05000265/LER-1988-006, Errata to LER 88-006-02:on 880404,station Notified That Eleven Flued Head Anchors Did Not Meet Design Requirements. Caused by Misinterpretation of Scope & Design Structures.Mod Initiated to Revise Structure

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Errata to LER 88-006-02:on 880404,station Notified That Eleven Flued Head Anchors Did Not Meet Design Requirements. Caused by Misinterpretation of Scope & Design Structures.Mod Initiated to Revise Structure
ML20101K134
Person / Time
Site: Quad Cities Constellation icon.png
Issue date: 06/04/1992
From: Bax R, Kunzmann D
COMMONWEALTH EDISON CO.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20101K122 List:
References
LER-88-006-01, LER-88-6-1, RLB-92-099, RLB-92-99, NUDOCS 9207020099
Download: ML20101K134 (7)
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LER-2088-006, Errata to LER 88-006-02:on 880404,station Notified That Eleven Flued Head Anchors Did Not Meet Design Requirements. Caused by Misinterpretation of Scope & Design Structures.Mod Initiated to Revise Structure
Event date:
Report date:
2652088006R00 - NRC Website
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RLB-92-099 June 4, 1992 U.S. Nuclear Reguletory Commission Document Control Desk Washington, DC 20555

Reference:

Quad-Ctile! Nuclear Power Station Docket Number 50-265, DPR-30 Unit Two Enclosed is Licensee Event Report (LER)88-006, Revision 02, for Quad-Cities Nuclear _ Power Station. This revision provides additional information regarding flued head anchors.

This report is submitted in accordance with the requirements of the Code of Federal Regulations, Title 10, Part bO.73(a)(2)(11)(B), which requires the reporting of any event or condition that resulted in the nuclear power plant being in a condition that was outside the design basis of the plant.

Respectfully, COMMONHEALTH EDISON COMPANY '

QUAD-CITIES NUCLEAR POWER STATION

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THis REPORT is SUBMIT 1ED PUR$UANT TO THE REQUIREMENTS OF 10CFR 1(htd one or mRtt_DLiht_iQllD*_if01 (111 4 20.402(b) ___ 20.405(c) _' 50.73(a)(2)(iv) _ 73.71(b)

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On April 4, 1988, Quad-Cities Unit Two was in the RUN mode at 93 percent thermal power. At 1410 hours0.0163 days <br />0.392 hours <br />0.00233 weeks <br />5.36505e-4 months <br />, the Station wLs notified by d e BWR Engineering Department that eleven flued head anchors did not meet the at: bn rep (rements of the Final '

Safety-Analysis Report (FSAR). _ NRC notification of__this condition was completed at 1423 hours0.0165 days <br />0.395 hours <br />0.00235 weeks <br />5.414515e-4 months <br /> to satisfy 10 CFR 50.72. A subsequent inspection found that there was also a pin missing from one of the affected flued head anchors. The missing pin was avaluated and the flued head anchor was considered operable.

.The cause f_or this condition was due to misinterpretation of scope in that these strut.tures were not reassessed for design base requirements based on IE Bulletin 79-02 and 79-14 programs. The cause for the missing piii could not be determined.

Modification 04-02-88-0;7 has been initiated to revise the structures to comply with FSAR requirements. A program is in place to analy2e the Unit One structures in a similar manner.

This report is provided to comply with the requirements of 10 CFR 50.73(a)(2)(11)(B).

DVR-107 I

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_ Qual Cliitt unit Two 0 i 5_LA_Lo_Lo_L2161- s s18 -

o LoJ_6_. _ o12- alz or sit TExV PLANT AND SYSTEM IDERIlflCATIO11 General Electric - Bolling Water Reactor - 2511 MHt rated core thermal power. Energy fndustry Identification System (EIIS) codes are identified in the text as [XX).

EVENT IDENTIFICATION:

Eleven Unit Two flued head anchors do not meet the design requirements due to analysis deficiency. ,

A. CONDITIONS PRIOR TO EVIRIl Unit: Two Event Date: April 4, 1988 Event Time: 1410 '

Reactor Mode; 4 Mode Name: RUN Power Level: 93% .

This report was initiated by Deviation Report D-4-2-88-017

RUN Mode (4) - In this position the reactor system pressure is at or above 825 psig, l and the reactor protection system is energized, with APRM protection and RBM I interlocks in service (excluding the 15% high flux scram). ..

B. DESCRIPTION OF EVENT:

On Apr11-4, 1988, Quad-Cities Unit Two was in the RUN mode at 93 per cent of rated

, core thermal power. At 1410 hours0.0163 days <br />0.392 hours <br />0.00233 weeks <br />5.36505e-4 months <br />, the Station was notified by the Bolling Water Reactor Engineering Department (BHRED) that eleven (11) flued head anchors [SPT] did not meet the design requirements specified in the Ouad-Cities Final Safety Analysis l Report (FSAR). All of the Unit Two flued head anchers were rev,lewed and the ones j in question are located at the following penetrations:

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1. X.11 High Pressure Coolant Injection [B)) Steam Supply
2. X-13A Residual Heat Removal (RHR)/ Low Pressure Coolant Injection (LPCI) injection [BO)
3. X-13B RHR/LPCI Injection
4. X-16A Core Spray [BM) Injection S. X-16B Core Spray Injection
6. X-23 Reactor Building Closed Cooling Water (RBCCH)

(CC) Supply

7. X-24 RBCCH Return
8. X-36 Control Rod Drive (CRD) [AA) Return
9. X-47 . Standby Liquid Control [BR) Injection
10. X-8 Main Steamline Drains [SB)
  • 11. X-7A, B, C, D Hain Steam X-9A, B Feedwater [SJ)

X-10 Reactor Core Isolation Cooling [BN) Steam Supply X-12 Shutdown Cooling Suction [BO)

X-17 RHR Head Spray [BO)

' Gang Anchor The NRC, via the Emergency Notification System (ENS), was notified of this condition at 1423 hours0.0165 days <br />0.395 hours <br />0.00235 weeks <br />5.414515e-4 months <br />, to satisfy the requirements of 10 CFR 50.72. The design concern for the eleven (11) flued head assemblies was the result of a concern identified at Dresden Station. It was identified duri.19 the Dresden review that the flued head anchor structures at Dresden and Quad-Cities were not included under the I.E.

Bulletins No. 79-14 and 79-02 scope of work.

The flued head anchors in question were assessed for considerat' ion of continued

. operability. The results of the assessment concluded that the flued head anchors aill perform their intended functions, thereby, establishing an acceptable operability basis.

Subsequent to the initial submittal of this LER, it was discovered that one shear pin was missing.from the Flued Head Anchor structure at penetration X-16A. Thii specific concern was assessed by Boiling Hater Reactor Engineering Department tnd the anchor was considered operable without the pin. It was recommended, however, to install a pin in order to fulfill the FSAR reautrements.

DVR 107

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_Dantiille13 bill s .0_L11LLLlJL12LtLLeLL._ JLL9_LL - o_ ILc1L_0Lolt TEXT C. AEEAREtLLCAUSE OF EVI!R1 This report is submitted to comply with the requirements of 10 CFR 50.73(a)(2)(11)(B), which requires the reporting of any event or condition that resulted in the nuclear power plant being in a condition that was outside the design basis of the plant.

The exclusion of the structures in the 79-02 and 79-14 programs was due to misinterpretation of the scope requirements. Therefore, analysis for these ~

structures was not reassessed for design base requirements. Although the eleven (ll)-flued head anchors did not meet FSAR design requirements, these flued head anchors were considered operable as determined by an analysis in January 1988.

The cause of the missing pin for the structure at penetration X-16A could not be determined.

3 D. SAFETY ANALYSIS _OF EVENT:

Os .iealth and safety of the public and of plant personnel was not adversely affected by this event. Since the anchor assemblies were analyzed and considered operable, the associated safety significance is minimal.

The concerns '.dentified are to be resolved to comply with the necessary FSAR requirements.

E. CORRECTIVE ACTIONS:

When the flued head anchor assembly concerns were initiated Co'mmonwealth Edison ~

(CECO) reviewed the basis-for the exclusion of the assemblies under the IE Bulletin No. 79-14 and 79-02 scope of work. Subsequently, CECO presented justification for the operability criteria as well as the basis for the exclusion in relation to IEB 79-14 and 79-02. The NRC has disagreed with this exclesion basis, hence, CECO has initiated a comprehensive program to demonstrate the adequacy of the aforementioned flued head anchor structures. An engineering walkdown was conductet during the ongoing Quad-Cities Unit Two refueling outage.

The eleven (11) Unit Two structures were reviewed and Engineering Change Notices were issued. -Modification (M-4-2-88-017) has been initiated to revise the structures to a condition that compiles with FSAR design requirements. The eleven (11) flued head anchors are complete and the anchors meet the required FSAR design requirements. In general, the modification involved the addition of structure baseplates and anchors, the welding of some minor support additions, and changeout of certain concrete expansion anchors (Nuclear Tracking System 2652008801701).

DvR 107 1

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' /,/jI /, NumbtL_ ///j _l!VrbtL JusLLillesaniLIvo LLL13 I o I o i 2LeLLaLL . _JJJJ_L-_LL2_ all JL J1L TEXT In regald to the missing pin, a new pin was installed and a keeper tab was welded to hold the pin in place. Th', should prevent recurrence.

Further analys b of Unit One flued head anchor assemblies showed that assemblies located at the X-11, X-13A, X-13B, X-16A, X-16B, X-23, X-24, X-36, X-47, X-7A, X-78, X-7C, X-7D, X-8, X-9A, X-9B, X-10, X-12 and X-17 penetrations did not meet FSAR .

design requirements. The station implemented modification H-4-1-88-017 to resolve any FSAR design requirement deficiencies. Modification H-4-1-88-017 was completed during the Q1R10 refur. ling outage and authorized for operation on 11-22-89. Under this modification piping to the X-36 penetration was cut and capped at both ends.

This reduced the pipe load to zero and eliminated the need to modify the anchor.

F. EREyl00S EVENTit l

LERJ2BEB TIILE 254/86-022 Containment Atmospheric Monitoring [IL) Line does not meet code allowable stress limits.

254/86-024 U-1 and U-2 Residual Heat Removal Service Hater (80)

Piping Supports exceeded code stress allowable limits.

254/86-025 Torus attached Small Bore Piping does not meet Code Allowable Limits 254/87-008 1C Residual Heat Removal Service Water Pump [P) piping in excess of allowable stress due to sheared anchor bolts.

254/87-011 Residual Heat Removal Support Embedment Pla'te in excess of allowable stress due to improper anchor strap spacing.

254/87-026 Piping Supports Outside Compliance with Safety Analysis Report due to Desigh/ Construction Errcr.

254/87-030 Anticipated Transient Without Scram [JC] Instrument Sensing Lines Inadequately Supported due to Personnel Error and Inadequate Design.

265/87-019 Piping Supports Outside Compliance With Safety Analysis Report due to Design Error.

254/88-004 Reactor Head Vent Line outside Safety Analysis Criteria for Allowable Stress due to Design Error.

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TEXT G. COMP 0HEELIAlLURE DATA:

Since the structures are not considered inoperable, no component failure is identified in this event.

DVR 107

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