t: REACTOR /TUR9INE TRIPS, FW IS3LATI315 - SG FW FLOJ C01TR3L PROBLEMS, C01 TROL CARD N3T ADJUSTED, PCT, VALVE FAILURES #
CATAWBA 2 - 14Y 86 - #ARIOUS P3 DER LEVELS (PRE 3P! RAT!3NAL) - bI P.06.E.36745: REACTOR TRIP DN MAIN FW PUMP TRIP - HEATER DRAIN PUMP TRIPPED -
CHECC VALVE STUCK OPEN - GALLING 34 NINGE PIN /
3CONEE 1 - MAY 86 - 45% POWER P.06.E.0375t: TURBINE, REACTOR TRIPS ON HIGH SG LEVEL - FEED PUMP DISCHARGE LINE CHECK VALVE ALL3dED REVER$E FLOd DN STARTUP - DISC P!VOT PINS MISSING - SETSCREWS NOT ANCH3RE)
INDIAN PT. 3 - MAY 86 - 25% P3 DER P.07.A.3401 : SI CHECK VALVE SEAL WELD LEAK - INADEQUATE APPLICATION 3F WELDING AND 3RINDING T E C H 1I Ql'E S 1AINE YA1KEE - FES 83 - SHUTDOJN l
P.07.A.0407 : COLD LEG INJECTION ACCUMULAT3R CHECK VALVES LEAKED '
MCGUIRE 1 - MAY 83 - STARTUP, 35 E 53% POWER P.07.A.3420 : HPI STOP CHETK VALVE STUCC - DISC TO SEAL CONTACT ARE4 T03 WIDE I
- DESIGN DEFICIENCY
--- 3 AVIS-9 ESSE - SEP 83 - SHUTDOWN O
P.0T.A.3437 : SI CNECK VALVE LE AKING - INC3MPLETE CONTACT BETWEE4 DISC 1 SEAT FARLEY Z - SEP $3 - REFUELINS
-~~- -. .~. ~ . _ . . . . . . . . - - . . . . -.
Gj
e PAGE 39 a
P.37.4.3447 : VEL ---44 CHECKESSE DAVIS-3 VALVE ANTI-R3TATION
- OCT $4 - ST3P JAM 9ING - DESIG4 PR03LE9 REFUELING e
P.07.A.3452 : EXCESSIVE SIT CHECK VALVE LEAK 4GE - SEAL PLATE C3CKED, V4.VE SEAT COMPENS4 TIN 3 JOINT 54LL GALLED FM ST. LUCIE 2 - DEC 34 - H3T ST4103Y gj P.07.A.3456 : EXCESSIVE SIT CHECK VALVE LE4KAGE - ETHYLENE PROPYLEME 3-TING 14TERIAL DEGRADATION CALVERT CLIFFS 1 - J44 85 - 133% P0hER
()
P.07.4.3457 : UNIT SHUTD3W4 - LOW ACCU 1JLATOR 83404 CONCENTRATI01 - LEA (IN3 VALVES
--- MCGUIRE 1 - APR $5 - 33% POWER 1
P.07.A.3463 : LPSI PUMP, St TAN ( 110PER48LE - MAINTENANCE, CNECK V4LVE .EAKAGE
--- PALISADES - N3V 85 - 781 POWER P.07.A.3474 : HIGH SI TANK LEVEL, LOW 83R01 CONCENTRATION - PRIMARf C3CLANT .*
LEAK 4GE THROJ3H SI CHECK VALVE PALISADES - N3v 85 - 78% POWER ,
P.07.A.0485 : RCS --- 40T ISOL48LE FR01 LP COOLANT SYSTEMS - PCIV LEAK 4GE PWR$ IN 3ENERAL - JUN 86 ((8.37.D.406)) -
P.07.8.3167e: IDDINE VALVES REMOV4L LEA <E3,SYSTEM INOPERABLE - N40N TANK DILUTED - CNECK
$1GNT GLASS I14CCUR4TE ST. LUCIE 1 - FE8 86 - 130% P3 DER P.07.C.3126 : RCFC --- MOTOR HE HOUSIN3 DR4!N CHECK VALVES INSTALLED 84CKJ4RDS I#
IION 2 - NOV 85 - REFUELING P.07.0.3061 : C3NTAINMENT ISOLATION VALVES LEAKED - SE AT ADJUS TMEN T, SIRT, JACKING SCREd ADJUST 1ENT ST LUCIE 1 - 14R S3 - REFUELING '
P.07.0.0063 : CIV LEAK RATES EXESSIVE - DEBRIS, SE4 TING SURFACE DE3 RAT!3N, s G BROKEN PLUNGER SPRIN3 $ PIN NORTH ANNA 2 - May 83 - REFUELIN3 J
P.07.D.3065 : CONTAINMENT 3YPASS LEAKAGE P4TH MECH 4NICAL PENETRATI31 V46VES LEAKED - PERSONNEL ERROR, GENERAL WE4R, DIRTY COMTACTS, F3 REIGN ,
MATERIAL 1CGUIRE 2 - JUN 1 JUL 83 - C3LD SHUTD3WN P.07.D.0073 : CIVS LEAKED - VARIOUS CAUSES f'
--- CONN TANCEE (HADDAM NECK) -
J41 TO MAR 83 - kEFUELI4G P.07.D.0076 : CIV LEAKAGE - FM, INTERNAL C3MPONENT FAILURE J
--- PT BEACH 1 - OCT 83 - REFUELIN3
.A P.37.D.30$9 : EXCESSIVE CIV LE4K4GE DURING LLRT - SEAT DAMAGE, P4CKIN3 .EAC
--- SALEM 1 - MAR-AUG 84 - REFUELI1G 4A P.07.D.0095t: EXCESSIVE CIV LEAKAGE - VARI 3US C4USES
, --- CONN. YA4KEE (NADDAM 1ECK) - J4N 86 - SHUTD0dN P.07.E.0168 : FIRE PUMP FAJLTY - LEVEL SWITCH, CNECK VALVE CORRODED
()
--- 1AINE YA1KEE - APR 83 - 100X P3WER O
i FAGE A0 P.08.A.3689 : BAST'S OUT OF SPEC FOR 27 HR - RMd LEAKAGE, SAMPLING PR3CEDURES
--- GINNA - SEPT $3 - 103% POWER P.09.A.3729 : CHARGING PUMPS IN3PERABLE -
DISCHARGE CHECK VALVE FAILE3, FM IN DISCHARGE RELIEF VALVE
--- SAN ONOFRE 3 - APR 84 - 100% P3WER '
P.05.A.3747 : 80RIC ACID C3NCENTRATIONS BELOW SPECS - CHECK VALVE LEAKE3
--- SALEM 2 - JUL 83 - 20t P3WER g)
P.05.A.3757 : CHAR 31NG PUM35 GAS BOUND, VCT OUTLET CHECC VALVE STUCC CL3 SED -
PULS4 TION DA1PENER BLADDER LEAKED ()
--- PALO VER3E 1 - FES 85 - 100% P3WER P.09.A.3764 : ADDITION T3 VIII.A.757 RE: CHARGING PUMPS GAS SOUND, WCT 30TLET CHECC VALVE STUCK CLOSED - PULSATIO1 DAMPENER BLADDER LEAKED -
MONTHLY PM I1ITIATED
--- PALO VER3E 1 - FEB 86 -
100% P0 DER P.05.8.3569 : C3NTAINMENT --- SPRAY RHR CHECK WALVE EXCEEDED ALLOW 48LE LEAK RATE -
COOK 1 - AUG S3 - REFJELING j P.0$.8.0574 : CHARGING PUMP SW PUMP DISCHARGE PRESSURE PROBLEMS - -
j INSUFFICIENT NPSH DUE TO SYSTEM DESIGN, CHECK VALVE SEATI1G j FAILJRE E CL3GGED STRAINERS DUE T3 F1 IN SW SURRY 1 8 2 - JUN TO 3CT S3 - 103% PodER & COLD SHUTD3WN P.38.8.3609 : SEAWATER ---
PUMP DISCHARGE CHECT VALVE FAILE3 - CORROSI21 CRYSTAL 41VER 3 - MAY 84 - 94% P3WER to P.03.8.3634 : CCW SUPPLY CIV LEAKAGE EXCEEDED LIMIT - STICKING DISC - SJSHING HAD SHIFTED TOWARD SdING ARM
- 77. BEACH 2 - APR $3 8 OCT 84 - REFUELING '
P.05.B.3639 : CCd 8 UMPS TRIPPED - M370RS FLOODED - SW LEAKED THR3 UGH SW PIPE #
, OPENING g
--- s INDIAN PT. 2 - AUG 84 - COLD SHUTDOWN P.03.8.0648 : Sd REALIGNED TO STAND 8Y Sd P3ND - PU1PHOUSE PIT ISOLATI34 VALVE #
FAILED TO OPEN - TORQUE SdITCH SET L3W, P351 TION INDICATI3N MALFUNCTIONE3 CATAdBA 1 - APR 85 - COLD SHUTD0dN '
P.08.C.3033 : POTENTIAL DA1 AGE TO REDUNDANT SAFETY EQUIPMENT FROM SACCFLOW THROUGH EQUI
- MENT AND FLO3R DRAIN SYSTEMS CALVERT CLIFFS 1 8 2 --- PWA'S I4 GENERAL - JUL S3 P.09.D.3297 : TURBINE / REACTOR TRIP - CONSERVATIVE DISCHARGE PRES $URE SWITCH J SETP3INTS - FW IS3LATION VALVE FAILURE, R3D CONTROL ALAR 9, SOURCE RANGE DETECTOR FAILURE SUMMER - APR 84 -
12% P0 DER '
P.07.D.3334 : REACTOR TRIP - LOW SUCTZ3N PRES $URE - PNEJMATIC TRANS9ITTER FAILED, TUR9INE CONTROL SYSTEM DI33E FAILED - SG PORV E T34FP 00 DISCHARGE CHECK VALVE FAILURES - PRESSURE INSTRU9ENT P.37.0.3355 : REACTOR TRIP DN HIGH- SG LEVEL - Fd PJMP SPEED CONTROL FAILED - I CHECC VALVE STUCK CONTROL DIL IMPJRITIES dATERFOR3 3 - JUL 85 - 5St P3 DER Oj t
9 PAGE 41 s
P.07.H.3761 : HYDR 3 GEN $4MaLING PUMP SEAL DATER CHECK VALVE INTERNALS REQUIRED CLEANIN3 COOK 1 S 2 - APR & JU1 83 - 133% POWER ,
P.07.H.3795 : C3NTROL ROOM BOTTLED AIR PRESSURIZED SYSTEM PRESSURE BEL 3d LIMITS - HEA3ER LEAKAGE, LOMPRES$3R FAILURE, INADVERTENT SI gj 10RTP AN1A 1 & 2 - MAY, 3CT 83 - 100% POWER P.07.H.3907 : SINGLE FAILU4E SEISMIC CRITERION 13T MET 3Y 2 HR 8ACCJP INSTRUMENT AIR SYSTEM - DESIGN ERR 3R THREE MI.E 15. 1 - MAR 85 - C3LD SHUTD0JN O
P.11.A.3688 : CHECC VALVE FAILURES IN 03 RAW WATER COOLING SYSTEMS - INTER 1AL DISASSEMBLY - ABRASION, CJRR3SION
--- 8hR'S IN GENERAL - MAR 83 ((S.11.A.4933)
P.11.A.3F92 : DGS IN3PERABLE - CONSTANT VENT CHECK VALVE SINDING, DISTRI8UTOR HAD EXCESSIVI susHING & R3 TOR WEAR '
FARLEY 1 & 2 - APR & AUG S3 - 103% POWER P.11.8.3561 : LOSP - TRANSFORMER FAILURE - CA8LE FAULT - INSULATION DEGRADED - #
AGE, ENVIR31 MENTAL DAMAGE PWRS IN 3ENERAL - JUN 86 --- SAN ONOFRE 1 - NOV $5 *
((B.11.B.5263) ,
P.15.A.3455 : POTENTIAL FOR PIPING SYSTEM L PIPE SJPPORT DAMAGE - PIPE BENDING - TEMP DIFFERENCES I1DUCED Bf STRATIFIED FLOJ
--- PWR$ IN 3ENERAL - DEC 84 ((8.15.3003)
P.15.A.3457 : ADDITION TO XV.448 RE: SNUBBER SELF-ALIGNING BALL SUSHI1GS #
FAILED - SNU39ERS RESUILT
--- R3BINSON 2 - MAY B4 O
i 8.15.A.3455 : CONTAMINATIO1 0F 3REATHING AIR SYSTE15 - AIR COMPRES$3R$ FAILED, RADIDACTIVE GAS & F1 IN AIR SURGE AT SYSTEM STARTUP '
ROBINSON 2 - JUL S3 - COLD SHUTD3WN --- PWR$ IN GENERAL - -
g JAN S5 (C I P.15.A.3468 : LIMITORQUE ACTUATOR WORM SHAFT GEAR FAILURES - IMPACT LOADING #
0F CLUTCH MECHANISM DURIN3 M3DE CHANSES AT HIGH SPEED PWR$ IN 3ENERAL - AUG 85 --- DTHER UNITS AS LISTED I4 ARTICLE ((8.15.A.3103) ,
j P.15.A.3472 : LIMITORQUE M3 TOR OPERATOR LITHIU1 BASE LUSRICANT N3T ENVIRONMENTALLY QUALIFIED - CALCIU1 BASE GREASE TO BE USE)
STRON 1 - JUL 85 - C3LD SHUTD3dN {
I I
P.16.C.1890 : PORV FAILED TO FULLY STR3(E - INSTRU1ENT AIR CHECK VALVE INSTALLED SACKWARDS
--- SURRY 1 - FEB 83 - 853 P3WER, SHJTDOWN j
P.15.C.2524 : AUX FW PUMP AUTC STARTS DJRING CHECK VALVE TESTING - LOJ SG LEVEL - WRITTEN 11STRUCTI3NS NOT !$5'ED J - ADMIN /PR3CEDURAL DEFICIENCY g
--- CATAWBA 1 - N3V $4 - HOT STANDSY P.16.C.2620 : CONTAINMENT SPRAY ACTUATI3N DURING TEST - INITIATION RELAYS NOT RESET - RCPS SECURED, SEAL DAMAGE - CCW PIPING PRES $3RE SPIKE
- l. --- dATERFORD 3 - FEB 85 - H3T STA105Y gg
t P401 42 P.15.C.2747 : ECCS TRAINS IMOPERABLE - CONTAINMENT RECIRC SUMP ISQLATIO4 VALVES NOT TESTED PR3PERLY -
PERS31NEL ER404 PALD VER3E 1 - A*R 85 - HOT SHUTDOWN ' ,
P.15.C.2813 : REACTOR, TUR3!NE TRIP, ESF ACTUATI3NS - HIGH MSR LEVEL - 1 EATER DRAIN TANC DRAIN PATHS IS3 LATED g;
CALLAWAY 1 - AUG 85 - 255 POWER P.16.C.2853 : REACTOR TRIPS - SPURIOUS CLOSURES 3F MSL EXCESS FLOW CHECC VALVES - RUPTURE DISC FAILURES - 13U1 TING 83LTS UNDERTORQJED 1AINE YA1KEE - OCT SS - 25, 335 80WER O
P.16.C.2933 : SI DURING ACCUMULATOR DISCHARGE CHECC VALVE FLOW VERIFICATION -
DISCHARGE VALVES OPENED - $$PS FUSES REM 0dED - PROCE3JRAL, PERS3NNEL ERR 3R l --- NORTH AN1A 1
- DEC 85 - COLD SHUTDOWN
............. ALL DONE, PRESS CRETURN> (EY TO G3 T3 MENU .................T:
THERE ARE 19 ARTICLE (3) T3 8E DISPLAYED: 19-4UG-1986 14:13:113 PWR - CHECC VALVES - RCS SYSTEM j
P.05.e.0018 : RC MAKEUP LINE CRACKS, THERMAL SLEEVE DAMAGE - FATIGUE FAILURE SUSPECTED CRYSTAL RIVER 3 - JAN 82 --- 3C01EE 1, 2 & 3 --- PWR'S IN #
GENERAL P.05.8.0020 : RCS CHECK VALVES LEACED #
MCGUIRE 1 - FE8 82 - HOT SHUTDOWN P.05.C.3047 : P3RV STUCK OPEN - SOLENDID OPERATED CONTR3L VALVE LOCCED IN #
ENER31IED STATE BY VENT RESTRICTI31 31NNA - JAN 82 - RAPID C30LD0dN O
P.05.C.0050 : PORVS INOPERABLE - LEAKIN3 DIAPHRA3M, AIR SUPPLY CHECC VALVE
--- SURRY 1 - OCT 82 - SHJTD3WN /
g P.05.C.0059 : RCS LEAKAGE - PORW SLOCK VALVE, SI CHECK VALVE, SG TUSE MILLSTONE 2 - MAR 83 - 130% P0 DER P.05.E.0063 : PORV LEAKAGE PROBLEMS, C31TR3L AIR ISOLATED, PCS VENT PAT 4 LDST
- RAND 0M MAI1TENANCE PROBLEM - PORV LEAKAGE PROBLE1 J1RES3LVED SALEM 2 - JAN $3 - C03LD3dN, SHUTDOWN P.05.F.0004 : PROBLEMS WIT 4 VALVE OPERATING SYSTEM
--- SHIPPINGPORT - 1960's >
P.05.F.0016 : CHECC VALVE INTERNAL 3 REPLACED WITH MORE DURABLE MATERIALS
--- YANKEE R3WE - AUG 74 - REFUELI16 SHUTD0dN #
P.05.F.0043 : PRIMARY COOLANT SYSTEM LEAK *3E - CHECK VALVES FLUSHED T3 IMPR3VE SEATING I J
--- PALISADES - JUL 54 - HOT STAND 3Y P.05.F.0052 : PLANT SHUTD0dN - EXCESSIVE RCS LEACA3E - WALVE PACCING, RCP b FLANSE LEA <E)
CATAwan 1 - OCT 85 - 1033 P0 DER
)
P.06.E.0594 : REACTOR TRIP DN 9AIN FW PUMP TRIP -
FW PUMP RECIRC VALVE FAILED TO h4hDLE FL3d TRANSIENT -
CONTR3LLER DESIGN PR03LEM - SG PORv5 FAILED TO OPEN - DUT 3F CALIARATI31 52
e PAGE 43 CATAWBA 1 - JUN 85 - 64E POWER P.37.A.3457 : UNIT SHUTDOW1 - LOW ACCU 13LATOR B3401 CONCENTRATION - LEACIN5 VALVES 1CGUIRE I -
APR SS - 335 POWER tr P.07.4.3474 : HIGH SI TANK LEVEL, L3W 93401 CONCENTRATION - PRIMARf C30LANT LEACASE THROJGH SI CHECK VALVE
'ALISADES - 43V $5 - 981 POWER C)
P.37.A.3485 : RCS ---
NOT IS3LABLE FROM LP COOLANT SYSTEMS - PCIV LEAKASE PWR$ IN 3ENERAL - JUN 86 ((8.31.D.406)) ()
P.37.D.3073 : CIVS LEAKED - VARIOUS CAUSES CONN YAN(EE (MADDAM NECE) -
J41 TO MAR 83 - REFUELING ~
P.07.0.0389 : EXCES$1VE CIV LEACAGE DURING LLRT -
SEAT DA.14GE, PACCIN3 LEAC SALEM 1 - 1AR-4UG 84
- REFUELING P.15.C.1297 : CHECC ---
VALVE CL3SURE WEIGHT MISSING NORTH ANNA 1 - MAR 81 - SHUTD0dN j P.16.C.1890 : P3RV FAILE3 T3 FULLY STR0(E - INSTRU1ENT AIR CHECK VALVE -
INSTALLED SACCWARDS SURRY 1 - FE5 83 - 855 P3WER, SHJTDOWN P.15.C.2520 : CONTAIN1ENT SPRAY ACTUATI3N DURING TEST - INITIATION RELAf5 NOT RESET - RCPS SECURED, SEAL DAMAGE - CCW PIPING PRESSJRE $8IXE 4ATERFOR) 3 - FE3 85 - H3T STA103Y
............. ALL DONE, PRESS (RETURN > (EY TO GO T3 MENU .................f: O THERE ARE 95 ARTICLE ($) TO BE DISPLAfED: 19-4U3-1986 14:15:43 PWR - CHECC V4LVES - SI/RECIRC/ CONT. SPRAY P.36.8.3389 : CRD ---
3UIDE TulE SUPPORT PI1 FAILED - SGS DAMAGED JITHIM H3JR$ #
dESTIN3H3USE PWR'S IN GENERAL - JUL 82 - N3 P.05.8.3018 : RC MAKEUP LINE CRACKS, THERMAL SLEEVE DAMAGE - FATIGbE FAILURE SUSPECTED
--- CRYSTAL RIVER 3 - JAN 8 2 --- OC 01E E 1, 2 8 3 --- P W R' S I N GENERAL P.05.C.0059 : RCS LEAKAGE - PORV BLOCK VALVE, SI CHECK VALVE, SG TJ3E 11LLSTONE 2 - MAR 83 - 130t P3 DER '
.s P.07.4.0013 : CHECC VALVE ASSEMBLED INC3RRECTLY PALISADES - MAY T2 J
P.07.A.0020 : INDICATIONS IN VALVES 14INE V41KEE - SUMMER 1972 (AFTER HOT FUNCTI3NAL TESTING) 3 J
P.07.A.3025 : WELD SLAG UN)ER VALVE SEAT - SI TANK CONCENTRATION L34
--- 1AINE YA1(EE - DEC T2 Of P.07.A.0032 : CHECC VALVE SOFT SEAT FAILURES TURCEY PT 4 - MAT 73 O
P.07.A.0060 : CRACC INDICATIONS IN VELAN CHECK VALVES CALVERT CLIFFS 1 - SEP T3 (FUNCTIONAL TESTIN3)
O
e
. PAGE 44 P.07.4.3063 : CHECK VALVE LEAKASE -
ACCUMULRTOR DILUTION
$1NNA - SEP 74 i
P.07.A.0085 : CHECK VALVES DID NOT SEAT, PROCEDURAL PROSLEM - 90R01 CONCENTRAT!01 LOW SURRY 1 - AUG 75 - HOT STAND 8Y 48 P.07.A.0105 : LEAKING CHECC VALVE - ACCJMULATOR 30RON L3W ROSINSON 2 - J AN 76 - 103% P3WER ()
P.07.A.3114 : LEAKING CHECC VALVES CAUSED 50RON DILUTIO1 IN SI ACCJ1ULATOR SURRY 1 - JUL 76 - 133% POWER ()
P.07.A.3120 83RON DILUTI3N IN SI ACCU 1ULATOR CAUSED Sf LEAKING CHECK WALVES SURRY 2 - AUG 76 - 133% POWER j P.07.A.0122 : SACK LEAKAGE IN ACCU 1ULAT3R DISCHARGE LINE CHCCK VALVES ZION 1 - JUN 76 - 50% P0 DER P.07.A.3126 : WRONG SIZE GASKET INSTALLED IN ACCUMJLATOR CHECK VALWE
--- ZION 2 - 3CT 75 - 30% P0 DER '
P.07.A.3143 : IMPR3PER SEATING 3F SIT DJTLET VALVES CAUSED DILJTION OF 3AT'S -
ZION 2 - MAR 77 - 50% P0 DER, APR 77 - 48% P0 DER ^;
P.07.A.3172 33LTS & BRACCET F3R CHECK VALVE DISC MIS $1NG
--- INDIAN PT. 2 - MAY 78 (L3d POWER PHYSICS TESTS) #
P.07.A.0175 : TILTING DISC CHECK VALVE INSTALLED I4 WRONG ATTITUDE SAN ON3FRE 1 - MAY TS - APPR3XIMATFLY 55% PodER p P.07.A.0182 BORON CCNCENTRATION LOW -
SI TANK CHECK VALVES LEAKED
--- CALVERT CLIFFS 2 - SEP 7S - 983 POWER ()
P.07.A.3187 : STOP CHECK DISC SPRING J AMME) - SPRI1G RETAINER ADDFD
ARKANSAS ONE 2 - JUL 78 - PRE 3P TESTING g)
P.07.4.3200 : CHECC VALVE LEAKED - VALVE 83DY OdERSIZED ARKANSAS ONE 2 - JAN 79 - 19% P0 DER (POWER ESCALAT!3N #
TESTING)
P.07.4.3204 : INCORRECT VALVE WEIGHTS USED IN PIPE STRESS CALCULATI3NS -
HANGERS MODIFIED NORTH AN1A 1 - MAR 79 - 92% POWER P.0i.A.3225 : SI TANK BORON CONCENTRAT!3N LOW
- CHECK VALVE LEAKED MILLSTONE 2 - APR 77 - 9)E POWER P.07.A.3229 : SAT SORON C01 CENTRATION L3W - VALVE LCAKED - LIMIT Sd!TCHES READJUSTED ZION 2 - 3CT 79 - 94% P0 DER ' >
P.07.A.3230 : SWINS CHECK WALVE BEARING CAP 0-RING FAILED 1AINE YANKEE - OCT 79 - 97% P0 DER g P.07.A.3240 : HPI CHECK VALVE SEAT HOLD-D0dNS F3JND LOOSE THREE MILE IS 1 - FE3 dd - SHUID3WN g)
- .'.f.A.3256 : C3NTAINFsNT PENETRATION C4ECC VALVE SEATS LEAKED
--- CONN YANCEE (MADDAM NECK) - MAY SD - REFUELING E)
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. PAGE 46 P.07.A.3316 : ADDITION To #11.4.243 RE: TILTING DISC CHECC VALVES
--- THREE MILE IS 1 -
JU1 81 - COLD iMUT33W1 P.37.A.0339 : CHECC VALVE )lSC MISSIhG - C3NSTRUCTION ERROR COOK 2 - OCT 81 - SHUTDOJN g, P.07.A.3341 : SI ACCUMULAT3R BORON CONCENTRATION HIGH - VALVES LEACED -
PROCEDURES REVISED ()
NORTH AM14 1 - N3V 81 - 79% P3JER P.07.A.0343 : LEAK!hG SI SYSTEM CHECK VALVES SEATED FOLLOWING FLUSHING PT 8EACH 1 - OCT 81 - dEFUELING P.07.A.3345 : $1 ACCUHULAT3R BORON CONCENTRATION HIGH - VALVES LEACEO, FAULTY PROCEDURE NORTH AN14 2 - OCT 81 - 971 PadER P.07.A.0346 : ADDITION TO WII.A.331,335 & 337 RE: ACCUMJLATOR 50404 CONCENTRATI015 10RTH AN14 1 & 2 - SE8 81 - 133% POWER P.07.A.0352 : 91T INLET VALVE FAILED - SHAFT PACCING 703 TIGHT j
--- SURRY 1 - DEC 81 - COLD SHUTODJN P.07.A.3363 : LPSI SWING C4ECK VALVE WE4R, DAMAGE - DESIGN
--- PALISADES - SEP S1 - REFUELING P.07.A.0372 : HP I1JECTION STOP CHECK v4LVES STUCK - UP TO 580 PSID REQJIRED g, 70 ESTABLISH FLOW
--- DAVIS-9 ESSE - JUN 82 - REFUELING
' P.07.A.3375 : ADDITION TO WII.A.372 RE: HP INJECTI3N ST3P CHECK VALVE STUCC -
INCORRECT SEAT ANGLE 3 AVIS-3 ESSE - JUN 82 - REFUELING --- PWRS IN GENERAL - AUG /
82 ,
EE P.07.4.3384 : SIT 3UTLET C4ECK WALVE LE4KED RINGS DETERIORATED CALVERT CLIFFS 1 - JUL 82 - 3T4RTuP #
P.37.A.3386 : COLD LEG INJECTION CHECK #ALdE LEACED - CAR 8ON STEEL CL3SJRE STUDS CORRODED NORTH AN14 1 -
SEP 82 - REFUELING P.07.4.0390 : ADDITION TO WII.A.275 RE: VELAN CHECC VALVE SINDING PROBLEMS -
DISCS REPLACED SEAVER VALLEY 1 - OCT 82 - SPECIAL TESTING P.07.A.0392 : 3I CHFCC VALVES STUCK - DISC STUD PR3TRUDED A80VE NUT, DISC M I S ?.t ;G N E D 4RKANS45 01E 2 - DCT S2 - REFUELING P.07.A.0396 : MINOR LEACAGE INT 3 SIT C31P00NDED 3Y SIT LEVEL INDICATING g SYSTEM FAILU4E CAUSED TECH SPEC LI11TS 70 BE EXCEEDED - V4LVE LEAK 4GE, TEH8 EFFECTS ON 4EFERENCE LEG
--- PALISADES - SEP-DEC 82 - 100% P0 DER
()
P.07.A.3397 : 51 TANK LEVEL LOW, 83RON CONCENTN4713N LOW - V4LVE LEAKAGE PALISADES - N3V 82 - 100% POWER ggl
< l phGE ET l
c P.3T.A.3400 : ADDITION TO WII.A.399 RE: CHECK VALyt DISC MIS $1NG
--- COOK 2 - DEC 82 - REFUELING i
P.37.A.3401 : SI C4ECK VALdE SEAL dELD LEAK - INADEQUATE APPLICAT!34 3F dELDING AMD GRINDING TECH 410UES 9AINE YA1KEE - FEB 83 - SHUTD0dN gj l
P.07.A.0403 : ACCU --- 9ULATOR 30R01 CONCENTRATION L3d - CHECK VALVE LEACE3 SURRY 2 - SEP 82 - 103% POWER ()
i P.07.A.3407 : COLD LEG I4JECTION ACCUMJL AT3R CHECK VALVES LE AKED -
1CGUIRE I - MAY $3 - STARTUP, 35 5 50t POWER (}
P.37.A.0420 : HPI- STOP CHECK VALVE STUCK - DISC TO SEAL CONTACT AREA T33 WIDE DESIGN DEFICIENCY y
--- DAVIS-3 ESSE - SEP 83 - SHUTD3d1 P.37.A.3422 : ADDITION TO VII.A.384 RE: SIT DUTLET CHECK VALVE LE ACED '
RINGS DETERI3 RATED - 9ATERIA. CHA43ED
--- CALVERT CLIFFS 1 5 2 - JUL 82 - STARTUP i
P.07.A.3437 : ST ---
CHECK VALdE LEAKING - INC3MPLETE CONTACT BETWEEN DISC L SEAT FARLET 2 - SEP 83 - REFUELING '
P.37.A.3447 : VELA
--- 4 CHECK WALVE ANTI-ROTATION ST3P JAM 4XNG - DESIGN PROBLE4 DAVIS-BESSE - OCT 84 - 2EFUELING P.07.A.3452 : EXCESSIVE SIT CHECK VALVL' LEAKAGE - SEAL PLATE C3CKED, WALVE SEAT C3MPENS ATING JOINT 3 ALL G ALLED -
FM
--- ST. LUCIE 2 - DEC $4 - H3T STAND 3Y g, P.QT.A.3456 : EXCESSIVE SIT CHECK VALVE LEAKAGE - ETNTLENE PROPYLENE 3-RING MATERIAL DE3RADATION CALVERT CLIFFS 1 - JA1 85 - 103X POWER
)
4
/
P.07.A.3457 : UNIT SHUTDOW9 - L3W ACCUMJLATOR 80401 CONCENTRATION - LEACING VALVES ,
4 1CGUIRE 1 - APR 85 - 33% POWER
,)
P.07.A.0468 : LPSI---
PUMP, SI TANC INOPERABLE - MAINTENANCE, CHECK VALVE LEAKAGE PAi!SADES - N3V 85 - 7d% POWER J
4 P.07.A.3474 : HIGH SI TANK L E '.'E L , L O W 83R04 CONCENTRATI3N - PRIMART C33. ANT 1* LEAKAGE THROUGH SI CHECK WALVE PALISADES - N3V 85 - 78% POWER P.07.4.3485 : RCS NOT ISOLASLE FRO 1 LP COOLANT SYSTEMS - PCIV LEAKAGE
, --- PWR$ I4 3ENERAL - JUN 86 ((B.37.D.406))
i ' P.07.8.0012 : CRACKED ROCK 3 HAFTS AND CRACKED 80DT IN CHECK VALVES SURRY 1 1 2 - NOV 74 ! ~#
P.07.8.0053 : CTS 10ZILES LEAKED - VALVES MOT SHUT TIGHTLY - CHECK VALVES INSTALLED SACKWARDS b COOK 2 - NOV 78 - SHUTDOJN P.07.8.0068 : SPRAY PUMP M3 TOR STATOR C3ILS/ PHASE LEAD CHAFED, $NORTE3 (
CALVERT CLIFFS 1 - SEP 77 - 103I POWER P.07.8.00$0 : CHECK VALVE 3ISC AND RELATED PARTS N3T INSTALLED
I StGE 48
--- s SEQUOYAH 1 - APR 80 - PREOPERATI3NAL P.07.8.3084 : 9 DST LEVEL L3d - RSS SYSTEM VALVE FAILED TO RESEAT, ALARM FAILED e
--- 3CONEE 3 - APR 83 - HEATUP P.07.8.3123 : CONTAINMENT
--- SPRAf ADDITIVE TANC CHECC VALVES LEACED RANCHO SECO - APR 82 - SHuTDod1 P.07.8.0131 : CONTAINMENT SPRAY $ N40H ADDITIVE SYSTEM CHECK VALVE FAILED -
SEAT & DISC 44CHINED 3INNA 1 - SEP S2 - 103% POWER P.07.8.0134 : SPRAf PUMP CHECK VALVE PIN /$d!NG ARM WORN O
--- 3INNA - JUN 82 - 100% P0 DER s
P.07.8.3142 : CONTAINMENT ARM dORN SPRAY PU9P DISCH4RGE CHECK VALVE FAULTY - PIN / SWING 3INNA - JUL 82 - 100% P0 DER P.07.8.3167#: 10 VALVES DINE REMOV4L LEA <E3, SYSTE9 IN3PER48LE - N40H T4NK DILUTED - CHECK SIGHT GLASS INACCUR4TE ,
ST. LUCIE 1 - FES 86 - 130% P3 DER P.37.0.30958: EXCESSIVE CI4 LEACAGE - V4RIQUS CA'SES J
--- C0hN. YANCEE (HADDAM 1ECC) - J4N 86 - SHUTD0dN P.07.E.3237 : CORR 3510N FAILURES OF 413 SS VALVE STEMS - IMPROPER HEAT
TREATMENT
--- PbR$ IN 3ENERAL - JUL 85 --- 3C01EE 1 - DEC 71 --- FARLEY 1-FE5 84 ((9.37.F.271))
P.35.A.0230 : CHECC V4LVE LEAKED - RELIEF VALVES, FLOW TRANSMITTERS 04M4GED -
CHECC VALVE REWORCED DAVIS-9 ESSE 1 - JUL ?? (4FTER INITIAL FUEL L3ADINS) - HOT
(
STAN38Y f P.05.4.0459 : dRON3 INTERN 4LS USED IN VALVE s dI
--- NORTH 4N44 2 - SEPT 83 (PRIOR 70 POWER OPERATION) - C3LD SHUTD0dN P.05.8.0465 : dELD ---
CRACKS FOUND FOUND 11 83 RATED W4TER PIPING IAN ON3FRE 1 - SEPT 77 - 1003 PodER P.05.C.0033 : PDTENTIAL DA1 AGE TO REDUNDANT SAFETY EQUIPMENT F401 84CCFLOW THROJGH EQUIP 1ENT AND FL33R DRAIN SYSTEMS CALVERT CLIFFS 1 & 2 --- PWR'S IN GENERAL - JUL $3 P.07.H.0798 : CONTROL ROOM 80TTLED AIR PRESSURIIED SYSTEM PRESSURE BEL 0d #
LIMITS - HEA3ER LEAKAGE, COMPRESSOR FAILURE, INADVERTENT SI
--- NORTH ANMA 1 $ 2 - MAf, 3CT 83 - 100% POWER t
P.16.C.1776 : BAAT---
PUMP DISCHARGE VALVE FOUND CL3 SED - PROCEDURE ERROR DAVIS-8 ESSE 1 - MAY 82 - COLD SHJTDOW1 0
P.15.t.2747 : ECCS TRAINS IN0PERA8LE - CONTAINMENT RECIRC SUMP ISOL4 TID 1 VALVES N3T TESTED PR3PERLy - PERS31NEL ERROR
- ALO VER3E 1 - APR 85 - H3T SHJTD0dN g)
P.18.C.2933 : $! DJRING ACCUMULATOR DISCHARGE CHECC VALVE FLOW VERIFICATIO1 j' DISCHARGE VALVES OPENED - SSPS FUSES REMovfD - PROCE3JRAL, IO
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l - - - - - - -
1 NPE I PLANT I PLANT I I EVENT I FAILURE I I NUMBER I (DATE) 1 STATUS 1 SYSTEM 1 DESCRIPTION I MODE I I I I i i i I I I I I I I I I V. D. 20 1 McGuire 1 Hot 8 RCS 1 Three RCS pressure isolation check IL I I I (02/02) l Shutdown i 1 valves failed leak rate test (failed to a i I I I I I seat properig). I I "
4 I I I I I I l V. C. 59 i Millstone 2 1 100% Power i SI I Metal gasket replaced on one check IL 8 1 I (03/83) 1 I I valve. 1 I O I I I i 1 I I I V.C.50 i Surry 1 i Shutdown i PZR I Air suppig check valve fer PDRVs IL 1
! ! (10/82) 1 I I leaked. 1 I O I I I I I I I I VI.D.52 1 Surry 1 I I RCS 1 The check valve in the bypass line IL I l I (11/82) 1 I I around the trip valve of the main i I l i I I I I steam line C leaked--flaw in the carbon i I I I I I I steel valve body. I I I I i 8 I I I I VI.D.55 i Kewaunee ! Construc- 1 MS 1 Cracks in isolation check valve seats IL I I I (Late 1972) I tion i I (two valves) (discovered during i I 1 I I i 1 quality records audit). I I I I ! I I I I I VI.D.55 I Zion i Construc- I MS I Cracks in two of the four MSI check IL 1 I I (Earig 1973) I tion i I valves. 1 I I I I I i 1 I I VI.D.60 I Zion I Cold i MS I Weld cracks in main steam check valve IL I 1 I (09/73) I Shutdown I ! disc stops (three of twelve valves). I I I I I I I I I I VI.D.73 i Maine Yankee i I I Various problems in Shutte & Koerting iF I I I (Late 1973) i I I excess flow check valves. I I W I I I I I I I I VI.D.73 i Surry 1 1 I I Various problems in Shutte & Koerting iF I o
i I (1973) l I I excess flow check valves. 1 I O l I I I I i 1 1 VI.D.78 i Fourt Calhoun 1 Power i MS I Tain-walled check valves. I - I I I (04/74) l I I I '
I . O I I I I I I I I VI.D.148 I Trojan I Cold i MS I Main steam line check valve shaft. 8 - l '
1 I (08/76) I Shutdown i I I I J B I I I I I I I VI.D.264 I Robinson i Refs,eling i MS I Check valve rock shaft displaced I- t ,
o i I (06/82) l I l asially--set screw loosened. I I "
i I I I I I I I VI.D.274 5 Sequoyah I Refueling i MS I Check valve disc / stem separation IF l D I I (11/82) i I I (cyclic fatigue failure). 1 I -'
I 8 I I I l l I VI.D.294 1 1 I I Additional information to VI.D.274DK. IF I
- I I I I I I I I VI.D.290 i Trojan ! ? 8 MS I Main stream nonreturn check valves iF 1 1 I (03/83) 8 I I stuck open due to increased friction i I
" i I I 8 8 due to overtightening of the packing i l i i I i I gland (four valves affected). I I I I I I I I I b I VI.D.303 I North Anna 1 0% Power 1 MS S Steam generator main steam trip valve I FC I b I I (09/83) 1 I I closed slow--corrosion on control i I i 1 1 8 I venting check valve. I I g
b I I I I i i 1 Page 1 0
I VI.D.305 i TreJsn I - I- 1 Additian to VI.D.290. IF I t I I I I I I I VI.E.22 i Surry l Power i FW I Check valve gasket failed (interesting IL I I I (10/72) 1 I I event). I 1 1 I I I I I I I VI.E.29 i Surry 2 1 - 1 FW I check valve in feedwater line a leaked iL i I I (05/73) I I I through or flaw in the valve body 1 I 1 i I I I (small leakage). I i -
I I I I I I I I VI.E.106 I Palisades I ? I FW I Feed water check valve gasket lea k ed. I FD 1 1 I (<1974) 1 I I I I O 8 I i 1 1 1 I I VI.E.128 i Robinson I 100% Power i FW I check valve failed to hold? (Burr on 1 F l I I #08/77) 1 I I hinge prevented proper operation). I I O 8 I I I I I I I VI.E.155 1 San Onofre 1 1 Startup i FW I Feed water check valve pin broke I FC 1 1 I (04/76) 1 I I allowing the disc to fall into the ! I l I I I I I bottoming of the valve. 1 I I I I I I I I
~
l VI.E.171 1 Trojan i Hot i FW I CST check valve stuck open (no cause 1 WP 1 1 I (07/78) I Shutdown ! I identified). I I I i 1 1 I I I I VI.E.185 I Fourt Calhoun i Refueling i AFW I check valve found to be oriented in i FC i i I (11/78) l I I the wrong position. I I I I I I I I I i I I I I I I I VI.E.278 8 Crystal River 31 Cold i FW 3 Check valve failed to close--steam iL i I I (04/80) 1 Shutdown i I generation feed water check valve disc I 1 I I I I I retainer pin missing. I I I I I I I I I I VI.E.286 I Oconee 2 3 48% Power i FW I Steam leaked from feed water check iL I I I (06/80) 1 8 1 valve. I I &
I 1 l t 1 1 1 I VI.E.319 I Cook 2 1 72% Power i FW l Feed water check valve leaked (seat i FC I I I (01/81) l I I leakage). I I O I I I I I I I I VI.E.325 8 Cook 1 1 100% Power i AFW I Auxiliary feed water check valve stuck I L I q
I I (02/81) i I I ope (caused by dirt). I , 8 O
I I I I I I I I VI.E.331 i Sequoyah 1 1 Cold I FW l Feed water check valve leaked-seat / IL I s I I (03/01) l Shutdown I i disc reguired lapping. 1 I J l 1 I I I I I j i VI.E.340 1 Turkey Point 3 i Refueling i NFW I Main feed water check valve disc stud iF I
- ~
I I (04/81) l I I nut missing due to the failure of its I I
, I I I I I associated locking device. 1 I I I i i i I I
' 8 VI.E.362 I Divis Besse 1 1 Startup i FS I Auxiliary feed pump check valve leaked. IL I
- I I (03/00) I I I I l
, I I I I I I i 1
' I VI.E.370 1 Point Beach I >79% Power i FW I Feed water pump discharge check valve IL l -
i i I (11/80) 1 I I failed. 8 I I I I I i l I b I VI.E.378 i Surry 1 i Startup 1 FW I Check valve Feed water-89 disc I FC 1 I I (04/80) i l I retaining nut cotter pins failed--valve i 8 I I i 1 I stuck open (improper material). I 1 g
\"
I i 8 I I I i i VI.E.378 i Surry 1 i Startup 1 FW I Check valve feed water-27 disc was 1F I I I (01/80) 8 I I found detached (cause unknown). I I
, 1 I I I I I i
' " 8 Page 2
i I VI.E.386 i Ntrth Anna I 100% Pawst 1 FW I Stoco supply check valve perted Isogsd iF I 1 I (12/01) 1 I I in aus FW pume trip valve--disc not 1 I I I i 1 I retaining pin installation suspect. 1 I I I I I I I I 3 VI.E.398 i Cook 1 1 100% Power i FW l Auxiliary feed water pump leakoff line i FC I I I (04/02) I I I check valve failed to seat completely. 1 I I I I I I I I 1 VI.E.402 I Turkeg Point 4 1 Startup I FW I Feed water check valve stud nut I F 1 d
i I (06/02) i I I fastener failed. I I I I I I I I I g
t i 1 8 8 1 1 I VI.E.412 Arkansas One 2 i Refueling i EFW I EFW pump turbine steam suppig check i F i 1 I (10/82) I I i valve failed--the dtsc stud was broken I l
~
l i I I I avid missing. I I I
I I I I I I I I VI.E.417 i Salem 2 1 82% Power i EFW 4 EFW pump steam supply check valve disc I F l 1 I (10/82) l I I stud broken, nut and cotter pin i I J l i I I I missing. I I I I I I I I I
~
I Crystal River 31 Refueling i FW l Feed water air accumulator check valve IL l l VI.E.444 1
! I (03/83) l I I leaked due to water in the instrument i I I I I I I air system. I I 1 I I ! I & I l' I VI.E.481 1 Farley 1 I 100% Power i FW I MDAFWP discharge check valve leaking i L I I I (12/83) 1 I I hinge pin bushings missing, worn. 8 I 1 I I I I I I I VI.E.495 i McQuire 2 1 25% Power i FW l SQ inlet check valve stuck open (cause i FC I I I (05/84) 1 I I not known). I I
- l' I I I I I I I I VI.E.524 i Millstone 2 1 62% Power i FW l Binding check valve stem--feed water i FC 1 I I I I I check valve to SO 2 failed to seat. I I I I I I I I I ##
1 VI.E.524 i Millstone 2 1 62% Power ! I Binding check valve stem. IF l i I (11/84) 1 I I I I 1 I I I I I I O I VII.A.32 1 Turkey Point 4 I - 1 SI I Seat failure (safety injection high I L I I I (05/73) i I I head lines to RCS cold leg)- "0" ring i 1 g
I I I I I missing (ma s. leak 1/3 gpm) I , 1 I I I I I I I I VII.A.60 1 Calvert ! - I SI I Crack in check valve body. 8L 1 .
" d i I Cliffs 1 1 I I I I i i I I I i l i t
i VII.A.63 I 01nna i 10%/Hr. I SI I Check valve in loop B cold leg IL I I I (09/74) 1 Power Red. I I leaked--accumulator dilution. I I i
! I l 8 I I 1 I I VII.A.85 i Surry 1 I Hot i SI I Check valve associated with l FC i l ' I I (08/75) 1 Shutdown l I accumulator IC did not seat. I l s l I I I I I I I VII.A.105 I Robinson 2 1 100% Power i SI I Check valve associated with I L I
\" I (01/76) l I I accumulator B leaked. 1 I " '
1 8 I i l i i I I VII.A.114 i Surry 1 1 100% Power i GI I Two check valves associated with IL I
\" i I (07/76) 1 1 1 accumulator B leaked. I I '
I I i i i I I I VII.A.120 l Surry 2 1 100% Power i SI I Two check valves associated with IL I b I I (08/76) i I I accumulator C leaked. 1 I b-I I I I i I i i VII.A.126 I Zion 2 1 30% Power i SI I Wrong site. gasket installed in iL I f I I (10/75) l I 1 accumulator check valve (0.25 gpm) 1 I O Page 3 0
l i I I I (frctary sessably serse) (m). I 1 1 I I I I I I I VII.A.128 1 Zion 1 1 50% Power i SI 1 Accumulator 1D discharge line check I L l i I (06/76) 1 I I valve leaked. I I I I I I I I I I VII.A.172 I Indian Point 2 I Low Power i SI I #21 HH SI discharge check valve IL 1 l 8 I (05/78) l Test I I leaked--bolts and brackets for dise I I
! ! I I I missing. 1 I d
! ! I I I I I I VII.A.175 I San Onofre 1 55% Power i Si i Titting disc check valve installed in i FC 1 1 I (05/78) i I I vertical rather than horizontal pipe 1 I C I I I I I line - failed to close. I I I I I I I I I t VII.A.182 i Calvert i 98% Power i SI I Safety injection tank check valve !L I i I I Cliffs 2 I I 1 leak. I I I I I I I I I I VII.A.187 8 Arkansas One 2 i Pre-op i LPSI I Check valve on pump discharge failed to IF 3 e
! I I Testing i I close (disc spring Jammed). I I I I I I I I I
~
l VII.A.2OO I Arkansas One 2 8 19% (power i SI I Check valve body oversighted. I L i e I i 1 escalation ! I I I I I I testing) 1 I I I I I I i 1 1 1 d i VII.A.240 1 Three Mile i Shutdown i HPI I 5 eat hold-down found loose. IL I I I Island 1 I I I I I I I (02/80) i I I I I I I i i 1 1 I I VII.A.256 I Haddam Neck i Refueling i HPSI I Containment penetration check valves IL 1 I I (05/80) 1 I I leaked (> technical specifications) due i I #
I I I I I to uneven seating surface component i I I I I I I cooling water check valve. I I I I I i 1 8 i C I VII.A.260 i Oconee 1 1 73% Power i HPI I I I I I (06/80) i I I I I
' I i VII.A.262 I I i Crystal River 31 Shutdown I I I I b i SI l Core flood check valve failed (cause !L 1 I l (07/80) 8 I I core flood /n2 check valve isolation i I
' I I I I I i 1 unknown). I f I b I I I I I I VII.A.273 1 Davis Besse i Hot i SI I Cross back leakage through cone flood iL 1 I I (10/80) i Shutdown i I check valve n2OO (unknown)- valve i I -
I I I I I improper 1g assembled. I I I I I I I I I b I VII.A.275 i Deaver Valley I Shutdown i SI I l-H SI pump check valve antirotation i F 3 -
1 I (10/80) 1 I I device binding due to flow or pressure i I I I I I I surges (check valve did not reseat). I I
'
- I I i 1 I I I -
1 I I i 1 I I I VII.A.270 i Sequoyah I Hot i 1 1 I b I I (09/80) i Shutdown 1 81 1 Check valve stuck open disc weld / IL I I I I I I valve body interference. 1 I I I I I I I I b I VII.A.230 i Maine Yankee 1 97% Power i SI I O-ring failure on the bearing cap of IL I l l (10/79) ! I l the swing check valve. I I b
I I I i 1 8 I I VII.A.285 i Salem 1 i Startup 1 SI I Check valve in the interference between i FC I b I I (12/00) i I I the RCS Hot Log II and Safety Injection i I I I I I I pumps failed to close during test. I I b I I i i i I I Page 4 0
I I I I I I I 3 VII.A.291 1 Surry 2 1 100% Pawsr ! SI I SI accumulator check valve leaked IL l
! I (01/01) 8 I I through causing boron dillution in i I I I I I 8 accumulator (normalig closed manual I i 1 I I I I valves were left open intentionally i I I i l I I making the check valves value the 3 I I I I I I interface). I I 1 I I I I I I -
I VII.A.294 8 Oconee 1 1 0% Power i LPI I Check valves leaked escessively - disc 1L I I I (02/80) i I I cocked by deposit buildup (leaking i I I I I I I valve was the final value in LPI loop B 3 8 g
I I I I I before reaching reactor vessel). I I I I I I I I I 3 I I I I I I g
I VII.A.306 l McCuire i Power i SI I Accumulator discharge check valves 1L I I I (04/61) i i 1 peaked (two values on cold leg I I I l l l t injection accumulator al. I 1 '
I I I I i l I I VII.A.307 1 McCuire i Power i SI I Discharged check valves associated with I L I 3 I (04/01) i I I accumulators d and c leaked. 1 I '
l 8 1 I l i I I i VII.A.311 i McGuire i Power i SI I Accumulator check valve failed IF I I I (05/81) 1 Cold I I (both RHR trains inoperable in cold i 1 '
l 1 1 I Shutdown i I shut down). I I I I I I I I i
~
l VII.A.315 i Point Beach i Startup i SI I Si check valves I-853 c and d (unknown) IL I >
I I (07/81) I I I (first check valves between the RCS I i
! I I I I and the low head SI core) leaked i I 1 I I I I escessively. 1 I '
I I I I I I I I VII.A.316 i Three Nile I Cold 1 MU 1 Addition to VII.a.240 re. 1 - I I I Island 1 1 Shutdown 1 I tilting disc check valves. 1 I I' I I (06/81) 1 I I I I I I I I I I I i VII.A.339 8 Cook 2 i Startup i SI I Check valve disc missing - Construction IL I O I I (10/81) i I I error (discovered during valve test). 1 I I I I I I I I
~
l i i i I I . I b i VII.A.343 i Point Beach l Refueling i SI l Check valve leaked during test (first I l' I I I (10/81) I I I check valve from RCS for (minor) tow I I s I I I I I head SI cone deluge line). I I s I I I I I I I I VII.A.363 i Palisades i Refueling I LPSI I Excessive wear in swing check valve. IL I
~
l I (09/81) I I I I l -
1 I I I l 8 I I VII.A.372 i Davis Besse i Refueling i HPI l HP Injection stop check valve stuck iF 1 1 I (06/82) 1 I I (up to 580 psi required to establish ! I -
1 I I I I flow) cause--incorrect seat angle. 1 I I I I I I I I b I VII.A.375 1 Davis Besse i Refueling i HPI I Addition to VII.A.372. 8 - 1 I I I I I I I I VII.A.384 i Calvert i Startup 1 SI I SI tank outlet check valve leaked 1L I b I I Cliffs 1 I i 1 ring deteriorated (200 gpm). 1 I -
1 I (07/02) 1 I I I .I 1 8 I I I I I I
' ' I VII.A.386 I North Anna 1 i Refueling i SI b I Cold leg injection check valve leaked I L I I I (09/82) 1 I I - carbon steel closure studs corroded. 8 I I I I I I I I b I I I I I I i i Page 5 0
l VII.A.393 I Bsavsr Vollsv 11 Special i HPI I Additig.n to VII.A.275. I- I 1 I (10/22) I Testing i i i l i i I I I I I I I I I I I I 1 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) 1 I I the open position - two different i I I I I I I causes. I I
~
l i I I I I I d i VII.A.401 1 Maine Yankee i Startup 1 SI I Safety injection check valve seal weld IL 8 I I (02/83) ! I I leaked (check valve in the loop 1 RCS I 1 1 I I I I safety injection line closest to the 1 I O I I I I I cold leg). 1 I I I I I I VII.A.403 i Surry 2 8 100% Power i SI I Accumulator check valve leaked. IL i O I I (09/82) I i 1 1 1 I I I i 1 1 1 I VII.A.407 i McCuire 1 1 St rtup I SI I Two cold leg injection accumulator IL I 1 1 (05/03) i 358.50% 1 I check valves leaked (unit was shutdown 1 I I I I Power i I to repair valves). 1 I .
! I #
I I I I I I VII.A.420 1 Davis Besse 1 Startup i SI I HPI stop check valve stuck-disc to IF 1 8 I (09/83) i I I seal contact area too wide-design 1 I ,
I I I I I deficiency. Value opened on 3 to 4 psid 1 I I I I I I instead of 100 paid. I I I I I I I I I I VII.A.422 i Calvert i Startup 1 SI I Addition to VII.A.384. I - 1 1 1 Cliffs 1 I I I I I I I (07/82) i I I I I
." l l I I l i I J l VII.A.437 I Farleg 2 i Refueling i SI I RCS pressure isolation leaked-- IL i i I (09/83) 1 8 I incomplete contact between disc and i 1 I I I I I seat. 1 I 8#
1 I I I I I I I VII.A.452 i Saint Lucie 2 i Hot I SI I Safety injection tank check valve I L 8
~
I I (12/84) I Standbg i i leaked. Seal plate cockeda valve seat i I O I I I I I compensating joint ball gated. I I I I I I I I I I VII.A.456 I Calvert Cliffs 1 100% Power i SI I Estessive SIT check valve leaked due to 1L, 8 O
I i 1 and 2 I I I 0-ring material degradation. I I I I (01/05) l I I (Unit 1 = 1.6 gpm - Unit 2 =27.2 gpm.) 1 8 s 1 I I I I I I J l VI.E.462 i Surry 2 8 100% Power i AFW I Steam cut setss general check valves. IL I I I I I I I I
~
! VI.E.412 1 Arkansas One 2 i Refueling i EFW I EFW pump turbine steam suppig check i FD 1 -
1 I (10/02) i I I valve falted--internal damage. I I I I i i I I I b I VI.E.415 1 San Onofre 2 1 Preop i NFW l Low N2 actuator pressure on steam IL I J l I (09/82) 1 Testing i 1 generator feedwater isolation valve. I I I i 1 1 1 N2 leakage on plug. 1 I
J I I I I I I I l VI.E.353 i Robinson 2 1 100% Power i AFW I Valve fails to open. Ausiliary feedwateri L I I I (06/81) l 8 1 discharge to steam generator A. Caused 1 i b I J I I I I by heat due to back--leakage of down- ! I l i I I I stream check valve. I I I i 1 I I I I b 1 VI.E.199 1 Jeaver Valleg 18 Startup 8 AFW I Steam suppig trip value would not close.1 FC 8 b I I (03/79) i 100% Power I i l i I I I I I I I b i VI.E.32 1 Ginna l Power I NFW I "B" steam generator feedwater i L i O Page & O
t I (07/73) I I I central volvs plug esparatsd i I
! I I I I from stem. caused loss of feedwater I I I I I I i and water hammer 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 l 1 (04/82) I and Power I i check valves (CS and RS). I I t i I I I I I 1 VII.B.123 i Rancho Seco 1 Starting i SI I Various problems on check valves caused IL I -
1 I (04/82) 1 I I backflow from BWST to horated water i I I I I I I additive tanks. 1 I I I I I I I I O I VII.B.131 1 Cinna i 100% Power i CS I Core spray pump check valve failed I FC 1 1 I (09/82) I I I to close seat and disc machined. I I 1 I I I I I i O I VII.B.134 i Oinna 1 100% Power 1 CS I Spray pump check valve pin / swing are i FC i 1 1 (06/02) 1 i 1 worn. I I I I i i I I I J l VII.B.142 i Cinna i 100% Power I CS I Core sprog pump discharge check valve I I I I (07/82) i I I faulty. Pin / swing arm worn. Air- 8 1 1 I I I I operated valve. I I J B 1 I l i I I I VII.D.34 I Cook 1 1 100% Power i CI I Containment isolation valves VCR-10 i FC 1
- J 1 I (9-12/79) l I I and 20 would not close. I I l l I I I I I I I l l 1 December 1979--valve VCR-10 failed to I FC I
" J 1 I I I t c l o s e' l I l i I I I I I I I Cook 2 8 75% Power 1 CI I Centainment Isolation check valve 1 WP t ,
1 I (06/80) 1 I I VCR-156 installed backward i I '
I I I I I I I I I Cook 1 1 1 I November 1980. VCR-20 fa!!ed to close i FC I
. 1 (11/80) I I I I I U l i I I I I I I VII.D.47 I Yankee Rowe i 100% Power ! CI I Containment steam heat suppig CI check I L I 1 I (04/82) l I I valve flanges leaked. Renasation of I I g
I I I I I flange bolts. I I
,l 1 1 I I I I I i
- I VII.E.78 i Arkansas One ! I 100% Power I I Check valve failed to open - corrosion 8 FO I O I I (03/80) 1 I I I I I I I I I I I -
1 VII.E.28 I Davis Besse 1 1 Refueling I CI I Sin CI check valves leaked. Discs and IL 1 1 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 60% Power i I Hydrogen dilution blower check valves 1 FO I
, 8 I (04/78) i I I stuck (suction check valves) 1 I I I I I i 1 1 I VII.B.28 i Rancho Seco I Hot i DHR I Two stop check valves stuck closed - 1 F 1 1 I (1974) 1 Functional i 1 Design error I I I I I Testing i 8 3 I L I I I I I I I I V11.D.43 i Robinson 2 1 100% Power i CCW l Debris check valve caused CCWP trip, IL I I I (01/75) I i 8 cracked pump seal. I I i b I I I I i 1 I I VIII.B.82 1 Point Beach 2 I Refueling I CI I CI Check Valve in CCW Suppig to "A" !L i I I (03/76) 1 I I RCP Leeked. Dirt Accumulation on Seat. 1 I b I I I l I 8 I I VIII,B.118 1 Palisades i 100% Power 1 SW l SW Pump Discharge CKV stuck. I FC 1 I I (02/77) l I I Lubrication. I b I I I I 1 1 1
I g
Page 7 1
I VIII.B.12S I Cost 1 I I:sfueling i SW I SW CKV Locked. Depenits frta Lche IL I 1 I (02/77) 1 I I Water. I I I I I I I l i I VIII.B.127 i Calvert C11Ffs i 100% Power i SW I SW Pump Discharge Check Valve Stuck i FO I 1 I (04/77) 8 I I I I I I I I I I I I VIII.B.151 1 Calvert Cliffs I Power i SW I SW Pump Discharge CKV Stuck Corroded i FO I 1 1 1 . I Flapper Hinge Pin. I I -
1 I i l l 1 1 I VIII.B.152 i Davis Besse 1 i Power i SW I SW System CKV Rusted and Inoperable i FC I g 1 I I I I I I I I I I I I I I VIII.B.183 i Palisades I 90% Power i SW I SW Pump Discharge CKV Stuck Shut. I FO I g I I (04/77) i I I Corroded Hinge. 1 I I I I i 1 1 1 I VIII.B.228 Cook 2 1 100% Power i SW I SW Suppig CKV to Diesel Failed to Seat. I FC I 1 I (01/79) 1 1 i Dish Wear. I I -
1 I I I I I I I VIII.B.233 3 Calvert I 94% Power i SW 8 Saltwater Pump Discharge CKV Failed to I FO I 1 I Cliffs 1 8 I I Open. Foreign Debris. I I I I (02/79) i I I I I I i i 8 i i I I VIII.B.290 3 Cook 2 1 Refueling i SW I NESW Check Valves Leaked. Sand Deposits IL 3 1 1 (12/79) 1 I I Pitted Seats ! I I I i l l I I
~
. l VIII.B.308 i Connecticut i Refueling I CI I Containment Penetration Check Valve IL l I I I Yankee i I I Seats Leaked. Weld Problems. 1 I I I (05/80) I I I I I I I I I I I I I I VIII.B.310 1 Beaver Valleg 11 Starting I SW 8 River Water Check Valves Deteriorated i FD I
, I I (04/80) 1 I I I I l ~ l I I i 1 8 8 4
I VIII.B.411 1 Cook l Refueling I SW l Leaking NESW Check Valves. To Be IL I I I (07/81) i I I Replaced With Diaphragm Type. t l I I I I I I I O I VIII.B.423 i Surry 2 1 100% Power i SW l Charging Pump SW Check Valve Failed to 1 FC I I 8 (08/81) i I I Seat. I I 1 I I I i 1 8 O 3 VIII.B.461 1 Surry 2 1 100% Power I SW I Charging Pump SW Check Valve Worn. I FC 8 I I (08/81) i I I Stuck Open. I I -
l I I I I I I
- I VIII.B.493 i Davis Besse 1 1 Cold i SW 8 SW Swing Check Valve Disc Seized Fully i FC I I I (06/82)) 1 Shutdown i I Open. Corrosion Deposits. Disc Stud i I I I I I I Penetrated. I I ~
l i i 1 1 I I I VIII.B.569 i Cook 1 I Refueling I CS I Containment Spray and RHR Check Valve iL 1
" I (08/83) I Leaked Disc Lapped.
8 l I I 1 -
1 I I I I pump SW discharge) I I l I I I I I I b I VIII.B.574 5 Surry 1 and 2 1 100% Power i SW I Check Valve seeding fatture. (Charging IL 1 -
1 I (6-10/03 I and l l pump SW dischare) I 1 I I I Shutdown l I l l I I I I I I I -
l VIII.B.609 i Crystal River 31 94% Power i SW l Seawater Pump Discharge Check Valve i FC .I lb 8 I
8 (04/84)
I i
i I
1 I Failed. Corrosion 8
i 3
I I b I IX.F.101 1 Rancho Seco 1 95% Power i SI I Make-up Tank Pressure Above N2 tiender 1- 1 I I (10/02) i I I Pressure Cause Value/ Leakage. (Not a l I b I I I 8 i Value Problem) I I Page 8 0
I I I I I I I 1 50% Power I IX.F.120 l McCutre 1- 1 VUCDT (Iow totalizer erratic choc valve iL I l I (8 and 9/82) I I I leaked. loss of 11guld at the flow I I I i ! i I element. I I 1 I I I I I I I IX.F.44 i Maine Yankee i 90% Power icontain-l Check Valve on Containment APD pump i L 3 8 I (04/75) i Iment APDI failed. Worn seats. I I 1 I I I i I 3 *'
8 IX.F.689 8 Cinna i 100% Power IContain-1 Containment Gas R/A monitor Check Valve l L 1 8 I (6 and 9/82) I iment Gast Leaked. FM on the Seat. I 1 I i 1R/A 8 a 8
1 1 g
I I I IMonitor I I I I I i i 1 i !
1 IX.F.78 8 Cook 2 i Refueling I SW l 13 NSt! Check Valves Leaked. Lakewater i L I O I I (02/77) 1 I I Deposits on Seating Surfaces. I I I I I I I 23 containment Purge Isolation Valves i I I I i 1 i Leaked. Loose Actuators. I I #
- LEGEND:
FC = Fall to Close -
F0 = Fail to Open FD = Failed on Demand F = Failed L = Leakage WP = Wrong Position
.)
Sb O
, Q
.)
)
.)
5, 1
- e Page 9 O
ATTACHMENT C SAN ONOFRE EVENT OF NOVEMBER 1985 (Reference 2)
E 624. Emmetor Tr*p De less Of Tltal los - lose Of W Systes latearity - Water Rammer - Main FW Queck Talwes Failed s E in Gene 1 - Jan 86 San'onofre 1 - v 85 - 601 power on 6 Jan, the NRC issued IE Information Notice 86- The reactor trip also caused level shrink in the 01 " Failure of Main Feedwater check Valves Causes SCs, causing SC level to drop below the actuation loss Of Feedwater System Integrity And Water-Hammer level for tha aux W ( AW) pumps. The electric-Damage , to inform recipients of a recent eveng ,
ally driven AW pump received an actuation signal, caused by 5 main FW (M W) check valve failures at a but no longer had electric power available. The PWR plant. De failures resulted in a loss of Mg stese-driven AFW pump, after a 3 min auto warsup system integrity and significant water hammer dam. period, began to deliver relatively cold FW to the age. SCs at a point in each of the MW lines between the On 21 Now, San Onofre 1 was operating at 60% power MFW regulating valve discharge-check valve and the SC. The AW to the SC FW 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 FW pipe in east (electric) MW pump. ne west (electric) Mg primary containment. Although the operators were pump remained energised from the unit main unaware that the check valves had failed, they then generator because of an abnormal electrical lineup. closed all MFW regulating velves and their When the east NFW pump tripped, its discharge-check associated isolation valves per procedures. (The valve (FWS-438) failed to seat properly. As e effect of the MW valve closure had yet to be result of the discharge-check valve failure, the west MW pump supplied FW backward through the determined.) Contact between steam in the FW lines and the cool AFW in the horia pipe resulted in a discharge-check valve and overpressurized the east water hammer. The water hammer caused damage to FW heater-condensate train. Several t'ub e s appar.
the FW line pipe supports and stretched the bonnet ently ruptured in the east FW train fifth stage bolts os the 'B' W regulating valve bypass line W heater as a result of the over.
(1.P) check valve (FWS-378), causing the metal valve pressuritation, causing the shell side of the gasket to extrude. The flapper on the 4 in. check heater to rupture. Also, several main turbine valve was later found to have been damaged by the rupture discs failed. Following these events, the water hammer impact. We extrusion of the valve operators tripped the reactor and turbine per gasket resulted in a substantial steam-water leak procedure because of the loss of power to a vital from SC 'B' to the FW mezzanine area and the ata bus. That also causel the west MFW pump to trip.
Both 12 in. MFW pump discharge-check valves ( WS- which was not isolable for some time because of the open, proximity of the associated isolation valve to the 438 and -439) were later found cocked leak. SC 'B' boiled dry because all AN 'B' flow supported by their disc antirotation lugs that had was carried out through the leak. Plant personnel rotated under the check valve hinge arm. were finally able to close valves FWS-342 and FWS-376 to isolate the leak and continue the plant When the west MW pump tripped, att 3 SC MFW regu. cooldova - 6 he af ter the event started.
lating valve discharge-check valves (FWS-345 -346 and -398) also failed to seat. hio of those 10 in. The NRC sent a 5-member incident investigation team check valves were later found to have their flap. (IIT) to the 3an Onofre 1 site shortly after the I 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 any work in progress r.r planned (as allowed by later found to have failed in the same mode as WS. plant safety considerations) until the licensee and 438 and -439. The check valve failures in the HFW the NRC had an opportunity to evaluate the event.
system resulted in leak paths from the SCa bsckward The licensee also agreed to keep Unit I shut down through the MW regulating valves and the east MFW until concurrence was received from the NRC to r(-
pump to the ruptured east train W heater. Also, turn to power. The IIT completed a preliminary in-the west MW train might have been pressurized from vestigation of the event and expected to issue a the 3Cs. The net effect was that the inventory in report in Jan 86. (wvg) all 3 SCs began to blow steam and hut water back through the east MFW train.
i 4
RAI 75 ' *,
In the description of RHR pressure boundary failure mode 9 it is stated '
that the maximum value of stresses due to pressurization to 2250 psia in ,
the limiting RHR piping are approaching the yield stress and the stresses
> in other metallic components are at a small fraction of,their respective yield stresses. Describe the analyses conducted to support this '
conclusion and provide a summary of the pertinent results. In addition, i clarif y whether the pressure loading has been applied as a dynamic pulse coupled with corrosion degradation ef fects (such as heat exchanger tube embrittlement). If these ef fects have been considered, describe the analyses and the dynamic loads. If not, provide the bases for not con-sidering these effects. -
RESPONSE 75
, i The analysis that support the conclusion that RHR system piping is the weak link with regard to gross rupture f ailure modes of the RHR system consist of a systematic qualitative evalution of each RHR system [
component, including the piping, heat exchangers, valves, instrumentation, j the consideration of the design pressures for each component, and stress calculations performed for two limiting components: the RHR system low pressure piping and the RHR heat exchangers. The results of these calcu-lations, as stated in RMEPS, clearly show that the piping is limiting in the initial plant configuration for gross failure and the RHR pump seals are limiting for leakages. These initial calculations are based on dynamic loads. Subsequently, in response to an NRC review query, calculations were performed to show that the above conclusion holds even when it is assumed that more than half of the RHR heat exchanger tube thickness is lost due to errosion/ corrosion effects. Wh.n the RHR heat exchanger tubes have their thickness reduced by 1/2, the stresses at 2250 psia are still less than yield stressess of that material. There is no evidence that metals fail at yield stresses. The stress calculations and other information supporting the pressure boundary analysis are sunmarized below. Thi s information includes separate discussions on:
o RHR pressure boundary failure modes o Additional V-sequences with failed open check valves o RHR piping failure probability a o RHR heat exchanger degradation c
/ o Frequency of unfilled RHR piping [
t f-
'* ---r ew , - - - - - - ,_ , _ . , , _ _ _
RHR PRESSURE BOUNDARY FAILURE MODES A wide spectrum of f ailure modes of the RRR 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
,4
- intermediate leak ( .09 - 1.05 in 2)
- 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 hi6hly 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 f ull range of alternative assumptions chosen to bracket 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 include ethylene-propylene elastomer o-ring material. While the pump 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 piping and heat exchangers remained intact but in several of these instances, RRR 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 pump seals). This range bounds the magnitude of possible leak sizes caused by failure of the primary seal. A larger leak would require f ailure of the mechanical seal assembly. As noted on page 3-31 of RMEPS, such a f ailure 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 ail 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 f or both pumps.
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 f requencies associated with this uncertainty.
Hence, it really doesn't matter whether the seal completely blows out or remains fully intact.
l l TABLE 3-14 RMEPS CASE A CASE B EVENT TREE - EVENT TREE SI = 0.0 L1 = 1.0, L2 = 1.0 END STATE L3 = 0.0 VI l VS VI l VS VI l VS LOCA 4(-6) 3(-6) 5(-6) 3(-6) 4(-6) 3(-6)
DLOC 4(-7) 0 0 0 4(-7) 0 DILOC 3(-4, 3(-7) 0 0 3(-9) 3(-7) 8C 7(-10) 0 0 0 5(-10) 0 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)
IFV 3(-9) 2(-9) 3(-9) 2(-9) 3(-9) 2(-9)
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 assembly and studs, and RHR heat exchanger would not f ail 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-tions of their respective yield stresses.
Our assessment that the RHR heat exchangers is probably the stronges t link in the pressure boundary is supported by calculations that show that even if the tubes are assumed to 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 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 syst ems). On the other hand, the achievement of yield stresses does not in itself produce failure. Finally, even if the heat ex-changer tubes leak, the impact on source terms and consequences is not believed to be significant.
t
Additional V-Sequences Involving Failed Open Check Valves The following are responses to certain questions that arose out of recent discussions about the V-sequence in the above reference. We have chosen to respond by:
- 1. Reviewing the relevant scenario phenomena.
- 2. Considering single valves that separate the RHR system f rom othe r systems and f rom potential release paths.
- 3. Describing new scenarios that arose out of I and 2.
- 4. Providing order-of-magnitude estimates of the frequency of the new scenarios.
- 1. REVIEW OF SCENARIO PHENOMENA A V-sequence is initiated by leakage of reactor coolant through valves that separate the reactor coolant system f rom the RHR system. Leakage is of concern only if it is large enough to cause loss of inventory to exceed the makeup capacity of the CVCS (about 150 gpm). Leakage of suf ficient size would initially reveal itself by causing RHR relief valves to open.
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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 psig, these valves are capable of relieving a f ew 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 lines 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 gpm 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 reset 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. RRR system pressure is determined by the relief valves.
As the rupture size increase, 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 iniating valve rupture size would determine the pressure.
The worst scenario would be a sudden, catastrophic failure of a paire of valves separating the RHR system f rom 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 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 f raction of the pressurizer volume per minute. Therefore, an "s" signal on low pressurizer 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
. . _ . _ = __ _ _ _
are not likely to be available because of pressure and temperature conditions well beyond their desin basis.
MAAP calculations were performed f or 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 i 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.
The above discussion was intended to point out a f ew important features of the V-seque nce . 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 cpmponents decay awat 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 , i t 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. Although 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 connent to the RHR system and are separated from it by valves.
4 Table 1 delineates the type and normal position of valves that separate high design pressure systems f rom the RHR system.
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I 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 MOV (train B) RH-V38 Normally Closed Interlocked MOV 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 chect valve and failure to close a normally open motor-operated valve.
- 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 scenarics 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
4 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.
1 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 u. cur. 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 piping and heat exchangers remain intact.
The ESD cuestions 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 MOV 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 j be available to make up loss of inventory for valve ruptures equivalent to an j 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.
l The IFPV plant damage state was assigned.
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 scenar:o 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 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 s'uction 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 conseryccively higher than the mean values that a detailed uncertainty analysis would reveal. The point estimate analysis indicates that 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 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.
P(EPMP) = 10-4/ demand.
P(SHS) = 10-4/ demand.
i
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 response procedures to initiate ECCS or cooling to the steam generator. Since 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 quantification 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 far less likely. In other words, although the ESO 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 fail open on demand is P(CV0) = 5.5 x 10-4 We judged that failure of either MOV 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(0PMOV) = 0.1 Our conservative point estimate of all the additional scenarios that lead to plant damage state 1FPV may be obtained by 4 (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.
l From Reference 1, f(VS) = 3.3 x 10-6/ reactor year.
f(VI) = 4.5 x 10-6 /reactor year.
Numerical substitution yields f(added 1FPV scenarios) = 3 x 10-9/ reactor year.
uency of
.plant Thus,damage our conservative state 1FPVpoint of only estimate about 10% yields ofan theincrease frequency in (3 thex freg/ 10-o reactor 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 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.
J 1
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I RHR PIPING FAILURE PROBABILITY The calculation of Ehe RHR piping f ailure 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 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 would 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 f ailure probability, and f urther conserv-atively assigned a 17. f ailure probability to the yield strength.
There is no evidence that any failures of the ASME test specimens ever occurred at the yield s trength.
To account for undetected design errors, material defects, and in-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 fragility curve around 2250 psia is strongly believed to be conservative.
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 system. 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. Bo rated reactor coolant that meets reactor coolant chemistry specitications 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 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 RilR heat exchanger degradation is likely to occur beyond that con-sidered in the design basis.
Under the maximum internal pressur zation 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. Aay inelastic defo rmation would relieve the stress.
The following equation may be used to estimate the amount of tube thinning that must occur to yield the tubes because of internal pressure.
2b2 ,l Shoop = P b'-a' 1
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P = internal pressure.
b = outer tube radius.
a = inner tube radius.
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thinning must then be at least 58% to approach yield.
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s FREQUENCY OF UNFILLED RHR PIPING The enclosed writeup provides a conservative bounding analysis of the frequency of RHR piping'f ailure due to an undetected presence of air in the syst an'. 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.
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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 oa 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 twc 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. lhe 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.
A review of about 280 reactor years of PWR 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
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 inje: tion 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 not standby. These data cover about 276 reactor years of experience on a calendar basis. Assuming our average value of refueling intervals of 1.6 years, this covers about 184 refueling outages.
Therefore, the average frequency of RHR air ingress problems has been about (10/184) = .054 events per refueling outage. Over the period of this same data base, the amount of reactor power operation, assuming a 60% average capacity fact reactor years or166x8,766=1.5x10gr,hasbeen(0.6)x(276)=166 hours. The average frequency of RHR y problems, while at power in hot standoy, is then 'l.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 powe~ 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 th:t the initiating event occurs more than 2 months after the previous refueling outage, it is necessary to postulate failure of multiple monthly a 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 si:rveillance test is 1 x 10-2 and again conservatively assumed that if 15 is not picked up in the first test, it is not picked up in any
% bsequent tests, depending on when tne 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 .0S4 air ingress events per refueling outage x .001 failure to detect air in preoperational testing x ((1/12) + (11/12 tests] = S x 10-6,) x .01 failures to detect air in surveillance To calculate the frequency of an initially unfilled itHR 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
during power operation was (2 x 24)/(1.5 x 10-6) = 3.2 x 10-6 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.
Ar. 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.
Component Maintenance Frequency
- Number per Component (events /nour) 2 Motor-Driven Pumps 8.4 x 10-5 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 the.SSPSA, is considered highly conservative because 6.5 x 10-b 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 perfomed more than 1 month prior to the initiating event) the likelihood of an undetected air bound RHR system due to maintenance events otner 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 obtaine(' 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 17. 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
not important is fully justified. Stated another way, if we conservatively add the above bounding estimate of RHR failure frequency 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 Effect Following ... RHR System, PLG memorandum to K. N. Fleming, August 22, 1986.
3 Nuclear Power Experience, Vol . PWR-2, S. M. Stoller, Albuquerque, New Mexico, 1979-1985.
l
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) f Sheet 1 of 3 i Event Plant Status Date NPE Reference
- 1. Safety Injection Pump Cavitation. Turkey Point 3 Cold Shutdown May 1974 PWR VII.A.46 l l Air ingress believed to be caused (ECCS) j 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 PW XV.421 cavity, water level in the RPV PWR XVI.B.70 !
dropped to level of the hot !
legs--led to cavitation of RHR j (RHR system was not pumps.
solid.)
- b. RHR Pump Cavitation. Voids McGuire 1 Refueling April 21, 1983 ?WR XV.421 existed in the piping of the nonoperating RHR train due l to physical layout of the RHR i system and vessel water level . >
(Top of RHR heat exchanger ;
tubes was ~ Elevation 758' !
and RPV level was t
~ 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. '
l f
TABLE 1 (continued)
Sheet 2 of 3 Event Plant Status Date NPE Reference
- 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.
- 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.
- 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 Valley 1 Cold Shutdown April 1980 PWR VIII.B.296 PWR VIII.B.304
q l
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TABLE 1 (continued)
Sheet 3 of 3 Event Plant Status Date NPE Referer.ce !
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.
ATTACHMENT F
_ _ _ ~ -- -_
Val. PWR-2 VII. 52faty Systems
- 44. BORIC ACID 11tANSTER PUMP SHAFT FAILURES A. Emerg. C?ra Cool.
. p. 13 Robinson 2 - Aug & Dec 73, Mar 74 Two previous incidents of boric acid transfer pump shaf t failures (in vicinity of shaf t 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).
D e "A" Pump was in service recirculating the contents of "A" BAT.
The "B" pump breaker then tripped due to thermal overload, h e breaker was reset and the pump re- 46. AIR IN SI PLHP started. Pump discharge pressure was checked and found abnormally low. The BIT closed with no Turkey Pt. 3 - May 74 appreciable increase in pump discharge pressure.
The pump was secured and declared inoperable. The 3A SI system pump was started in order to add water to the SI accumulators. ne pump started but n e pump motor was checked electrically and found to indicated abnormally low running current and low be sound, therefore, it was apparent the impeller discharge pressure. The pump was imediately had separated from shaft or the shaft had broken. stopped and valve lineup was verified to te correct.
Investigation proved the latter to be true. The With maintenance personnel present, the pump was pump was replaced with a spare from stock. De 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. De "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 h e "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 I' nit 3 6
- 45. BORIC ACID DANSFER PUMP SHAPT FAILURE 4 SI pumps were tested and performed satisfactorily.
Robinson 2 - Apr 74 he SI and RHR system piping and pumps were also vented to ensure they were free of trapped air.
They 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 imediately checked the pump and found it systems.
stopped. The pump breaker was normal, therefore, it was determined the pump had tripped due to ther. The exact cause of the air in-leakage could not be mal overload. ne 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 n e 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 r
vicinity of the keyway was determined to be the draining of a line. About 1 mo earlier a valve cause of the failure, n e pump 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 l 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 cocrnon 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. PIPINC VIBRATIONS DURINC FILLINC 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 investigatien. See VII. A. 58 & 67 f r subsequent (aqq)
During a shutd vn, and while filling a SI tank shaft failures. using a high pressure SI pump, the operator noticed a pressure drop on a pressure indicator. An increase was also noted in the Contaircent 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 m - e.- e Aug 79 I
t
\ _ - - - . _ - -
Vol. PWR-2 XVI. Operatione* Problems
- 5. Refueling
- p. 28
- 69. H.P. Technician Seriously injured, been closed during an attempt to reJuce leakage O
Contaminated In 40 Ft Fall From Crane into the RCDT, isolating the RPV level gauge used l To Reactor Cavity Floor to indi; ate that the level was approaching the ves- I sel fla,ge. Procedures Jid not require visual mon- l Cinna - Mar 83 - 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 XV.421 for additional conse-the refueling bridge of the manipulator crane in quences of this event. (sab) preparation for refueling operation, above the d ry reactor cavity. The manipulator crane was located at the south end of the reactor cavity against its 71. Addition To XVI.B.66 re: FAs Damaged south travel stops above the fuel trans fer mechan- During Core Loading - Remedial 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 rf the fire watches (between empty core center and was resting on FA W-51.
065b 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 Now 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 transportation 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.
j 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 ares /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 off-170 mree 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 be fore resuming refueling operations.
dpm/100 cm2 ). Air samples 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 25% of MPC for CoDO, the only isotope identifiable the in-core, mast-mounted TV camera and the manipu-on the air sample. Nose smeirs indicated minor lator crane were performed as required. Operator contamination so no significant internal contamina- requalification classes were to review fuel handl-tion was s us pe c t ed . Surface contamination levels ing procedures, provide classroom training on fual were in the range of 200 cpm on the neck to 1 ,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. (sje)
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 N Millstone 2 - Oct 83 - refueltng
Stopped - Level Not Monitored - Operator On 29 Oct, the spent fuel pool gate, a load of Valving Error about 4500 lb, traveled over irradiated FAs in the spe fuel pool. TN gate was being moved from the McGuire 1 - Apr 83 - refueling transfer canal to the cask laydown pit when discovery was made. The load was then returned on 5 Apt, while the Unit i refueling cavity was over irradiated fuel to a safe position by the being drained so that the RPV head could be placed transfer cansl. 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 stoppeJ. Since the positioned over the designated safe load path. To ND system was inoperable, RCS loop avg temp increased 260 prevent recurrence, special rigging procedures were The chem and vol control system was incorporated into the crane travel procedure to available to provide cool makeup water. The reac- illow the gate to travel over designated safe load tar coolant drain tank (RCOT) isolation v.a lve had paths. (su) cga Mar d4
I Vol . PWR-1 XV. Misc Systems
- p. 113 f SNUBBEit DEFICIENCIF.S 420. ADDITION To XV.192 re I Two dit[erent analyses were performed on the fluid ine by the Flcrida Pow.r Corp. (FPC)
(ontaminants: Millstone 2 - Mar 83 Materials technology Dept. and the other by the snub-ber afr. The Materials Technology report listed par- As of 22 Mar 83, all International Nuclear Safeguards ticlea of up to about 770 microns in size. Their Co. (INC) mechanical snubbers installed on hangers ennclusion was that the particles of Fe, Ni, and Cr that supported safety related piping systems had been cauld oc from SS. The others appeared to be of extrananus origin (based upon review of materials replaced with mechanical snubbers manufactured by (ofx,rve) making up various snubber parts). The report also Pacific Scientific Co.
concluded that some of the elements could be of paint pigment origin. 421. RHR PANCElt FAILURES - SNUBBERS 1.0CKED UP -
WATER HAletER DAMAGE in t9 Power Piping Co. analysis (done by Atmospheric P %earch Organization, Inc.), it was determined that McGuire 1 - Mar, Apr & May 83 - shutdown pirticulite chemistries indicated welding metal and llus, paint chips, and " dirt" as contaminants. The On 27 Mar, during unrelated maintenance, hanger INCA-report also showed an nievated phosphorus level found ND-H260 was discovered to have pulled loose from its in an unknuwn particle type. It was stated that this All RHR (ND) system mechanical snub-a wmastry was "not incompatible with welding rod wall anchors.
bers were subsequently full stroked on 27 Mar to en-elax." the one major dif ference in the 2 reports was sure sure their operability. Two snubbers (IMCA-ND-t u it the Materials Technology analysis was perform,d H177 & -H317) were found to be locked up. A walkdown nn fluit samples taken f rom snubbers removed trne the inspection of susptet piping wan performed. On 7 se era, building. The Power Piping analysis was Apr, hangers ND-H273 & -H287 were discovered pulled t- r to r.ned un fluid from a snubber shipped to the loose with broken concrete around their well anchors.
tield (Detroit Edison Co.) and returned to Power Continuation of the system walkdown on 8 Apr, revealed Piping Co. after about 3 mo without having been 6 more hanger failures (ND-H7, -H21, -H274, -H282, installed.
-H306 & -H308). The exact time that damage occurred FPC concl.aded that environmental conditions might could not be determined or attributed to a particular occurrence of water hamer (overloading the have contributed to the contamination problem. How-it was their opinion that there was enough hangers / snubbers). These hangers had been inspected "ver, J. eta to say that sufficient quantities of and approved in Jan 83. At least 2 ND water hammers avillable were thought to have occurred since that time.
O coqtamnints were present in the snubbers as supplied by Power Paping Co. to render them inoperable. An On 22 Jan, during shutdown, the ND system was aligned inoperable anubber a this failure made (aero bleed- to provide normal decay heat cooling to the reactor rato) could induca v.cessive amounts of stress in core. It was theorized that voids formed by flashing piping that was supported by the snubber. Therefore, EPC determined this issue to be reportable per 10 CFR of an isolated vol of water between valves N D-! &
ND-2 during leak testing might have contributed to a 21 because the snubbers contained defects which con- water hammer at that time.
stituted .a substantial safety hazard.
The problem of fluid contamination had not been noted On 5 Apr. during draining of the refueling cavity, until t ief ir then most recent refueling outage. No water level in the RPV dropped to the hot legs, and zero bied-rate probicas had been documented during the ND pumps lost suction (see XVI.B.170). In order to vent the loop, 6 pump start attempts were made the 2 pr:vious refueling outages. Due to the reser- while the system was not water solid. It was voirs Scing a non-pressurized type, vented to the believed that this might have caused water haenners in ato, tr. re was the possibility that conditions at the the ND piping, damaging or contributing to existing plant bring storige, maintenance and/or operation damage on the hangers.
centributed to the contamination problem.
Snubbera IMCA-ND-H177 & - H317 were replaced. Anchor Siace this problem was identified, they had developed bolts for hanger IMCA-ND-H260 were retorqued and a 4 pr ey va in watch all fluid reservoirs, both old and brace was added. Repairs to hangers -H273 & -H287 new, =m row letely flushed prior to filling with (moving baseplates to undamaged concrete) were not this cleanang process had revealed that tresh i h et.
u-w reser ma rs, obtatned f rom Power Paping Co. within completed within the 72 hr Tech Specs limit. The ND the prw ieus yrir, contained varying amounts of what 'A' train was declared inoperable on 10 Apr. On 12 appear..I to be welJ slag, small beads of metal and Apr, repairs were completed and 'A' train was They were also to institute a declared operable.
[tne metal tilings.
progra:a to .: heck th- fluid of any snubbers removed On 21 Apr, a m.nintenance worker on 733 ft elevation tot t.Tting .and/or rebutiding, for ev tdenc e of con- heard a loud ' crack' and watnessed the agitated move-iinued contamination. ment of ND piping throw dust into the air. Upon checking ND system supports, he tound th at the anchor
% ring the !bl retueling outage, wh-n thia problem bolts on hanger ND-H21 h.ad partially pulled loose s es (nona, all snubbers inside contsiament were com- from the wall and -H273 bolts had also pulled out, plately rebuilt , i.e., reservoirs ilushed, new seals, breaking concrete. This occurrence had coincided aew t l ad, and required hardware added. In addition, with the starting of ND pump '!A*. Further inspec-27 of the r.ena6aing 9) anubbers outside the contain- tion revealed hanger -H287 had pulled loose from the vnt were rebutit prior to Mode ! luwer operation. wall as well. /or some time before the swapping of From i review of piping analyses, tnese 27 snubbers ND trains, Pump '18' had been running with the RPV r+sid have .a t' fec t ed piping integrity due to zero Olved-rate . taring tuereal movement. the rot.uilding lovel at about N0 ft elevatio1. The 2 NP trains were sa the rom.nining 64 snubbeta was to be cornpleted by i IAt M. (ruol
< --c.c -
-e. .st 43
r_-
Vol. PWR-2 i
Vill. Aux Systems l
B. Aux Cooling
- p. 92 409. SW CONTROL VALVE SEAT DAMAGED l Salem 1 - Jul 81 - 97% power I have 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 Hz), the cctions were to review the pertinent plant procedurts operator noticed that flow control valve 12SW127 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 co.itrol valve was a Fisher Controls Co. 16 tions personnel. (mzp) in. Vee-Ball, type 478-6-16-U-CAV5 control 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 marred, restricting move-ment of the valve. No. Il CCW Hx was placed in ser-hillstone 2 - Jun 81 - hot standby vice, and No. 12 CCW Hx removed from service. Valve f 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 l on the "C" SW pump discharge pipe. 1he leak developed free operation. The valve was reassemtled, tested !
between the pump and pump discharge valve due to cor- satisfactorily, and returned to service within 22-1/2 rosion which was believed ta be the rssult of coating hr. The valve afr 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 125Wl27; it was removed and a new Hx was being replaced. (nam) installed. (nec pwd) 406. VENTILATION UNIT SW COOLING COIL LEAKED Cook I - 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 leak 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 b 1.5 gal /hr) from a 3/4 in. vent con-per cooling coil which was found to have a minor imper- nection veld on No.15 containment fan coil unit fection, oelieved to have resulted from the afg pro- (CFCU). The CFCU was removed from service and iso-cess. The af fected 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 planned. (nat) an existing design change 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 returne3 to FAILED service within 33 hr. The design change request was completed on No. I1, 12,13 and 14 CFCU. No. 15 CFCU Crystal River 3 - Jun 81 - 100% power was partially completed and was to be finished during the next suitable outage. (nex)
It was discovered that cooling fan (Trane) AHF-ISB for decay heat closed cycle cooling pump IB was exces- '
sively noisy and the fan was shut down. Investigation 411. LEAKIN", NESW CHECK VALVES TO BE REPLACED WITH revealed worn bearings in the fan motor. The sealed DIAPHRAGM TYPE VALVE bearings were replaced. The bearings were also re-placed in ANF-15A as a preventive measure. (nax) Cook 1 - Jul 81 - refueling While performing the B & C leak rate test, several 408. 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 I - 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. Il component cooling HK (CCHX), No. Il 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. Il CCMX to No. 12 CCHXe 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 CCMX due to the pipe nipple not 14 NESW check valves were to be replaced witn a dia-being 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)
_ _: a- Ju' 85
- . - - . . . , = . , + - , - . - - , - + - - - - , ,-7ws --------------g + T-, ,---r
Vol. PWR-2 XV. Misc Systems
- p. 114 been in service for 6 yr. The failed snobber was isolated on the discharge side of the 2 pumps. To prevent recarrence, a PM program wis These conditions, combined with the physical layout replaced.
of the ND system (top of the ND HX tubes was about to be instituted following the next refueltng to 758 ft elevation), could have contributed to voids rebuild hydraulic snubbers %. I through 4) and 02 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' during routine surveillance of pump was postulated as having caused the water on 27 May 82, hammer and subsequent hanger damage. hydraulic snubbers (Bergen Paterson), it was deter-mined that snubbers No. 32 and 45 (Engineerind Action taken included declaring ND 'A' train inop- Safeguards Pump Suction) contained insufficient oil erable, full-stroking of ND system mechanical snub- and were therefore declared inoperable. The snua-
%.rs, inspection of ND system hydraulic snubbers bers were found to have failed due tu deterior1*cd ud restoration of the damaged hangers. Further, o-rings. These snubbers ha d been in service for The f ailed snubbers were repliced.
SI (NI) system piping was inspected and mechanical 6 yr.
snubbers were full-stroked. Insulation at the nearent elbows to the damaged hangers was removed on 29 Jul 82, during routine surveilln:e of to allow closer inspection. Anchor bolts on -H287 hydraulic snubbers, it was determined that Bergen Paterson s mabbe r No. 3L (GC-8 Shutd Jwa C ioling LPSI
& -H21 were retorqued, restoring the hangers. Pump Suction) cont a inad insufficient oil and was Problems with -H273 could not be immediately T54 anubber was tnarefore declared inoperable.
resolved due to broken concrete around the anchor found to have deteriorated o-rings. This snubser balts. However, ND 'A' Train was improperly had been in service for 6 yr. The failed snubhr declared operable on 23 Apr, following an engineer-ing evaluation regarding the effecta of unrepaired was replaced.
hanger ND-H273. The engineering evaluation deter-mined that the ND piping was sufficiently on 21 Sep 82, during routine surveillance of Berg.en restrained and supported without the immediate Paterson hydraulic snubbers, it was Jetermined th er restoration of -H273; however, a factor not constd* Snubber No. 39 (GC-6 Shutdown Cooling LPSI hv there-ered in declaring the ND 'A' train operable was the Suction) contained insufficient oil til si.
snubber support function ot -H273. Tech Specs fore declared inoperable. Failure was found ta be required that ND sybteo safety-related snubbers be due to fluid loss through scores in we e/l;oJer operable in any mode retsiring the ND system to be wall of the snubber. The scores era attribut ed to operable. The hydraulic snubber of -H273 was a minor vibrations over a 6 yr period. The failed snubber was replaced. (rwc) safety-related 13,750 lb snubber, and was covered by this Tech Spec. With the 'A' train improperly declared operable. Unit I changed from moda 5 to Hydraulic t oubber 1.eaked - Piston mode 4 on 28 Apr in isolation of Tech Specs. O 423.
Restoration of -H273 was underway on 6 May when it Rod 0-Ring Failed was discovered that the 'A' train had been improp- Pt. Beach 1 - Dec 82 - 77% power erly declared operable. Work was completed and the O
'ianger and snubber were restored on 7 Apr. I memo-randum containing guidelines to insure performance A daty shift supervisor notice 1 an oil spill in the of all proper evaluations prior to removing any araa below the 'A' SG main stern line snubbers.
hangers from service ns distributed to appropriate The snubber oil reservo;r was chec ked and found to be empty. Also, a pressure gauge a t t a %=J to the personnel. oil supply line was reading slightly lower than Procedures were to be revised to prevent further normal. These findings led to the discovery Cist water hammers upon aligning to the NDS system dur* snubber No. iHS-1 was leaking oil past its shaft Loss of RdK system procedare was seal. The snubber was manufactured by Anker-Ho.tN, ing shutdown. Model No. 21.1250.007, and had a capacit/ of 50,J00 modified to indicate that the system had to be RHR oper- Ib.
vented between valves lhD-1B & IND-2AC.
attng 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 pwnps pri>r to swapping trains. This was intended although the oil reservoir was empty ther? The se to ftll any voids in the nnn-operattng train with- still oil in the supply line to the snubber.
out causing water hammer 4 (rva. s.u ) leak was located abou the cyliader ho ly anJ t hi snubber was still full at oil when it was removed.
It was found that the leak w as caused by a damaged piston rod o-ring. The snubber was repsirrJ i a 422. Ilydraulic Snubbers Inoperable - O-Rings Deteriorateu Cylinder Scored installing a new shaft wal kit, which inc i a t ed 4:
piston rod o-ring. After repairs, the snuwer sas Palisades - May to Sep 82 - 0% to 100" power functionally tested and returned to service wichtn O
2 days. As a result of this even:, the allowed on 19 May 82, during routine surveillance of interval uatil the next required visual snubber hydraulic snubbers (Bergen Paterson), it was deter- inspection was reduced from 18 to 12 mo in accur- (rwe) mtned that snubber No. 17 (main steasi line upstream Jance with Tech Specs.
of atm dump valve CV-0760) contained insufficient oil and was ther ef ore = Jeclared ino p.a r ab le . The snubber failure was found to be due to detertorated o-rtngs. The normal service life for the o-ii m was constdened to be 5-10 yr. Thts snuboer h ad c.** O a.e ==* **= a _ ov 83
C. MISCELIMEOUS (OPERATIONAL PROBLEMS) Vol. IVR-2 XVI. Oper. Problems C. Misc.
- 1. COVTAINMENT I?TTECRITY VIDIATION p. 1 Cinna - Jan 70 (1) During the period of steam generator inspec-9/ During a cold ahutdown to replace 6 failed reactor coolant system RTD's, both personnel airlock doors were opened to permit flow of maintenance tools tion, the minimum level in the primary coolant pipe will be 10 in above the centerline to assure sufficient NPSH to the RHR pumps.
and equipment. During partini draining of the reactor coolant system to replace the RTD's, it (2) A hot leg temperature indicator was added so was noticed that in-core temperatures were - 40'F that a more comprehensive monitoting of loop temp-above the Tech Spec limit for lack of containment erature could be attained.
integrity. One door was shut icnediately and the door interlock system was returned to service within (3) A temporary recorder to monitor for any flow I hour. fluctuations was connected to the RHR system.
To prevent recurrence, procedures were changed that Tech Spec violations that were probably violated prescribe running the RHR system during draining; were that containment integrity was not maintained replacement of one RTD at a time; maintaining con- above the cold shutdown condition of 200'F, and t nuous surveillance of in-core T/C si and setcing i
the coolant in the hot leg n):zle probably exceed-a max limit of 130'F on reactor coolant system above ed the heatup and cooldown rates of 60 and 50*F/hr which cooling must be initiated when containment respectively. (ft) integrity is not maintained. (fr. ft) 3. PRIMARY COOLAtTT DILUTION TECH SPEC VIOLATION
- 2. AIR IN RHR SYSTEM - TECH SPEC VIOLATIONS Ginna - July 70
""* * "*I During a shutdown to repair a steam leak in a The plant was in cold shutdown with both manwap 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 one of the "B" generator manways off. The loop plugs pomps. Work on the turbine stop valves used steam, had not been installed in the "A" generator. In or- I wering the coolant temperature. To maintain the 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***
g The RHR pump was running smoothly. A valve to the
- 4. DECOVTAMINATION OF ENTIRE PRIMARY SYSTEM refueling water storage tank was opened, flow and pressure were established, and the water level was raised 4 in. Shippingport - 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 pip 11g which is the of corrosion product activation within the core suction for the RHR system. and subsequent redistribution of the activated crud throughout the system. The Core I to Core 2 At one time a temperature of ~ 212'F was indicated. modification program, which consisted of extensive The level was raised another 14 in and RHR flow anu 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 proxicity 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 througn vortex-ing into the suction of the RHR system. This is what In additien, 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-was injected into the "A" loop hot leg through the tenance that might be required during these periods RFR suction line and that with the loop water level Of operation and testing.
raised, the system functioned properly and swept the remaining 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 s^* preventive measure = instituted were - To determine the ef fectiveness of decontamination on a nuclear power plant which has been in ser-vice > 6 years and had accumulated > 2 5.N0 FFril.
V;1. PWR-2 XVI. Oper. Prchl:as s:ures of sake-up witsr) opin1d brisfly. An attsmpt C. Misc. w s mads to virify 1sv21 et er (bove ths RTD mani-
- p. #18 fald which would hIv3 insursd prcper 1sv31 for purp operation. Water was reported present when the RTD manifold vent was opened, and the level in a tyl,on
" Operations Standing Order" was issued for both tube vented to the atmosphere agreed with the level ulits which prevented either unit from being operated gauge. Although cause for the pump's behavior could la conditions where the Specs were applicable until not be determined, it was returned to service as prrcedures couid 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 gpm.
Later, flow was throttled to just above 500 gpm after parameters had begun to oscillate again.
Af ter flow was throttled to 500 spm, level was N 1730. SUPPLY AIR RECULATOR FOR A/C THERMOSTAT FOUND MISAD M TED observed (using tygon tubing vented to the RTD man-ifold vent) below that specified for ND operation.
N Calvert Cliffs I - Mar 82 - 100% power The operator that checked level earlier had been misled by a small amount of water present in the Th3 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 thin the proper 20 psig. This caused a continuous As level was later increased, ND flow was increased, false high pressure output signal to the A/C com- without signs of air-binding, until level was 175 prsssor control pressure switch. The pressure in. and flow s 3,000 gpm.
switch would not have actuated to start the compres-ser as normally required by a rise in Control Roan The only instrument normally indicating water level tsmpe rature . The No. Il 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 wts available for operation. gauge was referenced to atmosphere because its design function was to indicate water level during Riadjustment of the thermostat supply air regulator refueling. Pressures other than 0 psig were pos-(Norgrem) restored control operation for No. 12 A/C sible during draining and filling because the allowing its cumpressor 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-rssponded 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 inadvertently adjusted, a facility change to add and the computer video display (Iow paessures could etmper-proof, locked metal guards to enclose pneumatic not be accurately read on this 0-100 psig control centrol panels and thermostats in the Control Room board gauge) to insure no pressure deviation occur ,
A/C System was to be completed. (pin) red that would significantly af fect indicated level. NC pressure could differ significantly from ~
PRT pressure if the drain or fill rate was high N IDI. R M PUMP AIR BOUND - 1.EVEL TRANSMITTER because the N2 must pass through a small hole in MISAPPLIED the PORY discharge line inside the PRT. When ND flow was throttled to 500 gym, PRT pressure indi-N McGuire I - Mar 82 - coM shutdown cated 0 psig with the drain secured. A pressure of 2.89 psig would cause a 170 in. level to indicate on 2 Mar, while draining the RCS (NC) System for SC 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 lA being tolerance.
stopped due to signs of cavitation. During the 50 cin 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 %
cut of service for manitenance. Upon restart, normal 105 to 1300F. A conservative linear extrapolation ND flow could not be maintained without signs of showed that ND could be shut down % 4 hr before etvitation reappearing. The control board level the onset of boiling. This boiling would not be giuge 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 operable) and any Ircops (175 in.was the minimum for ND operation). vapor generated was not released to containment However, 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 Isenediate corrective action was to raise level to specified 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-tecomplished 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 inaccurate indication by the transmitter for the psig PRT overpressure could greatly affect the 1svel gauge, due to a misapplication of the level levels indicated on the control board gauge. A transmitter. modification to have the reference leg of the level transmitter connected to the PORV discharge line To insure pump IA could be restarted safely af ter was planned. This avoided problems associated with thowing signs of cavitation, it was vented and the a pressurizer tap and was satisfactory since a PORV fueling water storage tank supply to ND valve (a was required open during draining. The modification T - Jul 82
~
t
V21. PWR-2
- IVf. Operational Problems C. Misc
- p. 714 3 2425. Cs Test Talve Opened - EE Water Dralmed verified. The closure of that damper had a direct '
Out - peroommel Error effect on the concentration of CO2 in the cable vault area. The last operability inspection of the N McGuire 1 - Jun 84 - 1001 power fire dampers was conducted on 21 May. Conduit was not found in the control room cable vault damper, A 3/4 is. vant 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-RRR 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-check 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 ane could cause safety and radiological concerne original data sheet.
by forcing high velocity water out of the discharge g vint. On 16 Apr, personnel placed a drain hose on An investigation revealed that personnel resetting g etch spray header to determine if the header was the dampers after surveillance tests or actual free of water. A drain valve and vent valve were actuations occasionally used conduit to hold the rpened to perform the task. dampers open, to aid in reattaching the latching mechanism. Not 'all situations that required reset-It was theorised 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 left open during preliminary draining on 27 Jun. dampers were operable and unobstructed following The pipe cap was found sitting next to the vent testing, maintenance or actuations that disturb a valve, and a reducer coupler and a 1/2 in. nipple damper's readiness. A safety evaluation tut were attached to the valve. Interviews with per- addressed the obstruction of the dampers concluded connel who were in the same room as the vent valve that adequate fire protection existed in both tnd 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. RER Trains Simitaneously Placed In Recire cauged high velocity audible air to exit the open Mode - Operator Error vint valve, if the vent valve was open. However, "]
background noise might have prevented personnel N Farley 1 - Feb 84 - cooldown /
from he6 ring the noise.
Before entering Mode 4, on 10 Feb, the shift super-The radioactive spill was cleaned up successfully, visor placed the RHR system in the recirc mode to without any workers receiving an above-limit dose. ensure that the boron concentration in the RHR sys-However, the performance test supervisor's shoe was ten was greater than or equal to that in the RCS.
contaminated when he closed vent valve INS-68. That required shutting the RRR to RCS cold leg MOV Operatlocs checked other vent and drain valves on for each train to prevent auto low head S1 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 operster decided to place both trains restoration procedure in removing station equipment of RHR in recire. Normally, the procedure was per-z.o t covered by an established operating procedure formed 1 train at a time. The decision was dis-vas 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 3pec 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 connection valves for train 'A' & 'B' on the RRR & that only 1 train at a time should be prepared for CS spray headers. Also, a review was to be done to cooldown until the unit was below Mode 3. (uhx) determine if independent verification might be applicable to other header valves. (uqe)
N 2428. Inadvertent SI - Wrons Titel Bus Synchronisina Switch Used - Operator Error l 3 2426. Fire Demper Obstructed by conduit -
l Personnel Error N Surry 1 - Mar 84 - cold shutdown l N Cook 1 - Aug 84 - 100% power While operators were removing vital bus transformer l
(SOLA Electric Co.) 1-I on 1 Mar, vital buses I&
l 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, l
veillance test of the carbon dioxide fire protec- causing spurious containment EP and high steam flow tion system. An inspection scand a short section signals, and the initiation of an $1 signal. The of conduit positioned so that it prohibited the vital bus synchronising switch for vital buses II &
damper 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 crosstied.
~4 l
Vol. PWR-2 IV1. Operational Problems C. Misc q p. 801
)
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 circuit.
4160V bus IF. All major loads supplied by that aus were de-energized. To prevent rec.:rrence , 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-aive viring changes but did require internal cabi-net wiring changes. The changes on Unit I had been N 2673. Contaiammet Intearity violated Durina impicnented. The Unit 2 modification was in prog-Bydre test - Sisseltamaans Test and Brain ress when the specialist doing the work identified Of M System and M R Eg - Procedure Error a problem with the wire routing and questioned how the electrical separation criteria could be main-N Cook 1 - Apr 85 - hot standby tainea. The technicians identified the train sep-aration problem when they noticed the differently On 12 Apr e an equipment clearance was being placed colored cables entering the bottom of the cabinet.
on the RRR 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 containrient 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 RER HI 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 containment components in the front of the cabinet. On the and the outside atu. The operator inside contain- front side of the cabinet, train 'A' wiring was ment noticed a greater than expected flow of water routed from the outside edges of the cabinets into and notified the unit supervisor. The supervisor the operating components. 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. Personnel 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 g Although the system never completed draining, a and required removing cabinet panels when routing potential flow path was established between con- new vires. The procedures used by the technicians
-) tainment and the outside atm fqrJ5 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 installation, but was not noted and viring were not color coded or identified as at the time. To prevent recurrence, 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. (vtt) ing, an electricain accidentally depressed a limit switch inside the reactor trip breaker 'A' cabinet which initiated a main FW isolation signal.
N 2674. Reactor Trip Switchaear Wirina Out of Spec -
Procedcral. Persommel Errors An admin and procedural deficiency was found to be the cause of the incident becauset 1) the wiring N 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 spec e s and 2) the technicians in-ciency was discovered in the reactor trip switch- volved in the modification had not been trained on gear cabinets. The error involve 1 a single train the installation specs manual. A personnel error
'A' vire which had been routed with the train 'B' caused the unplanned main W isolation signal.
wiring during a previous 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- E 2675. control noom Air Intake nadiation Element cult to the auto breaker trip circuits. The spe- typassed Too Lona - Persommel Error 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 j power to operate and if the single fuse had blown because of an aux circuit component failure, then element CK-RE-05 was bypassed for routine filter replacement. Tech Specs allowed that condition for the shunt trip circuit would have been inoperable. $, I hr, af ter which time the control room emergency ce,- Aug 85
ng 1 Vgl. PWR-2 VIII. Aux Systems CORRODED TOROCE SWITCH, MALADJUSTED POSITION B. Aux. Cooling N201.
LIMIT SWITCH ON SERVICE WATER OPERATORS 198. SW PUMP BEARING WORN - COOLING LUBRICATION N Oconee 3 - Aug 78 - 100% power 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 coming from a service water (SW) pump (Wheelabrator- Reactor Building cooling units (RBCU) from service.
Frye) found the seal assembly was out of position. A unit shutdown (10%/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-cut of service, placing the facility under an mined that 3LPSW-21 was inits full open position and LCO. Only 2 SW pumps were needed during post LOCA the apparent valve failure was actually a failure
-cInditions; 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 cause of the inoperability of 3LPSW-24 was the end returned to service within the time allowed failure of the Limitorque valve operator torque by the LCO. The failed 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 tne apparent inoperability shift. This failure was believed to be caused by of 3LPSV-21 was the failure of the valve's position insufficient lubricating / cooling water supply to indicator to properly show the valve's position thi bearings. A vendor design change to the because of maladjustment of the open/close limit pump bearing cooling water supply was being eval- switch (which was readjusted). (gnd) ulted and was to be implemented when possible.
(gko)
N202. RHR vAtvE STEx D1SENCACED rROn D1SC 199. RHR PUMP SHAFT SLEEVE CASKET LEAKED, SEAL FACES WORN N 8" "7 ' - 3""' 78 - ""*
ituring 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 Storage Tank did not appear to be opening when act-During preparation for plant startup, No. 32 RHR uated. Further checking showed that no water would Pump Seal was found leaking. The pump (Ingersoll- pass the valve when the valve was opened manually.
Rzad model 8X20W) was removed from service to The valve was disassembled and was found with the idIntify the cause of the leakage and No. 31 RHR stem disengaged from the disc and the disc fully Pump was verified operable. Upon disassembly of seated. During normal operation this valve was shut.
ths No. 32 RHR Pump Seal it was found that the It was opened only to provide a means of pumping shsft sleeve gasket was leaking and the seal faces the reactor cavity to the RWST following refueling.
v1r2 votn. The pump seal and gasket were re-pitced. (gls) The valve (Aloyco) had been previously disassembled during the outage for maintenance. The valve was 200. assembled by a qualified workman using an approved RHR PUMPS AIRBOUND proceoure. 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 in the assembly and was working according Tha RCS was drained to the centerline of the to an approved procedure. The valve was reassembled cold legs. The primary side manways on SG's A and operated properly. (ghg) and B were removed for ISI and RCS temperature wts M 1450F with the "A" RHR Pump in service when indication of low RER flow was received. The DC START TIME EXCESSIVE - C00LINC WATER VALVE "A" RHR Pump was running with no discharge flow. N203. SEAT LEAKFD Ths "B" RHR Pump was started but it also produced no discharge flow. An investigation revealed N Calvert Cliffs 2 - Mar 77 - 10n?. power botn RHR pumps were air bound. The "A" RHR Pump ssels and casing were vented and the "A" RHR Pump During a weekly test DC No. 21 failed to reach wis returned to service tv1 hr later. rated speed within 10 sec. On 3 successive attempts.
the times to reach rated speed were 19.6, 16.0,
, With mixing of the RCS re-established. Tavg ar.d 11.0 sec. The service water inlet valve w:s determined to be N 1750F. The heatup from (Masonellan) 2-CV-1587, which provided cooling 1450F to 1750F for the water volume in the water to the DC was found to t'e leaking excessively reactor vessel was consistent with the expected by its seat. This leakage resulted in low jacket hiatup rate for 2 IW of decay heat which wts typical for 38 days into a shutdown from water cooling temperatures and thus was the apparent full power operation. The RHR pumps were cause for the sluggish operation of the diesel. The bsing vented periodically and the RCS was to be valve was rebuilt with a new seat which corrected the refilled upon completion of maintenance activities. leakage problem. The DC. had was successfully test started 5 times, over a 4 wk period, after this (git) repair. (gnh) s - e =- a.--
s - =. hv is
iraining a new slbow rnd e u nuti the wasr rings wara irrsgular in locatien rnd did aor
- m!v) 743 w:s inst:112d. propogita tround ths ixptller wrer ring area. Erch m af this new piping siction, th? Errk htd a stsrt tnd stop point, which wts typical of -
i hyderd ct 625 psig. T;sts w2ri som debris bring tripped in th2 annulus. Shsft Jerformed on the new manual globe inspection by PT showed no indications. The inspection nd s;cti n was analyzed. and the covered the whole shaft and concentrated in the keyway ta b2 c vitation and/or impinge- area near where previous failures occurred. Hub material
!y w n initiated to determine if was removed from the B pump impeller to further assure lud2 rreurrence. (8pe hbs) adequate clearances as on the ' A' pump. NRC's review concluded the pumps were reliable and acceptable. (gsb)
T FAILED - CASING fuPELLFR RASED 212. SERVICE WATER VALVE ACTUATOR LINKACE N"T MISSINC 3 - Apr 78 - refueling Davis-Besse 1 - Oct 78 - 100% power a low flow condition existed The Control Room operator unsuccess fully attempted to 1 Pump (DHP) 1-A. Upon dis- open the CCW HX No.1 Service Water Outlet Valve (Hammel m rsd thtt the pump shaft Dahl) SW 1424 to control the HX outlet temperature. He
,p111st end. The pump tfr. then attempted to manually stoie or operate SW 1424, 1sulted and pump shaft break- but it remained closed and CCW Loop No. I was declared initleted. Due to the severity inoperable.
'nd occurrence (see VII.B.14B) pump, tha reliability and Upon visual inspection it was discovered that the nut
- ypt of pump was in question. on the actuator linkage for the valve was missing. The tcps ba taken to establish valve had failed ia the closed position since the 11ty (nd operability of these positioner was not connected to the actuator. They replaced the linkage arm bolt and reset the stroke. The valve operator was recalibrated, and the position
- htft friture was primarily switches set. Af ter satisfactorily completing stroke to-impiller rubbing. Sub- times, the valve was declared operable
- Ik hr later.
rabuilt with a new shaft wrs shivsd back to assure proper R Other SW valves associated with CCW HXs were inspected essing (nd the impeller. A for similar conditions with no problems noted. See
'orm d on this pump to establish R vIIr.B.293 for adettionat information. casv hbh>
h cxistad was no longcr present.
>trined utilizing accelerometers on th] pump, connected piping 21 3. LPSI PUMPS AIRBOUND FOLLOWING TRANSFER OF The ccquired data was reviewed PURIFICATION SYSTEM RESIN cnd noise source identification e,th;r dIts sets to identify Calvert Cliffs 2 - Oct 78 - co'd shutdown o end trcnds in performance uits indicated that the A LPSI pump No 21 operating in shutdown cooling will cs the B pump. Also the (SDC) mode began cavitating and lost suction. LPSI on thi A pump casing showed pump No. 22 was immediately started but it too began BIsid en this data they cavitating and was secured. Both pumps were sub-w:s ni non-lubricated me; =1 sequently determined to be airbound. The pumps were ctsd by sxtremely low no: e friqu ncies, no high fr- uency vented and SDC was restored 2 hr later. ,
}f 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 system resin while the purification and SDC systems were cross connected. (gsw)
In July, the Unit 3 pump malfunct was dismantled and the shaf t was broken at the keyway. The pump t tested satisfactorily.
214. PUMP SEAL FAILED - MOTOR BEARINGS DAMAGED In July, the thrust bearings seit pump. The pump was dismantled as Surry 1 - Oct 78 - 100% power overheated thrust bearings. The replaced and the pump tested sat:
Chargicg rump Component Cooling Water Pump (Ingerno11 Rand) 1-CC-P-28 tripped on overload. The exact cause of the shaf t and The redundant pump was placed in operation. was not determined. It was suspe They found the overload was the result of corrosion have resulted from excessive inte by pump cavitation.
damage to the motor bearings caused by water leak-age from the pump seal. The failed pump was removed and disassembled for repair. The motor bearings A valve located at the SFP heat e were replaced and the pump rebuilt with new seal is called the " system discharge I material and new gaskets. The rebuilt pump was valve". Operating personnel veri reinstalled and tested satisfactorily. (gwx) operation was very noisy, indicat whenever this valve was > 50% oper conducted an evaluation of systen 215. determine the optimum position of SPENT FUEL PIT COOLING PUMP SFAL FAILED - charge line control valve. Evalu INSUFFICIENT LUBRICATION - RELEASE indicated they could avoid cavita provide cooling system flow to me Turkey Pt. 4 - Sept 78 - refueling rent SFP heat transfer demand by A spent fuel pit cooling pump's mechanical seal limiting the valve position to ab open, failed causing leakage of radioactively contaminated
( l 294. CORE FLOW LOST DUE TO RHR PUMP AIR BINDING Vol. PWR-2 VIII. Aux Systems Beaver Valley 1 - Jan 80 - refueling B. Aux. Cooling
- p. 67 The plant was in Operational Mode 5 with RCS temp at 100 F 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-l vessel vent eductor was in service in preparation for refueling. A low flow alarm was received and ance ef forts to open POV-5 continued. At 0005 hr on 11 Mar, POV-5 was opened and the SWCP C-13A low flow and low motor current were indicated. placed in service. At 0010 hr, with the pumps C-13A A second pump was started and was also air-bound.
ahd G-13C determined to be operable, the pump Cal 3C The pumps were vented, core flow was rapidly re- was stopped and unit operation proceeded in accor-stored and there were no implications to the health dance with the LCO. 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 below the alarm condition setpoint of 97'F.
was in service and caused a negative pressure in Tem-the RCS because the eductor air volume was greater peratures of pump bearings cooled by the CCW system than the influent air. This dp overcame the dp were monitored during the incident and no signifi-in the steam generator U-tube section and the cant changes observed.
excess steam generator water and entrained air drained into the loops. This entrained air later Investigation revealed that the SWCP C-138 shaf t 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 were repaired and the pump returned to service.
vessel vent eductor into service. This had been The failure of POV4 to open automatically and the a recurring problem when loops were drained to mid-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 On 18 Mar during maintenance activities on the SWCP constantly vent the RHR pumps automatically. C-13B, pipe support 9-416-14 was found to be in a Qmm) degraded condition. Corrosion was evident on the anchor bolts that attached the vertical support legs to the face of the concrete pedestal. These 2 295. SWC SYSTEM FAttERES - PtPE SUPPORT ANCHOR BOLTS LOOSFNED, VALVE 6 PUMP 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 90 - 1007. power existed between the face of the concrete and the structual member. The pipe support was redesigned At 2115 hr on 10 Mar with salt water cooling pump and rebuilt to transfer loads to the floor slab
-C (SWCP) C-13B in operation, SWCP low flow and low rather than to the concrete pedestal. Following completion of modifications made to the SWCP line discharge pressure alarms were received on the main support, SWCP C-138 was returned to service on 21 control room aux board annunciator panel. Con
- Mar.
currently, SWCP C-13A automatically started due to low pressure in the discharge line of pump C-13B and pump C13B motor amperage was observed to be in- Subsequent inspection of the corresponding pipe
,yppogg, g ,.15-12, on the SWCP C-13A line revealed dicating low. They found both pumps were running with discharge pressures of 0 and 40 psig indicated some signs of deterioration; however, this support at pumps C-138 and C-13A. respectively, and the was in full contact with the concrete and determined pneumatically operated discharge valves (POV's 5 & to be operable. On 12 Mar SWCP C-13A was removed
- 6) uf each pump were in the closed position. POV-5 from service to install a modified support. The (discharge of pump C-13A) should have been open while modifications were identical to those made to s;pport POV-6 (discharga of pump C-138) should have been 9-416-14 Installation was completed the same day.
closed. Ef forts 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 perable during these repairs. (jro) from the control room. However, a low flow cond!-
tion as indicated by low pump motor amperage wss observed. Investigation of the aux pump and piping 296. RHR FLOW t.0ST WHILE INCREASING FLOW -
system indicated that the low flow condition was PUMPS AIRB0t'ND, FLOWMETER RECALIBRATED due to apparent insuf ficient pump priming and the aux pump was then stopped. To re-establish SWCP teaver Valley 1 - Apr 80 - cold shutdown flow, the screen wash pumps were started from the 4
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 81oF E-208 normally served by SWCP C-13A. At 2133 hr, and the "B" RHR Pump was in service. A com-SWCP flow to E-208 was observed to be ~ 20M gpm plete loss of RHR flow occurred when they were attempting to increase RHR flow from 1000 to 3000 and CCW te perature exiting E-20B was decrsasing' having reached a peak value of 82*F. At 2156 br, gpm so the RCS could be recirculated through adequate priming was restored to the aux SWCP C-13C, Mixed Bed Demineralizer CH-t-1A. The resin in aid 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 Specs required 3000 gpm core flow when diluting.
During the period described above, an LCO was not cet and prepirations were made to commence 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 WCP 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 -e a -e- = ?tav W
i VJ1. PWR-2 N 2". satwarra/stavtcr W^Tra "x rust trixto u C N cat"n inns 2 - su 80 - 70s power
- 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 dX 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 was 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 FLOd TO CHARCINC PUMP COOLERS INADEQUATE to require an operator to vent at all times during this evolution. See VIII. B. 294 for a previous O North Anns 1 - Jan 79 - shutdown similar occurrence. (jsj)
N ue vIII.e.30s for additional infor-tion. A sPecial test was conducted on the Service Water sys-tem (SWS) to determine the flow characteristien 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 pump out for maintenance and fail-Component cooling water containment isolation ure of another to start, the design flow of 37 gpm to valve 626. RCP thermal barrier outlet, failed to the charging /high head SI pumpa 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 B containment isolation. The locknut was skid piping, and lube oil cooler temperature control 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 t44 hr. and associated instrumentation, piping for the HX's and instruments associated with the charging pumps.
Various Limitorque SMB valve operators had main- (jvy) tenance performed on them during the cycle VI-VII refueling outage as follow-up to IE Circular 79-4 (see XV. 180) The operator-to-valve stem locknut N 301. sOCxrT-Wrtm o ' s= - rtTTINcs sw0wro ca^cxs -
for valve 626 was staked on 7 May 79. The lock- GENERAL MFG PROBLEM nut backed over the stake resulting in failure of the valve to close. The locknut was staked N St. Lucie 1 - Apr 80 - refueling again and the repair was considered adequate. ,
As additional action, a representative sample of PT of 45' socket welded elbow-fittings (A182 sch. 401/2 the total locknuts staked during the 1979 refueling in. SS) in a non-class system showed cracks at the same outage were to bt checked to verify similar location in 3 fittings. This was reported because the i problems were not occurring on the other valve fittings were certified for use in class 2 (safety-operators. Additional corrective action was then to related) systems. The cracking appeared generic and PT be taken if the results of these inspections indi- 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 required 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 oPrRATOa ex0rittrito esvraOaxestatty cracking. Three other fittings (dif ferent heat) were ordered on the same Purchase Order but 2 were acceptable N Oconn 1 - Mar 80 - IM% power and the third was inaccessible. The fittings were or-dered in 1977 No other deficient fittings were on site.
The -otor operator (Rotork model 16 NA2) for valve ILS105 (installed in 1976 as 1 of 2 flowpaths sided The inaccessible fitting was associated with the 1A In-to the decay heat line to assure sdequate flow from take Cooling Water Pump (ICVP). It was located on the the RCS past-LO M to prevent boron precipitation) lA ICWP lube water strainer drain line, d wnstream of was found to be unqualified for in-contain ent use an isolation valve and was open to atmospheric pressure.
during a review required by IE Bulletin 79-1B (see (jvm)
Vol. BWR-2. IX. E. 429). The valve operator was to be replaced during the next available outage.
(jud)
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E, I 302.
DH PUMP LOST SUCTIOV - LOW RCS WATER LEVEL -
LEAKINC VALVE READJUSTED Vol. PWR-2 VIII. Aux Systems Davis-Besse 1 - Apr 80 - 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 304 the hot leg piping centerline as read at the Re- ADDITION To VIti.B.296 re RHR PUMP fueling Level Indicator L1214 (This lower level corresponds to 78 in, above the top of the reactor Beaver Valley 1 - Apr 80 - cold shutdown care.) The low RCS level caused an erratic decay heat (DH) flowrate due to inadequate DM pump suc. The SC's were drained and RCS level was mid-span in tion conditions. As a result, DH Pump 1-2 was the loops. RCS temperature was 101'F and the "B" stopped at 1920 hr. At this time, the RGS level RHR pump was in service. A complete loss of RHR flow 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 and found water flowing beginning this evolution pump flow dropped to from DN Cooler 1-1 Discharge Line Leak Test Connec- zero as the pump became air bound. At thin time, the "B" RHR pump was shutdown and the tion Isolation Valve DH73, which hsd been opened "A" RHR Pump to facilitate draining of D1 Loop 1. D973 was im- was started. "A" RHR Pump showed no flow also so mediately closed. Further investigation of the it too was shutdown. Both pumps were then vented loop isolation lineup revealed that D*1 Cooler 1 2 plus the RCS loops levels were increased. The "B" Discharge to Makeup and Purification System Iso- pug was then started satisfactorily and normal lation Valve DH62 was 2 turns open. This salve. 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 mskeup 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 was not attri-System to restore the RCS level to 75 in. DH flow buted to lack of venting because the flow was being to the RCS was restored when DH Pamp 1-2 was re- diverted, but not increased. A procedure was im-started at 1949 hr. Trippicg D9 Pump 1-2 placed plemented for a total loss of RHR flow and a con-the unit in violation of the Tech Spacs which tinous vent hose was installed and was to be put required that while in Mode 5 at least one reactor into operation as soon as possible (jyw) coolant loop must be in operation with an associated D or DH pump. The DH pump was off for 29 min and U,$ temperature remained significantly be,'ow the N 3os- v^tvt oost= 't"" Sw"c"ts amt=
ADJUSTMENT saturation temperature, rising from ~ 93 to 103'F.
The occurrence was attributed to component failure N St. t.ucie t - May Sc - refueting kg of the vstve operator. The remote valve operator L
(RVO) for DH62 was found to be out of adjustment See XI. A.400 for information. (kbx) can ing the valve to remain I to 2 turns open when the RVO was on its close stop. DH62 was to be ce. N 306.
adjusted. vAtvE OeERATOR RADtUS ARM OmTRUCTED w (jxf) MOUNTING P.0LTS 303. Davis Besse 1 - Apr SO VALVE OPFRATOR LIMIT SWTTCH MISADJUSTED While attempting to switch to #2 Decay Heat Re-Zion 1 - Apr 80 - 1007. power moval Train for maintenance on #1, Decay Heat Re-During surveillance both service water motor operat. moval Cooler No. 2 outlet flow control valve DH14A would not open beyond 207.. The improper operation i
ed valves (MOV's) to Aux W ( AW) pump IC backup of valve DH14A was attributed to a mechanical com-l 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 AW pump was already out of ing bolts allowed only marginal clearance. Any service for maintenance, 2 of 3 AW pumps were not movement of the radius arm towards the mounting operable, as required by Tech Specs. One of the bracket caused binding and limited movement of the limitorque (Type SMB Size 000) limit switch rotors radius arm. The radius ari on the valve shaft was of IMG-SW 103 was found out-of-position. It was moved away from the mounting bracket, thus provid-reedjusted and the valve was tested satisfactorily, ing clearance and allowing the valve to stroke They investigated the total operating circuit of fully. After meeting stroke time acceptance cri-IMOV SV 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.
(kek) i l \
- 'Sw=* e- July 50
a Vol. PWR-2 V111.* Aux Systsrs B. Aux Cooling steam void in the system. He also noticed the A p, g g loop wide 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 9nce through SC (OTSC), and McGuire events in order to ascertain it they cooling the A loop hot leg. About 7 hr later suf- brought to light any conditions which were not ficient cooling was achieved to drop pressurizer c nsidered during review and revision of natural level. The plant depressurized without void form- circulation and shutdown cooling procedures
up using RCPs. While reducing RCS temp and pres- * * " "E " " " * "E '" **~
sure to achieve a cold shutdown condition, a steam Cf81DIDE CI88888* (8yp) 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 SW LEAK AT CONTAINMENT VENTILATION have been for an operating plant. The system was COOLING COIL DRAIN VALVE repressurized, the RCPs ptamps were restarted to uniformly cool the system,and the plant was taken 2 - M 81 - WW to cold shutdown.
These events were apparently caused by insufficient 4 small SW system leak was found in the contain- .
ment ventilation cooling coil drain valve. The cooling of the large masses of hot metal in re- leakage amounted to 3 drops per 5 seconds from gions such as the RPV head, upper J leg (B&W), a bonnet to body leak on a i in. drain valve.
SG walls, and RPV nozzles prior to initiating They tightened the loose valve bonnet and,on DHR cooling. Local stagnation and stratification 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 blem, as required by IE Bulletin 80-24. They requested clarification of the reporting requirements of IE Bulletin 80-24 from the NRC. (mzl)
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- 404. RRR PUMP CAVITATED - INDICATED RCS per head region was essentially stagnant. This WI' NG
] allowed the coolant to be heated to saturation
'J temp, or to remain at relatively high temp because Trojan - Jun 81 - cold shutdown of the stagnant conditions. This resulted in steam void formation when the system was depressurized. While reducing RCS level, with the RCS level The operators had correctly diagnosed the system standpipe indicating % 66 ft, the RHR pump voiding at both Crystal River 3 and McGuire i was cavitating as indicated by fluctuating and took appropriate corrective actions to bring motor current. Diverting to the hold up tank the plants to cold shutdown. was terminated and, when the motor current did not stabilize, the RHR pump was stopped and While these events were not a concern because of RCS charging was established.
the availability of the DHR system at both Crystal 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 vent valve isolated. The vent was opened and in the RCS was a concern if operators fail to recognize these conditions. Also, normal natural standpipe level begaa dropping rapidly. The circulation flow was reduced by voids in hot legs RPV head vent was opened and standpipe level stabilized at % 61 ft. The RHR suction tapped and sufficiently large RPV voids. of f the RCS piping at just above the 60 f t level.
Previous experience had indicated that cavitation IE Circular No. 80-15, Jun 80, and NRR Ceneric may occur when RCS level approached that point.
Letter No. 81-21. May 81, were directed at the possibility of voids being formed in the RPV head An attempt to restart the RHR purp failed due j region durinR natural circulation cooldown. It was to air entrained in the RHR suction line. The i considered important that reactor operators RHR hot leg suction isolation valves were closed recognized that voiding could occur in other por- and the RWST suction valve was opened in order tions of the RCS (e.g., in the "J-leg" of B&W to provide a positive suction head to the RRR reactors) and under conditions other than natural pumps. Both pumps indicated flow when jogged circulation cooldown so that timely and correct and the normal valve lineup was restored. The action could be taken. RER pump was successfully restarted 75 min after it was stopped.
NRC reconenended these actions for holders of an operating license and near term operating licensees: The cause of this occurrence was determined to
- 1) Review operating procedures dealing with plant be personnel error. The procedure for draining cooldown and emergency and/or abnormal procedures the RCS had specified that the pressurizer vent that address natural circulation to assure that valve be opened. However, personnel performing sufficient information was available for operators 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 pressuriser condition. Special attention was directed to the vent path lineup, the RCS standpipe level would
==-a==== Aug 81 i
l vil.PWR-2 XVI. Optr. Problems ,
C. Mi n. all personnel not to move the setting unless authorized
- p. 174 by the shift foreman. The OWP regarding this limiter and the governor system was to be revised to N m. sravict WArta Sv5Tra at = =Cv tost sv incorp= ate a test to ensure proper settins on the IMPROPER TACCING limiter prior to returning the pump to service. ,
N s,ine vanwee - xay 78 - 93% power a
One of the 2 redundant Service Water system headers ""^'" t " - 'c5 """ '"5' we incorrectiv isoisted and tagged out to allow minor N'o5-maintenance. The resul t of the incorrect tagging u Trojan - Apr 78 - cold shutdown order was the loss of redundancy of the Service R Water pumps. Within 3 hr the incorrect lineup was RCS flow was being supplied by the luiR pumps with discovered and corrected. The operator who the plant in a partially drained condition. In-wrote the tagging order failed to realize the signift- mediately prior to this occurrence the RHR pump cance of Isolating the header and both that header's discharge flow had been shifted from reactor cool- i pumps. The incident was discussed with the personnel ant loop 1 and 2 cold legs to loop 3 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-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 TaIPetD ou ovraSPttD - Covranoa tI*ITra changes. About 20 =in after the RHR flow was READJUSTED - VALVING tRROR shif ted, the Control Operator received a report that level was increasing in both Containment sumps.
N Robinson 2 - Msy 78 - 100% power Concurrently the Control Operator cbserved oscil-laftons on RER pump actor currents. He immediately 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.
FV 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 RER pump was secured. Flow Work Permit (OWP) and an operator was dispatched to dropped to < 3,000 spa (the Tech Spec limit.)
turn of f the breaker for valve V2-16A and tag it out The RHR 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 (steara driven AFP discharge valve to SC A) was needed and this test An investigation revealed that a drain valve on the was implemented. Meanwhile, the operator in the field SI test line near valve CV-8879D was open. To pre-noti fled the control room that valve V2-16A breaker vent recurrence all vent and drain valves reposi- ,_
was off. Minutes later as valve V2-14A closed at tioned for work done under a systes tagout were /
4 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. (fsks inability to feed SC A with any AFP's. Normal feed at the time was being supplied by the main FW ,
pumps. A violation of a LCO was believed to have 5""c" i
eccurred and a reduction in power of 3/4%/ min was N'6- "'S 5 5""' '"" - 5 5"*'
MISPOSITIONED l started. The breaker for valve V2-16A, which had not been disabled, was returned to service. The Zion 1 - May 78 - 891 power power reduction was halted at 95% and the plant was returned to full power. It was later determined While testing one of 4 pressurizer pressure Trans-that no violation of the limiting condition for sitters (IPT 457) the control selector switch was operation Sy Tech Specs did result untti 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 psiJ. Pressure reduction stopped During 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 Hechanics were reinstructed l 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 (fal) could limit the speed of the steam turbine driving the AFP. It was normally set to some value below - N 6M . DC OIL PRtSSURE CACE LINE I.EAKED FOLLOWING the overspeed trip. This was done to prevent the CALIBRATION - LOOSE FITTING speed surge on starting, or when suddenly going to N reduced ficw condition which caused the speed to 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 DG No. 2 was speed was regulated below tne limit by the pump dis. removed, calibrated and reinstalled on 1 May 78.
charge pressure input to the governor. It appeared During a test on 1 June the generator was shut down l
j that af ter the above test, however, the limit ad. and declared inoperable because of a lube oil leek Justment of the governor was lef t set too close in the engine control panel. The leak resulted l from a loose Swagelock connection on the lube oil to the trip speed. Therefore, when flow through l
this pump was reduced suddenly, during timing of pressure gage apparently caused by not properly valve V2-14A, the pump did overspeed and tripped, reinstalling the pressure gage after calibration.
j The instrument line was installed correctly and DG l
The speed limiter was readjusted to its correct No. 2 was returned to operable status. Personnel were instructed to verify correct instrument instal-i setting and the system was returned to service.
f 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)
,. July 78 s m.e -
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