ML17264A967

Forwards Questions Re SG Tube Rupture Releases,Probability of Power Recovery,Containment Isolation Failure & Penetration Seal Failure

ML17264A967
Person / Time
Site:	Ginna
Issue date:	06/18/1997
From:	ROCHESTER GAS & ELECTRIC CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9708050174
Download: ML17264A967 (11)

Text

CATEGORY 1 REGULATOlINPORMATION DISTRIBUTION +TEN (RIDE)

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DOC.DATE: 97/06/18 ACCESSION'PR:9708050174 NOTARIZED: NO DOCKET FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G 05000244 AUTH. NAME AUTHOR AFFILIATION Rochester Gas & Electric Corp.

RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)

SUBJECT:

Forwards questions re SG tube rupture releases, probability of power recovery, containment isolation failure s penetration seal failure.

DISTRIBUTION CODE: AOOID COPIES RECEIVED:LTR Q ENCL J SIZE:

TITLE: OR Submittal: General Distribution T NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72). 05000244 I'-.,"

RECIPIENT COPIES RECIPIENT COPIES:, .

ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR 'ENC'L' PDl-1 LA PD1-1 PD 0 VISSING,G. R INTERN: F~E CE.NgE 01 NRR/DE/ECGB/A NRR/DE/EMCB NRR/DRCH/HICB 1 NRR/DSSA/SPLB NRR/DSSA/SRXB 1 NUDOCS-ABSTRACT OGC/HDS3 0 EXTERNAL: NOAC NRC PDR D

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JUN-18-1997 14: 13 FROM R.G. KE. DOCUMENT CONTROL TO 913814152182 P. 82

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f Question 6: Steam Generator Tube Rupture (SGTR) Releases The probability of the steam qenerator (S/G) atmospheric relief valves {ARVs) failing to close is determined in the Level analysis. There are essentially 3 cases in which the ARV on the l

ruptured S/G can fail to close. The first is the ruptured S/G is not isolated such that the S/G rapidly if feedwater flow to overfills and the ARV relieves water. In this case the ARV is assumed to stick open such that rapid cooldown to RHR shutoff head ARV on the intact S/G fails to open.

if is required. The second case is isolation is successful but the Xn this case, operators are instructed to use the ARV on the ruptured S/G to cooldown the RCS.

In this case the probability of the ARV on the ruptured S/G f'ail ing open is determined by the Level 1 data analysis portion of the PSA using plant spsc".ific Rata updated with generic industry data. The failure probability for the ARV to reclose following a steam release is 8.53E-04. The third case is one in which the operators fail to cooldown and depressurize the primary system prior to overfilling the ruptured S/G due to a failure of the PORVs to open.

Again, due to a liquid release through the ARV, stiak open.

it is assumed to Zn thc Lovol 2 analysis, no adieu tmcnt was made tc account for increased failure probability due to harsh conditions. The possibility of cLcbris entrained in high temperature gas being transported from the core through the RCS piping, through the ruptured tube which could potentially be under water, and up through the S/G and its moisture separators (which are designed to remove droplets or particles entrained in gas) was not considered to be credible. It should be noted that following the Level 1 requantification, the contribution to CDF from SGTR sequences dropped from approximately 334 to 164 such that this issue is of

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I Question 7: The Probability of Power Recovery As stated in the January 15, 1997 response to the RAI, new power recovery curves were developed for the Level 1 resubmittal. These new curves were used to develop power non-recovery probabilities for the Level 2 analysis. The preliminary results for the four cases of interest are shown in the table below. The table shows the time to vessel failure and the time to containment failure, along with the power non-recovery probability associated with each of those times. The large and, medium LOCA cases are evaluated, separately from the other SBO cases because in the Level 1 analysis large and. medium LOCAs coincident with a SBO were assumed to lead directly to core damage and. did not transfer into the SBO event tree. The two SBO scenarios take into account whether the TDAFN pump starts and runs (i .e., the f erst branc.h in the SRO event tree). Although PDS binning is not yet. complete, these non-recovery probabilities will be used in the binning process. Note that the Level 1 event tree for SBO includes the potential 'for power recovery prior ta core damage which will affect the binninq process.

VF CF VF Time N.R. Prob N.R. Prob Large LOCA ~ 9 klan'5 ~ 383 12 hcs. ~ 055 NA8$ >>m Ig)C'.0 1.6 hrs. ~ 267 13 hrs. .090 SBO (no AFW) 4 hrs. ~ 09 21 hrs. 267 SBO (AFW for 6 hrs.) 13 hrs. ~ 024 25 hrs ~ .985

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913814152182 P. 84 Question 9: Containment Isolation Failure The probability of containment isolation failure was determined by quantifying the containment systems event tree {CSET), which includes a heading for Containment Isolation failure. As previously stated, the five areas in NUREG-1355 are evaluated in detail. Specifically, items 1 through 4 form the basis of the containment isolation fault tree which is quantified (item 5) . As in any Level 1 fault tree, the model includes the appropriate top gates (failure of the pathways determined in item 1), all supporting systems (motive force for valves and signals required as determined in items 2 and 3), and plant specific failure rate data and testing and masntenance c3ata (item 4). Bectaon 3.).1.3 of the original submittal discusses this in more detail.

Preliminary requantification of the Containmeat System Event Tree (CSET) indicates that the current percentage of non-containment bypass core damage sequences which result in containment isolation failure is 3.0% (down from 5.24 in the original submittal). bf this 3%, approximately 1/3 is a result of the mechanical failure of AOU 371 to close during LOCA sequences where sump recirculation using the RHR system is required. Section 6.3.6.4.N of the 1/15/97 submittal discusses this failure path in detail (note that there is a typographical error in that paragraph; the phrase "which ovorflows to the Auxiliary building sump>> is inadvortontly repeated). Another 1/3 of the 3C is due to the failure of MOV 313 to close due to a loss of DC power on train A. This failure only leads to containment isolation failure if pressure in containment exceeds 85 psig such that the relief valve on the VCT opens creating an open path outside containment, or, if there is a failure of CVC3 piping. ft was conservatively assumed in the CSET quantification that if containment spray and the containment recirculation fan coolers (CRFC) fail, pressure in containment will exceed 85 psig. The remaining 1/3 of the 3R involves various other ranQom Cailurv c;vmbiclalivns.

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JUN-18-1997 14: 15 FROM R.G.SE. DOCUMENT CONTROL TO 913814152182 P. 85 Question X.Q. en r tion s NUREG-1037, pg 2-15, states Chat ~piping penetrations and aacoaiate8 piping for the six reference plants are not likely to contribute to containment leakage before reaching the capability graa~uras" ELecticaL penetration assenelies (aPA) have inboard and outboard seals which would( have to fail iu order to

'fai't. th~ penetration. MEN-1037, gg C-4 states that "Xf at least one set of the EE'A seals and/or sealants are at, oz'below the design 0mperature. then the potential for leakage is ~ected to be low."

Details of the piping penetrations and electrical penetration assedblies are discussed. below in items e aaO f. Since these penetrations are similar to those cesodJoed in MUREQ-1037, they are not expected to leak signiH.cantlY. Those penetrations having the greatest potential for Leakage include:

e ecyipuent hatch o personnel hatch e fuel, transfer tube

@Gage and vent system isolation valves a.'he 14'quipment Rat:cb is pressure un-seating with double tongue and groove silicone rubber seals. There are 36 swing bolts which are 1 3/4" diameter and have a specified torque of SQO to 3.000 ft-lb. The ecyxigment hatch is similar to the peach Bottom equipment hatch (12'iameter, 24 1-3/4 in. swing bolts with preload torque of 1900 ft-Xb) as described in NDIREG-1037, Appendix B, page 27. The NUREB calculates an upper hound lea3r. area (assumes no gasket> for the Peach Bottom equipment hatch of 4.15 square inches at 160 psig.

The 116" dimneter personnel hatches (2) are both pressure seating with double tongue and groove silicone seaIs. The ersonnel hatches are similar to the Zion hatch {122" ameter) shown in HUREG-XQ37, 'AppendQc H, Figure 10. 'Zhe HURSG calculates an upper bound leak area of 5.36 square inches at 134 gsie.~

C. The fuel transfer Cube is a 24" pipe sealed hy a double gas3ceted blind flange on the containment side, and by a gate valve on the spent Kuex pao1 side. 'X'he fuel transfer ne penetration. is (like'Zion and Surryj similar to the one shown in rig. 12 of Appendix B of NUREG-1037. The zUREG does not include calculations of the leakage for the fuel transfer tube hut cXues state l;hat "for a 1eak to occur between the PTT and its containment penetration sleeve, the leak must penetrate a berl3.owe'n the cantainneat side, the seal p1aC.e, AX a bellows on the outSiOe Of the contaixnnent." Xt is assumed. that this penetration vou1cL not 1eo1- si~fxcant1y,

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JUN-18-1997 14: 16 913814152182 P. 86 Z8 d,MM NOREG-z.037 states that "the large-diameter butterfly valves associated wit,h t: he purge aait vent, system are consiaerea to have the greatest potential for ccnta nt leakage~ and that "the main concern Ls tbdt Lh< ugn met,al3.fc seals between the valve body and disc wild. become degraded when subjected to the comb9.nneioa, of high pressures axe temperatures associated. rriTb severe accident comKtians -" at. 6Qxna, the 48'ontaizmeut purge ducts (supply and. mchauot) have double gasket flaugvs durmg normal operation. The mni-purge supply and exhaust

~3.vcr axo C~ XOHOX model 801 Vlimseal bueeeMly ml,vey with PBEE/metal back-up seats. This is nesidered a fire-safe seal in that the X'BBK aetcriol. 9.s the contact point during noxaml cqperations but iI they were to experience high temperatures that degraded the PRRK, there ia o nets hack up to eaixatain the seal. These valves MouM thus not he considered likely ta reeQt in significant lea3caga.

e The Ginna yipincj genetrations are generally eaibedded s3.eever except far the 3 drain lines free Sump 8 which are embedded.

yiye (2-S~, 1-4"). Theri ax'e 35-10~ an@ 3-8~ flanked sleeves or pipes. There are 8-8" and 3.3-10" Clued. her@/bellows penetrations. Tham az e 2-6~ end 3-X.O" flue'ead peuetxations. There M'e 2"24',/4"i 3.-14 1/4 andt 8-22 x/0" insulated. flued head/beXlows penetratfoas, There ie 1-24 a/4>>

insMated. Clued head/bellows penetration.

There are 50 einna electrical penetrations which were mannfactured by Cx'ouse-HincLs. The critical sealing function for these penetratious is ceramic to metal wbich, according to

Ãg33Ã/CR-3234, is an mccellent seal 8e89.cga.

The worst case temperature scenarios for Cixrna are the station blac}couts without power recovery or with power recovered, late.

Temperatures are seen between 300-375~ for 8-x0 hours. Worst case pressure sequences @re seen @hen there are ne fan coolers or containment sprays available concurrent with core-concrete interaction. '1'geese again tend ,to be the station blackout sequences.

penetratioa Based cn the above discussion, RGB failure is significantly less still belie~es that important than.

overpressure failure.

TOTAL PE 86