ML20198J898
| ML20198J898 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 10/17/1997 |
| From: | Ewing E ENTERGY OPERATIONS, INC. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-96-06, GL-96-6, W3F1-97-0232, W3F1-97-232, NUDOCS 9710220200 | |
| Download: ML20198J898 (8) | |
Text
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6 En e y perations.Inc.
,O KHlona. LA 70066 Tel 504 739 6M Early C. Ewing, til a ety & Regulatory Mairs W3F1-97-0232 A4.05 PR K
C October 1 r', ';C97 i
U.S. Nuclear Regulatory Commission ATTN: Document Control Desk
}
Washington, D.C. 20555 subject:
Waterford 3 SES Docket No. 50-382 lg~
c License No. NPF-38 Supplemental Response to NRC Generic Letter 96-06 L
Assurance Of Equipment Operability and Containment Integrity During Design Basis Accident Conditions Gentlemea:
The purpose of this letter is to provide a supplement to Waterford 3's original response to Generic Lettet (GL) 96-03 which was transmitted in Letter W3F1 0017, dated January 28,1997. The generic ;etter identified safety significant issues which could affect containmont integrity and equipment operability during certain accident conditions. One action with respect to referenced scenarios within the generic letter was to determine if rsing systems that penetrate containment are susceptible io thermal expansion of fluid so that overpressurizatior' of piping could h:
occur.
Identified in Attachmont B of the referenced response, are 12 containment penetrations which are potentially susceptible to overpressurization. Subsequent to j jI transmittal of the Waarford 3 GL 96-06 response and as a result of examination of 1
plant operaGng ' parameters following Refuel Outage 8, another potentially susceptible penetration (Letdown Line from the Regenerative Heat Exchanger) hns bean identified, bringiag the total to 13 containment penetrations which are potentially susceptible to overpressurization. This penetration was originally i screened out based on the minimum temperature downstream of the Regenerative Heat Exchanger. To update the origina: submitta!, a revised Attschment B which is
- annotated with revision bars is enclosed, u
9710220200 M1017:*-
PDR ADOCK 'JOOO382 '
P.
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Supplemental.'lesponse to NRC Generic Letter 96-06 Assurance of Equipment Operability and Containment Integrity During Design Basis Accident Con 6itions W3F197-0232 Page 2 October 17,1997 Please note that the addition of the Letdown Line from the Regenerative Heat Exchanger Penetration did not affect the conclusicns of the initial evaluation which stated that potentially susceptible containment penetrations do not exceed Burct Pressure and do retain their ability to perform their safety function, thereby maintaining containment integrity.
Please contact me at (504) 739-6242 or Tim Gaudet (504) 739-6666 should any nuestions arise.
Very truly yours, r
[
E.C. Ewing Diret, tor Nuclear Safe.ty and Regulatory Affairs ECE/ PRS /ssf Attachment cc:
E. Merschoff (NRC Region IV)
C.P. Patel (NRC-NRR)
J. Smith N.S. Reynolds NRC Resident inspectors Office i
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_1 UNITED STATES OF AMERICA y-
- NUCLEAR REGULATORY COMMISSION iln 'the matter of ~
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i' Entergy Operations, Incorporated
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Docket No. 50-382
- Waterford 3 Steam Electric Station -
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4 Early Cunningham Ewing, being duly sworn, i ereby deposes and says that he is i
Director, Nuclear Safety & ReCulatory Affairs Waterford 3 of Entergy Operations, Incorporated; that he is duly authorized to sign and file with the Nuclear Regulatory Commission the attached Supplemental Information Regarding NRC Generic Letter 96-06; that he is familiar with the content thereof; and that the matters set forth therein are true and correct to the best of his knowledge, information and belief.
h Earp) Cunnindiam EwinP Director, Nuclear Safety & Regulatory Affairs
- STATE OF LOUISIANA
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) ss PARISH OF ST.' CHARLES
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Subscribed and swom to before me, a Notary Public in and for the Parish and State
- above named this E 2 4 day of (C L d
,1997.
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he Notary Public LMy Commission expires = dAT ch
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Attachm :nt to W3F1-07-0232
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Page 1 of 5 ATTACHMENT B Evaluation for Susceptibility of Containment Penetration Piping Overpressurization Due to Thermal Expansion of Fluid An evaluation was performed of the Waterford 3 containment mechanical penetrations utilizing the following criteria developed for all Entergy sites:
Entergy Generic Letter 96-06 Penetration Evaluation Criteria l
1)
Scope: All containment (Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR)) and drywell BWR penetrations.
2)
A penetration piping system, including any connected heat exchangers, wm be considered to be Potentially Susceptible if it meets a_Il of the following four criteria:
A.)
The penetration must be full of liquid at the time of the accident.
Pipes containing air, gas or steam win be excluded.
B.)
The liquid contained in the penetration piping must be at a lower temperature than the surrounding environment during operational or accident situations. Piping that contains water at or near Reactor Pressure Vewel (RPV) or Steam Generator (SG) temperatures, such as feedwater, letdown, blowdown or Reactor Water Cleanup (RWCU), would actually have initial fluid temperatures higher tnan those expected during an accident.
C.)
The penetration must be isolated during an event, i.e., plant heatup l o, accident, that could cause a significant heat transfer to the fluid between the isolation valves. The valve arrangement used for penetration isolation must restrict flow out both directions, if the inboard isolation valve is a check valve or certain type and orier.tation of solenoid valve, (with a mechanism of pressurs reliel in the connecting piping) the penetration may possibly be excluded.
This exclusion would also include piping open to the suppression pool, SG, RPV, or containment air space.
Att: chm:nt to j
Page 2 of 5 In order to be excluded, the extended piping system available for fluid expansion inside containment must not constitute a closed systam, so that the fluid volum3 can expand and prevent damage to the containment isolation portion of the piping penetration.
Addi*ionally, another closed valve further down the line inside containment must not prevent expansion of the fluid volume in the penetration, thereby isolating a penetration with en expected available leak path I.e., check valve.
D.)
The potentially susceptible penetration will not have any pressure relief valves (with sufficient capacity and setpoint) or other method of overpressure protection (such as a check valve in parallel with the main inboard valve) between the isolation valves.
A penetration will additionally be considered potentially susceptible if it meets any one of the fo!!owing two criteria:
The penetration will be considered potentially susceptible if a single-failure coupled with an accident would cause isolation, heatup and overpressurization, of a normally open, low temperature, fluid filled penetration.
A penetration will be considered potentially susceptible if trapped pressure can prevent safety-related isolation valves from opening when required to mitigate an accident, i.e., pressure locking. Ref.:
A Potentially Susceptible penetration may be eliminated from concern if qualified calculations or analyses demonstrate that the penetration piping system, which includes the valves, remains within its Design Basis.
Penetrations that do not meet their Design Basis requirements shall be considered Susceptible, and have a basis for operability established in accordance with Generic Letter 91-18 guidance.
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Attechm2nt to W3F1-97-0232 Page 3 of 5 At Waterford 3, seventy-four (74) containment penetrations were evaluated per the preceding criteria. Of these 74 containment penetrations, the following 13 l
were considered potentially susceptible:
Per%trat. ion #
System Fluid Dwg 7
Pnmary Makeup Water G-161 Sh 2 24 CCW Return From RCP's Water G-160 Sh 4 26 Letdown Line from Regenerative HX Water G-168 Sh 1 l
28 RCS Sampling Water G-162 Sh 2 29 PZR Surge Line Sampling Water G-162 Sh 2 30 PZR Steam Space Sampling Water G-162 Sh 2 42 Containment Sump Pump Discharge Water G 173 Sh 3 43 Reactor Drain Tank Outlet Water G-171 Sh 1 44 RCP Controlled Bleedoff Water G-168 Sh 2 51 Refueling Water Supply to Refueling Cavity Water G-163 59 SIT Fill / Drain From RWSP Water G-167 Sh 4 62 Refueling Cavity Drain Pump to RWSP Water G-163 71 Demineralized Water Water G-161 Sh 2 Penetration # 26 was originally screened out cased on an assumed average temperature in the penetration of 261*F. This penetration has a unique configuration with the Regenerative Heat Exchanger (RHX) located between the two containment isolation valves. When RCS Letdown at 543*F enters the RHX and exits at 200*F, the average temperature in the penetration is 261*F. This assumption resulted in an administrative limit on RHX outlet temperature of 200*F. This 200*F limit proved too restrictive for normal operation. The previously performed detailed engineering analysis had evaluated penetration
- 26, assuming an average temperature of 200*F in the penetration instead of 261*F. No credit was taken for this analysis in the original submittal. Lowering the average temperature from 261*F to 200*F allows the operationallimit on RHX outlet temperature to be lowered to 130*F. The 200*F average temperature is below our 260*F screenq criteria; therefore, Penetration #26 requires inclusion as a GL 96-06 susceptible penetration.
l Analysis determined that all 13 potentially susceptible penetrations exceeded l
their normal code allowables and would require a basis for operability established in accordance with the guidance contained in Generic Letter 91-18.
In accordance with GL 91-18, a penetration is considered operable if analysis demonstrates that the penetration piping system maintains its ability to perform the safety function it was designed to perform.
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Att chm::nt to W3F1-97-0232 Page 4 of 5 A calculational analysis of the 13 containment penetrations has been prepared.
l This calculation determines the maxirnum internal pipe pressure due to thermal expansion of the water, and considers the expansion of piping due to pressure from the encased water and the increase in p;pe temperature. The change in pipe volume will result from the elastic and plastic deformation of the pipe due mostly to the internal pressure, and partly due to the temperature rise. No consideration was given for leakage past the isolaticn valve seats. This assumption is conservative and results in an internal pipe pressure greater than actual values.
The resulting internal pipe pressure was then compwed to the burst pressure of the pipe. The calculation concludes that none of the 13 penetrations exceed l
burst pressure.
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Att:chment to W3F1-97-0232 Page 5 of 5 Conclusion None of the 13 penetrations exceed burst pressure The piping will yield and l
deform to relieve the pressure, but the piping will not fail. Since the safety function of the penetrations is not compromir.ed, a basis for operability in accordance with GL 91-18 is established. This conclusion is further supported by available Local Leak Rate Testing (LLRT) leakage data; valve leakage would decrease the internal pressure and provide additional margin for operability.
These 13 penetrations were reviewed for effects caused by pressure locking or l
thermal binding in accordance with GL 95-07. None of the valves required to open post accident are susceptible to pressure locking or thermal binding.
Reportability Determination in performing the review requested by Generic Letter 96-06, Waterford 3 noted several containment penetrations that were potentially susceptible to overpressurization. While detailed engineering evaluations demonstrated that the postulated overpressurization of these penetrations would not jeopardize the ability of the penetrations to perform their safety functions (i.e., containment isolation); it was determined that in some instances that ASME Ill Class 2 code limita could be exceeded. Waterford 3 considers this to be a nonconforming condition as described in Generic Letter 91-18 and thus addressed the issue accordingly, in accordance with Generic Letter 91-18, the nonconforming condition was documented within the corrective action program and a prompt determination of operability determined and documented for each affected penetration. Following assurance that conminment integrity would be mab.tained (i.e., no safety concern exists), the nonconforming condition was evalubted for potential reportability requirements. In this instance the reportability determ5ation was contugent on the interpretation of the phrase "outside the design basis".
Waterford 3 be 'ieves guidance provided for making this interpretation directs one to focus on prr ervation of defense in-depth particularly as it relates to protection of fission barriers.
In inis case since all affected containment penetrations retain their ability to perform their safety function and thus containment integrity is maintained, this nonconforming condition is not considered reportable under the criteria of 10CFR 50.7.2/50.73. As with other nonconforming conditions, appropriate corrective action to restore the condition to within the required quality requirements will be taken in accordance with the safety significance of the issue.
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