05000293/LER-1992-001, :on 921219,Class I Piping Seismic Damping Ratios Found Were Not Approved by NRC for Use.Caused by Personnel Error.Class I Piping Evaluated Using Higher Damping Ratios & Declared Operable
| ML20248L931 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 06/05/1998 |
| From: | Desmond N, Dicroce K BOSTON EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| BECO-2.98.074, LER-92-001, LER-92-1, NUDOCS 9806120069 | |
| Download: ML20248L931 (7) | |
| Event date: | |
|---|---|
| Report date: | |
| 2931992001R00 - NRC Website | |
text
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Boetas Edison i
Pilgrim Nuclear Power Station Rocky Hill Road Plymouth, Massachusetts 02360-5599 Nancy L Desmond June 5,1998 Regulatory Relations Group Manager BECo Ltr. 2.98.074 U.S. Nuclear Regulatory Commission Attn.: Document Control Desk Washington, D.C. 20555 Docket No. 50-293 License No. DPR-35 The enclosed supplemental Licensee Event Report (LER) 92-001-01, " Class I Piping Seismic Damping Ratios,"is submitted in accordance with 10 CFR Part 50.73. This supplement cancels the original voluntary LER.
Please do not hesitate to contact me if there are any questions regarding this report.
hk N.L. Desmond JH/deg 9200101 i
Ui Enclosure: LER 92-001-01 [l
Boston Edison Company
, y.V 9806120069 980605 PDR ADOCK 05000293 S
PDR yvvd
cc:
Mr. Hubert J. Miller Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Sr. NRC Resident Inspector - Pilgrim Station INPO Records 700 Galleria Parkway Atlanta, GA 30339-5957 Standard BECo LER Distribution m
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NRC Form 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 (4-95)
EXPIRES 04/30/98 LICENSEE EVENT REPORT (I'ER)
ISWAM BN)EN M RLWNSE TO LWM3 %RH THIS INK)RMATION COLHCTION REQUEST: 500 HRS. RLPORTED 7
LESSONS 11ARNED ARE INCORPORATED INTO THE UCENSING (See reverse for number of PROCI SS AP.O ILD BACK TO INDUS'lRY. FOllWARD COMMENTS digits / characters for each block)
REGARDING BURDEN LSTIMATE TO THE INFORMATION AND OECORDS MANAGEMLNT BRANCH (T-6 F33), U.S. NUC11.AR REGULATORY COMMIS$10N. WASHINGTON.DC 2055M101, AND TO THE PAPERWORK REDUCTION PROJECT Ol50 0104L OfTICE OF MANAGEMENT AND Bt!DGET. WAUHINGTON. DC 20501 FACILITY NAME (l)
DOCKET NUMBER (2)
PAGE(3)
PILGRIM NUCLEAR POWER STATION 05000-293 1 of 5 TrrLE H)
Class I Piping Seismic Damping Ratios EVENT DATE (5)
LER NUMBER (6)
REPORT DATE (7)
OTilER FACILITIES INVOLVFD (N)
SEQUENTIAL REVISION F ACILTTY NAME DOCKET NUMBER MOVI'H DAY YEAR YEAR NUMBLR NUMBER MONTH DAY DEAR FACILITY NAME DOCKET NUMBER 12 19 92 1992 001 01 06 05 98 N/A 05000 OPERATING Tills REPORT IS SUBMIITED PURSUANT TO Tile REQUIREMENTS OF 10 CFR:(Check one or more)(l!)
1% ODE (9)
N 20 2201 (b) 20 220 ka42W v) 50.73(a W 2 Wi)
I 50.73(aN2Hviii)
POWER 22 220Ma)(O 20 2203(aW3 Wi) 50 72aW 2Wii)(H) 50.73(aW2Wx)
LEVEL (10) 100 20.220ha)(2xi) 20.2203(ax3xii) 50.73(ax2xni) 73.71 20.2203(aW2 Wii>
20.220 haw 4) 50 7VaW 2Wiv)
X OTHER 20 2203(aW2 Hiid 50.3Nc W I) 50.73(a W 2 W y)
Specify in Abstract helow 20 220 haw 2Wiv) 50 36(cW2) 50.7N a W 2 W vii) or in NRC Form 366A LICENSEE CONTACT FOR Till" I ER (12)
NAME
~
TELLI' HONE NUMBER (include Ar a Cale)
Kristin R. DiCroce - Senior Regulatory Affairs Engineer 508-830-7667 COMPLETE ONE I INE FOR EACII COMPONENT FAILi'RE DFSCRIBED IN TIIIS REPORT (13)
REM)RTABli 10 REPORT ABLE 10
CAUSE
SYSTEM COMPONEPfr MANUFACTURER NPRDS
CAUSE
SYSTEM COMPONENT MANUFACTURER NPRDS SUPPLEMENTAL REPORT EXPECTED (14) l EXPECTED lMMlH DAY YEAR YES X
NO SUBMISSION fit yes. comme EX"ECTLD SUBMISSION DATE)
DATE(15)
ABSTkACT (Linut to 1400 spaces. i.e., approxirnately 15 wngle-spaced typewruten imes)(16)
On December 19, 1991, the Chss I piping seismic damping ratios permitted for use since 1982 were found to nCc conform with the damping ratios approved by the NRC for use at Pilgrim Station.
The damping ratios were increased in two changes to the Final Safety Analysis Report (FSAR).
The changes incorporated the damping ratios from NRC Regulatory Guide 1.61 and ASME Section III Code Case N-411.
NRC approval for the damping ratio changes was not sought under 10 CFR 50.59 because it was not recognired that the changes could constitute an unreviewed safety question.
This condition was discovered during a review of the main steam line/ safety relief valve tailpipe stress analysis..
The plant was operating with the reactor mode switch in the RUN position.
Reactor pressure was 1035 psig and reactor water temperature was approximately 548 degrees fahrenheit.
The use of higher damping ratios was determined not to be reportable in accordance with 10 CFR 50.73.
This report is submitted voluntarily.
The condition poses no threat to the public health and safety.
All piping systems designed or modified using seismic damping ratios greater than L, do NRC approved FSAR ratios are operable as determined in an operability evaluation
{j%7rovedonJanuary13, 1992.
Long term corrective actions are under development.
i 1Curr'ently, planned actions include identification of stress analyses where the higher damping ratios were used, development of enhanced design specifications for piping and supports, and systematic review of safety grade piping systems to demonstrate conformance with the appropriate licensing basis.
NRC FORM mo(++s)
4 WRC Forrn 366A U.S. NUCLEAR REGULATORY CON 1%11SSION (4 95)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUAT10N FACILITY NAME (1)
IX)CKET NUMBER LER NUMBER (6)
PAGE (3)
(2)
SEQUENTIAL REVISION YEAR NUMBER NUMHER PILGRIM NUCLEAR POWER STATION m 293 92 001 01 2 of 5 TEXT (if more space as required, use addinonal copes of NRC Form 366A)(17)
REASON FOR SUPPLEMENT This supplemental report is submitted to cancel the original voluntaw LER.
The original LER (92-001-00) identified a condition, which we now confirm not to be reportable involving the incorrect application of seismic damping factors.
Initial deportability conclusions were based on the misapplication of ASME Code Case N-411 and FSAR changes to damping ratios made with incorrect information.
l An engineering evaluation was completed in October 1993 to resolve LER 92-001-00 issues.
The conclusions of the evaluation were determined by the NRC to be inadequate to close the LER issues.
Additional analyses were performed in 1997 for the piping systems referenced in the LER using original FSAR damping requirements.
The results demonstrated the seismic capability of the affected piping met design basis requirements without modification.
EVENT DESCRIPTION
On December 19, 1991, the Class I piping seismic damping ratios permitted for use since 1982 were found not to conform with the damping ratios approved by the NRC for use at Pilgrim Station.
The damping ratios were increased from those approved in two changes to the Final Safety Analysis Report (FSAR).
NRC approval for the damping ratio cuanges was not sought under 10 CFR 50.59 because it was not gecognized that changes in damping ratios could constitute an unreviewed safety question.
Increased seismic damping concurrent with the plant's seismic response spectra could reduce the conservatism and the margin of safety as described in FSAR Section 12 and Appendix C.
These sections of the FSAR describe the analytical methods used and list the ratios as a percentage of critical damping.
The affected items are Class I piping and supports that provide structural and pressure integrity for normal, transient and accident conditions.
Systems currently identified to have been designed or modified using the higher damping ratios are the Reactor Water Level piping, the Recirculation System piping, the Main Steam relief valve discharge lines, the Scram Discharge Volumes, and the Residual Heat Removal system drain lines.
The FSAR changec were made by two 10 CFR 50.59 Safety Evaluations (SEs) which incorrectly determined that NRC approval was not necessary.
The first change was pursuant to SE 1421, approved May 24, 1982.
The second change was pursuant to SE 1697, approved August 23, 1984.
The damping ratios were changed as follows:
FSAR ORIGINAL SE 1421 SE 1697 QB_E E_S_5 OBE SS_g OBE AND SSE S
5 12" Dia
- 0. 5%
1.0%
1.0%
2.0%
Note 1 2 12" Dia 0.5%
1.0%
2.0%
3.0%
Note 1 l
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,U.S. NUCLEAR REGULATORY COMMISSION (4-95)
LICENSEE EVENT REPORT (LER) i TEX 1 CONTINUATION FACILITY NAME (1)
DOCKET NUMBER LER NUMBER (6) l' AGE (3)
(2)
SEQUENTIAL REVISION YEAR NUMBER NUMBER PILGRIM NUCLEAR POWER STATION m 293 92 001 01 3 of 5 TEXT (if rnore space is reqmred, use addihonal copies of NRC Foria 366A)(17)
Note 1:
For all piping sizes, 5% from 0 to 10 Hz, decreasing linearly to 2% at 20 Hz.
This is the ASME Code Case N-411 criteria.
Potential Condition Adverse to Quality (PCAQ) 92-3 and Management Corrective Action Request (MCAR) 92-1 were written to document the use of higher-than-licensed damping ratios.
Class I piping evaluated using the higher damping ratios was declared i
operable based on a NED operability evaluation.
This operability evaluation is consistent with the guidelines contained in NRC Generic Letter (GL) 91-18 and was reviewed and recommended for approval by the Operatios Review Committee (ORC) on January 13, 1992.
I Discovery of the condition occurred during a review of the main steam line/ safety relief valve tailpipe stress analysis.
The plant was operating with the reactor mode switch in the RUN position.
Reactor pressure was 1035 psig and reactor water temperature was approximately 548 degrees Fahrenheit.
Reactor Power Level was 100 percent
CAUSE
The cause of this condition was personnel error in judgment concerning the proper application of Reg. Guide 1.61 and ASME Code Case N-411.
The relationship between damping and response spectra was not clearly defined.
The determinations made on the use of higher damping in SE 1421 and SE 1697 reflect the belief that since the technical issues had been resolved, NRC approval was not necessary.
The key question in both evaluations was whether the changes decreased the margin of safety as described in the basis of Technical Specifications.
The authors and reviewers of both safety evaluations did not recognize the change could result in a decrease in the margin of safety.
SE 1421 justified increasing damping ratios to those contained in NRC Regulatory Guide (Reg. Guide) 1.61, " Damping Values for Seismic Design of Nuclear Power Plants".
The higher damping ratios from Reg. Guide 1.61 were apparently meant to be used with the response spectra contained in Reg. Guide 1.60, " Design Response Spectra for Seismic Design of Nuclear Power Plants."
The Standard Review Plant (SRP) in 1975 defined an acceptable design basis for a commercial nuclear power gsneration facility.
The SRP Section 3.7.1 recommends, among other things, the use of Reg. Guide 1.60 for the development of ground response spectra, and Reg. Guide 1.61 for damping ratios.
The requirements to use Reg. Guides 1.60 and 1.61 together was not clear.
Furthermore, Pilgrim Station was bui.'t prior to the issuance of the SRP.
Used I
together, the Reg. Guides would be a more advanced analytical technique, which would not have reduced the margin of safety.
However, Reg. Guide 1.61 used with the original plant design spectra was less conservative.
i i
NRC IURM 366A (4-95)
.U.S. NUCLEAR REGULATORY COMMISSION (4-93)
LICENSEE EVENT REPORT (LER)
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TEXT CONTINUATION FACILITkNAME (1)
DOCKET NUMBER LER NUMBER (6) l'AGIb (2p SEQUENTIAL REVISION YEAR NUMBER NUMHER PILGRIM NUCLEAR POWER STATION 05000-293 92 001 01 4 of 5 i
TEXT (if more space is required, use addinonal copies of NRC Form 366A)(17)
The use of the higher damping ratios with the original plant design spectra could result in a decrease of the margin of safety as described in the basis of Technical Specifications.
SE 1421 addressed the technical aspects only.
]
The lack of recognition of the usage requirements of Reg. Guide 1.61 resulted in NRC i
approval not being sought as required by 10 CFR 50.59 and 10 CFR 50.90 ASME Code Case N-411 adoption was justified by SE 1657 on August 23, 1984.
NRC approval for use of ASME Section III Code Cases is delineated in Reg. Guide 1.84,
" Design and Fabrication Code Case Acceptability ;SME Section III Division 1."
Revision 22 of Reg. Guide 1.84 was effective in July 1984.
Code Case N-411 is not listed in Revision 22.
Sectioc C of that revision states that " code cases not listed herein are either not endorsed or will require supplementary provisions on an individual basis to attain endorsement status."
Code Case N-411 was subsequently endorsed in Revision 24 of Reg. Guide 1.84 in June of 1986.
Endorsement was limited to those plants that used the " current" seismic spectra.
" Current" is interpreted to mean the Reg. Guide 1.60 response spectra.
Pilgrim's use of its original spectra l
may not qualify for use of N-411 under the NRC endorsement.
The adoption of Code
{
Case N-411 without specific NRC approval resulted from a misunderstanding of the NRC endorsement requirement.
CORRECTIVE ACTION
Class I piping evaluated using the higher damping ratios was declared operable based on a NED operability evaluation.
This operability evaluation was reviewed and recommended for approval by ORC on January 13, 1992.
Currently, planned actions include identification of stress analyses where the higher damping ratios were used, resolution of discrepancies between the existing design and the regulatory guidelines, development of enhanced design specifications for piping and supports, and systematic review of safety grade piping systems to demonstrate conformance with the appropriate specification.
The need for further review of older safety evaluations will be determined under MCAR 92-1.
SAFETY CONSEQUENCES
The use of the higher seismic damping ratios for Class I piping poses no threat to the public health and safety.
I All piping systems that have been designed or modified using seismic damping ratios f
greater that the NRC approved FSAR ratics are operable as defined in Technical l
Specifications.
Consistent with NRC Generic Letter 91-18, the allowable stress limits used to determine operability of the affected piping and related supports at PNPS are based on the requirements of IE Bulletins 79-14, 79-02, and ASME III Appendix F.i N.
aU.S. NUCLEAR REGULATORY CO31511SSION (4-93)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
IMX'KET NUMBER LER NUMBER (6)
PAGE (3)
(2)
SEQUENTIAL REVISION YEAR NUMBER NUMBER PILGRIM NUCLEAR POWER STATION 05000-293 92 001 01 5 of 5 TEXT (if more space is required, use additional cognes of NRC Form 366A)(17)
There are no special compensatory measures required for continued operation.
All the systems in question meet ASME code required safety margins but operation in this condition 1; limited to one operating cycle.
This LER is submitted voluntarily.
SIMILARITY TO PREVIOUS EVENTS A review was conducted for similarity to previous Pilgrim Station Licensee Event Reports tLERs) written since January 1984.
The review focused on LERs submitted on a voluntary basis or pursuant to 10 CFR 50.73 subparts (a) (2 ) (ii), (a) (2) (v), and (a) (2) (vii).
No events were identified that involved personnel error in performance of safety evaluations.
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l l
ENERGY INDUSTRY IDENTIFICATION SYSTEM (EIIS) CODES
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l The EIIS codes for this report are as foll'ows:
COMPONENTS CODES
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l Supports SPT Pipe PSP I
i SYSTEMS-Control Rod Drive System (Scam Discharge Volume)
AA Main Steam System SD Primary Containment System (PCS)
JM Reactor Water Level System JD Reactor Recirculation System AD Residual Heat Removal BO l
I l
l l'
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