05000219/LER-1997-001, :on 970103,seven Penetrations in Five Sys Did Not Meet Requirements as Described in GL 96-06.Caused by Previous Analyses Were Performed Without Using Conservation Assumptions.Determination Was Performed
| ML20134E772 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 01/31/1997 |
| From: | Roche M, Tamburro P GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 6730-97-2043, GL-96-06, GL-96-6, LER-97-001, LER-97-1, NUDOCS 9702070088 | |
| Download: ML20134E772 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(viii) 10 CFR 50.73(a)(2)(ii) |
| 2191997001R00 - NRC Website | |
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l GPU Nuclear, Inc.
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U.S. Route #9 South NUCLEAR Post Office Box 388 Forked River, NJ 08731-0388 Tel 609-9714000 January 31,1997 6730-97-2043 U. S. Nuclear Regulatory Commission Attn.: Document Control Desk Washington, DC 20555
Dear Sir:
Subject:
Oyster Creek Nuclear Gener, ting Station Docket No. 50-219 Licensee Event Report 97-01; Seven Drywell Penetrations Do Not Meet the Requirements described in Generic Letter 9646 Enclosed is Licencee Event Report 97-01. This event did not impact the health and safety of i
the public.
If any additional information or assistance is required, please contact Mr. John Rogers of my staff at 609.971.4893.
l Very truly yours,
\\
Michael B. Roche Vice President and Director Oyster Creek 9702070008 970131 PDR ADOCK 05000219 S
PDR f
MBR/JJR 69)i Enclosure cc:
Oyster Creek NRC Project Manager Administrator, Region i Senior Resident Inspector
e NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION 14 95)
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FACIUTV NAWlE (1)
DOCLET NUMBER (2)
PAGE(3s Oyster Creek Unit 1 05000 - 219 1 of 4 TITL[ 44)
Seven Drywell Penetrations Do Not Meet the Requirements Described in Generic letter %4 EVtmi DATE (5)
LER NUMBER (6)
REPORT DATE (7) l OTHER rACILITIES INVOLVED (8)
MONTH DAY YEAR YEAR SEQUENTIAL REVISION MDhTH DAV YEAR l FACILITY hAME DOCAET NUMBER l
NUMBER NUMBER 05000 01 03 97 97 --
01
-- 00 01 31 97 l ' ^""" "^"'
l 05000 OPERATING N
THIS REPORT 15 SUBhin itu PUR!,UANT TO THE R EQUIREMENTS OF 10 CFR 6: (Check one or more) (11)
MODE (9) 20.2201(b) 20.2203(aH2Hv) 50.73(aH2)(i) 50.73(a)(2)(viii)
POWER
}QQ 20.2203(aH1) 20.22031aH3Hi)
X so.73(aH2Hui 50.73(a)(2Hx)
LEVEL (10) 20.2203(aH2)(i) 20.2203(a)(3Hu) 50.73(aH2Hm) 73.71 20.2203(a)(2Hu) 20.2203(aH4) 50.73(aH2Hav)
OTHER 20.2203(aH2Hiid 50.36(c)o )
50.73(aH2Hv) 20.2203(aH2)(iv) 50.35(cH2) 50.73(aH2)(vii)
LICENSEE CONTACT FOR THIS LER (12)
NAML TELEPHONE NUMBER (Irwiude A'en Codel Peter Tamburro 609.971.4141 COMPLETE ONE LNE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)
EAUSE SYSTEM 00MPONENT MANUIACTURER MtPUNI ABLE LAUSE SYSTEM COMPONENT MANUFAG!URER REP 0HIABLE TO NPRDS TO NPRDS UUPPLEMENTAL REPORT EXPECTED (14)
EXP EC TED MCNTH DAt YEAR SUBMISSION YES X
NO (if yes complete EXPECTED SUBMISSION DATE).
ABUTRACT (Limit to 1400 spaces. i.e., approximately 15 single-spaced typewntten lines) 116)
On January 3,1997, during a review as requested by Generic letter (GL) %4, " Assurance of Equipment Operability and Containment Integrity During Design Basis Accident Conditions",
seven penetrations in five systems did not meet the requirements as described in GL 964. Under the postulated conditions contained in GL %4, these seven penetrations are only required for containment integrity. An analysis was performed of each penetration while isolated. The analysis modeled the effects on internal fluid and piping in response to an external ambient temperature increase. The results revealed that although the piping did not meet the design requirements, the postulated pressures did not exceed ASME Section 111, Appendix F criteria for piping.
Additionally, the potential effects on the respective isolation valves were considered. A catastrophic failure of the valves is not considered credible. Therefore containment integrity for the penetration was maintained and the safety significance of this discovery is considered minimal.
The cause for this condition was that previous analyses were performed without using the more conservative assumptions described in GL 964. Operability determinations were performed and further evaluations are ongoing to determine the need for modifications or procedural revisions.
NHC FORM 366 (4-95)
NRC F.ORM 366A U.S. NUCLEAR REIULATORY COMM4SION (4 95) l UCENSEE EVENT REPORT (LE*J i
TEXT CONTINUATION FACILITY NAME (1)
DOCKET (2)
LER NUMBER (6)
PAGE (3) 05000 YEAR SEQU AL REV Oyster Creek, Unit 1
- - 219 97 --
01 00 2 of 4 1
TEXT fit more space is required. Use additamalcopies of NRC form 366A) (11)
DATE OF DISCOVERY l
The conditions described in this report were discovered on January 3,1997.
IDENTIFICATION OF DISCOVERY During a review requested by Generic Ixtter (GL) 96-06, " Assurance of Equipment Operability and Containment Integrity During Design Basis Accident Conditions", seven penetrations (EllC -
PEN) in five systems were evaluated as not meeting the requirements presented in GL %-06. The systems are Shutdown Cooling (SDC)(Ells - KE), Reactor Building Closed Cooling Water (RBCCW)(Ells - KG), Reactor Water Cleanup (RWCU)(Ells - KH), Isolation Condenser (IC)(Ells - BL), and Reactor Recirculation Loop Sampling (RLS)(Ells - KN). This discovery is conside red reportable under 10 CFR 50.73(a)(2)(ii).
t i
CONDITIONS PRIOR TO DISCOVERY The reactor was operating at approximately 100% power. At the time of discovery, system pressures and temperatures were normal for full power operation. However, the plant had been operated in all modes with these conditions since original plant startup.
DESCRIPTION OF DISCOVERY GL 96-06 requested licensees to consider the possibility of equipment damage following a design basis accident caused by the heating of trapped internal fluids. During the review of affected plant systems, it was determined that seven penetrations did not meet the design stress requirements.
The areas of concern are: 1) the supply and return penetrations for the SDC system; and 2) the RBCCW return line penetration. Additionally, the following penetrations become concerns if they were previously isolated for maintenance: 1) The RWCU supply line penetration; 2) both IC return line penetrations; and 3) the RLS penetration.
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e NRC F,ORM 366A U.S. NUCLEAR REGULATORY COMMISSION UCENSEE EVENT REPORT (LER)
+
TEXT CONTINUATION
~
FACILITY NAME (1) 0,0CKET (2)
LER NUMBER (6)
PAGE (3) 05000 YEAR SE U AL FiEV Oyster Creek, Unit i
- - 219 97 -
01 00 3 of 4
~ TEA t til resore space is required, use additional copies of NRC Forrn 366Al (17)
DESCRIITION OF DISCOVERY (Cont.)
The review of the isolated piping sections revealed that following a design basis less of Coolant Accident (LOCA) or Main Steamline Break accident (MSLB), heating of the fluid trapped in the sections would cause thermal expansion. This expansion increases the internal pressure of the fluid in the piping, and results in the seven identified piping sections exceeding ASA/ ANSI B31.1 Code allowable stresses.
APPARENT CAUSE OF TIIE OCCURRENCE l
l The root cause of the condition is that previous analyses were performed without using the more conservative assumptions described in GL 9606.
ANALYSIS OF DISCOVERY AND SAFETY ASSESSMENT During a review requested by Generic IEtter (GL) 96-06, seven penetrations in five systems were evaluated as not meeting the requirements presented in GL %-06. The five systems are nat required to operate to mitigate uOCA or MSLB accidents. Therefore, the sole function of these seven penetrations is for containment integrity.
A review utilizing the guidance provided in GL 91-18,"Information to Licensees Regarding Two NRC Inspection Manual Sections on Resolution of Degraded and Nonconforming Conditions and on Operability", was conducted to evaluate the operability of these seven penetrations. The review assumed that each penetration was isolated. The analysis modeled the effects on internal fluid and piping in response to an external ambient temperature increase. Actual configuration and material properties were utilized. The results revealed that the piping did not meet the design requirements.
However, even when the most conservative assumptions were included, the postulated pressures did not exceed ASME Section 111. Appendix F criteria for piping. Therefore containment integrity for the piping was maintained.
An analysis of the isolation valves (EllC - IV) was then performed. The SDC, RWCU, IC, and RLS valves are all either 600 psi class or 900 psi class valves. No credit was taken for seat leakage. Ilowever, it was determined that valve bonnet or packing leakage was likely. This would l
relieve the pressure in the penetration. Based on n review of the design, a catastrophic failure of l
any of these valves is not considered credible.
iI
NRC F,ORM 366A U.S. NUCLEAR REOULATO7.Y COMMISSION (4-95)
UCENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME 11)
DOCKET (2)
LER NUMBER (6)
PAGE (3) 05000 YEAR SE U AL REV Oyster Creek, Unit 1
- - 219 97 -
01 00 4 of 4 iEKT (11 more space is required. use additionalcopies of NRC Form 366A) (11)
ANALYSIS OF DISCOVERY AND SAFETY ASSESSMENT (Cont.)
4 The RBCCW isolation valves were evaluated. These are six inch 150 psi class gate valves with bolted bonnets and solid disks. Discussion and correspondence with the valve vendor indicated that the valves leak past their seats when overpressurized on the upstream side. Significant overpressurization upstream of the valve causes pressurization of the valve bonnet, which results in
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a curvature of the 11at txx!y neck. Resultingly, this forces the body seat out of plane and allows the valve to leak. Once the leakage has reduced pressure, the forces on the valve neck reduce and the valve would rescat. Therefore, catastrophic failure of these valves in not considered credible.
Based on the fact that: 1) these five systems are not required to operate to mitigate a LOCA or l
MSLB; 2) the penetration pressures do not exceed ASME Section III, Appendix F criteria; and 1
- 3) the affected valves will not fail catastrophically, the safety significance of these conce.rns are minimal.
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CORRECTIVE ACTIONS
Immediate Actions Taken Upon discovery of this concern, an operability determination was performed which determined that the isolated penetrations would maintain containment integrity.
Short Term Compensatory Actions Two SDC valves outside the containment were opened to reduce the peak pressure of the penetration during the LOCA or MSLB.
L(mg Term Corrective Actions The seven penetrations will be further evaluated to determine if modifications or procedural revisions are required.
SIMILAR EVENTS
LER 96415; Reactor Water Cleanup Valves May Not Operate During a Line Break Due to a Non Conservative Analysis LER 95-005; Non Conservative Anticipatory Scram Bypass Switch Deficiency due to Original Plant Designi l
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