ML20043G204
| ML20043G204 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 06/14/1990 |
| From: | Gina Davis, Mcgaha J ENTERGY OPERATIONS, INC. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| LER-89-006, LER-89-6, W3A90-0155, W3A90-155, NUDOCS 9006200037 | |
| Download: ML20043G204 (9) | |
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June 14, 1990 U.S. Nuclear Regulatory Commission ATTENTION: Document Control Desk Washington, D.C. 20555 -
Subject:
Waterford 3 SES l Docket No. 50-382 l License No. NPF Reporting of Licensee Event Report. .
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Dear Gentlemen:
Attached is Licensee Event Report Number LER-89-006-01 for Waterford l Steam Electric Station Unit 3. This Licensee Event Report is submitted 1 pursuant to 10CFR50.73(a)(2)(ii) and provides the resulte of a final )
engineering evaluation and updated corrective action. I Very truly.yours, J.R. McGaha -;
Plant Manager - Nuclear
- JRM/KTW/rk (w/ Attachment) cc Messrs..R.D. Martin J.T. Wheelock INPO Records Center-E.L. Blake W.M. Stevenson D.L. Wigginton NRC Resident Inspectors Office 9006200037 900614 2 PDR ADOCK-0500r s ,
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DOCKIT NUMOtR (2) PAGE(3, FACILITY NAME lit Waterford Steam Electric Station Unit 3 01510[0l01 31812 1 !O d Ol 8 title i4i Safety Class Break Reauirements no: met due to Ir
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A. TRACT m ~,,,,uv ,u ... - aN....y. u.n,....n.,,,...,n i At 1300 be its on March 27, 1989, Waterford Steam Electric Station Unit 3 was operating-at 100% power when Event Analysis & Reporting personnel discovered that two manual isolation valves which supply component cooling water (CCW) to the post accident sampling system (PASS) coolers were shown to be normally open on design drawings. Each valve is located at a safety class 3 (SC3) to non nuclear safety (NNS) class broak in the CCW system. Because these valves do not have remote operators, they are required to be normally closed for leak isolation in the event of a NNS piping break. This event is reportable as a condition outside the plant design basis.
The root cause of this event was a programmatic breakdown in administrative controls. A 10CFR50.59 evaluation was not reviewed by the Plant Operations Review Committee. A contributing cause was inadequate traiting because engineers involved in the review process overlooked the safety class break requirements. The valves have been closed and design drawings and 1 ocedures -
are being revised. A final engineering evaluation has determined titat cooling water would last at least 1.53 hours6.134259e-4 days <br />0.0147 hours <br />8.763227e-5 weeks <br />2.01665e-5 months <br /> assuming a NNS piping break with both valves left open. This would have allowed sufficient time for a leak to be detected and isolated. Thus, this event did not threaten the health or safety of the public or plant personnel.
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ERC Poem 300A - U.S. NUCLEAR KEEULATO2Y COMMISSION
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Ol1 () 2 OF 0l8 nxi w mee ame a w. me eanoonmc rano sasens tm At 1300 hours0.015 days <br />0.361 hours <br />0.00215 weeks <br />4.9465e-4 months <br /> on March 27, 1989, Waterford Steam Electric Station Unit 3 was operating at 100% power when Event Analysis & Reporting (EA&R) personnel discovered that component cooling water (CCW) valves (EIIS Identifier CC-V)
CC-80310B an.1 CC-8062 were shown as normally open on design drawings. The manually operated valves isolate CCW to the post accident sampling system (PASS) three ton chiller (EIIS Identifier IP-Ch'J) and the PASS heat exchanger number two (EIIS Identifier IP. UX) . The PASS titree ton chiller supplies chilled water to the PASS heat exchanger number one (EIIS Identifier IP-HX).
Each valve is located at a safety class 3 (SC3) to non nuclear safety (NNS) class break in the CCW system. The NNS portion of the piping for both valves supplies the chiller and heat exchanger. The SC3 portion upstream of CC-8062 and CC-80310B branches from the CCW supply to Containment Fan Coolers (CFCs)
(EIIS Identifier BK-CLR) 3A-SA and 3B-SB respectively. The CCW return lines from the PASS heat exchanger number two and the three ton chiller discharge'to the common CCW return header from the 'A' and 'B' CFCs respectively. The SC3-to NNS break en these return lines are provided with check valves which automatically close to isn'. ate a leak in the NNS piping.
CC-8062 and CC-80310B are locally manually operated because they are not provided with remote operators. The valves comprise a SC3 to NNS class break, so CC-8062 and CC-80310B should be normally closed to comply with the requirements of section 3.2.2.1 of the Final Safety Analysis Report (FSAR). This specifies that SC3 to NNS boundary valves should be normally closed unless.it can be proven that leaving the valves open would not result in a loss of safety function of the higher class component. An analysis was not performed to
' demonstrate the seismic integrity of the NNS PASS piping, so this piping is assumed to fail and create an unisolated leak assuming CC-8062 and CC-80310B are open. Thus, this event is reportable as a condition outside the plant's design bases. The event date of January 5, 1987, is based on the oldest available valve lineup which shows one of the above valves (CC-80310B) as positioned and verified open.
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0l1 d3 OF gl g TEXT (# more spece 4 segunoir, use esMeamst NAC Form JI541(th An engineering analysis of a break in this piping has determined that up avail-able margin of CCW makeup would last at least 1.53 hours6.134259e-4 days <br />0.0147 hours <br />8.763227e-5 weeks <br />2.01665e-5 months <br />. The activation acd continued operation of the CCW makeup pumps (Ells Identifier CC-P), which take suction from the condensate storage pool (CSP) (EIIS Identifier KA-TK), would provide indication of a CCW 1eak. The activation of the sump pumps (EIIS Identifier WK-P) would provide' indication of the location of a leak. Adequate time would be available for operators to detect and isolate a leak. This event would not have prevented the CCW system from supplying sufficient cooling water to components required for safe shutdown or from mitigating a design basis accident.
Various documents associated with this event were reviewed. On May 8, 1988, Problem Evaluation Information Request (PEIR) 60887 was issued by Operations personnel to Plant Engineering (PE) personnel to evaluate having CCW continuously lined up to the PASS coolers. Operations did not explicitly specify which valves they vere referring to in the PEIR. The reason given for requesting the evaluation was that PASS Cooler CCW isolation valves could bs inaccessible due to high radiation levels during a design bases loss of coolant accident (LOCA). The valves were thought to be located in the -4 foot level of the Reactor Auxiliary Building.(RAB) (EIIS Identifier NF) wing area. In fact,.they are located on the +21 foot level RAB ving area where radiation levels will remain low enough to permit access during accident conditions.
PE personnel concluded that both valves should be left open because they believed the valves were inaccessible during accident conditions. PEIR 70998 was issued by PE to Nuclear Operations Engineering (NOE) personnel to request that design drawings be revised to show CC-8062 and CC-80310B as normally open.
NOE personnel noted that the valves were located on the +21 foot level RAB wing area versus the -4 foot level. Due to the concern about minimizing post LOCA radiation exposure, NOE personnel concurred with PE personnel, missing the safety class break requirements in the FSAR.
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N0E personnel issued Document Revision Notice (DRN) M8800868 in response to PEIR 70998 to revise design drawing LOU 1564 G160 sheet 1. A 10CFR50.59 safety evaluation was performed on July 11, 1988, by N0E personnel which concluded that changing the val'aes f rom normally closed to normally open would not af fect j any portion of the accident analysis, equipment malfunction, or margin of safety. The DRN was completed by November 22, 1986, te show CC-8062 and j CC-80310B as normally open.
Procedure OP-02-003, " Component Cooling Water System," was reviewed. CC-8062 and CC-80310B were added to OP-02-003 in revision 4, whf ch consisted of a major
-rewrite of the CCW system valve lineup. The location cf CC-8062 and CC-80310B was correctly given as the +21 foot level RAB '_B' Switchgear Room at the
-containment (EIIS Identifier NH) wall and at the PASS panel respectively. The required position of CC-8062 was correctly specified as " closed"; however, CC-80310B was specified as "open". Revision 6 to OP-02-003, approved on October 29, 1987, incorrectly changed the stat. 'ocation fet valve CC-8062 to the -4 foot level RAB wing area. The required positiviis of both valves and the stated location of CC-80310B have remained the same since revision 4 was issued.
On March 27, 1989, a walkdown confirmed the location of CC-8062 and CC-80310B on the +21 foot level RAB wing area and determined that both valves were open.
This was due to performance of chemistry procedures CE-3-900, " Operation of the PASS," and CE-3-905,." Testing and Maintenance of the PASS." These procedures did not reshut CC-8062 and CC-80310B after PASS operation,' testing or maintenance was completed. Thus, the valves were left open after completion of the chemistry procedures.
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The above sequence of events describes a series of cognitive personnel errors involving several departments. TM required position of CC-80310B should not have been "open" in revision 4 to OP-02-003 to comply with safety to non-safety class break-requirements. Revision 6 incorrectly changed the location of CC-8062.
Because the required position of this valve was correctly listed as " closed" and revision 6 stated the valve was located in an inaccessible area during a design LOCA (the -4 foot level RAB wing area), operations personnel initiated PEIR 60887 to evaluate changing the valve position. The failure to specify the valve number in PEIR 60887 led PE personnel to believe the locations and positions of both CC-8062 and CC-80310B were in question. PE and NOE personnel involved in the review and approval of PEIRs 60887 and 70998 and DRN M8800868 did not know that changing the valve position would require an evaluation to meet the specifications of FSAR section 3.2.2.1.
The root cause of this event was a programmatic breakdown in administrative controls. The 10CFR50.59 evaluation performed m DRN M8800868 vas not reviewed by the Plant Operations Review Committee (PORC). This additional high level of review and discussion would likely have determined there was an unreviewed safety question, thus preventing implementation of the DRN. Nuclear Operations-Engineering Procedure (N0EP) 305, " Safety Evaluations," has been replaced with Nuclear Operations Procedure (NOP) 013, which requires PORC review of detailed 10CFR50.59 evaluations and requires a more in depth review than previously required by NOEP-305. Detailed safety evaluations associated with the DRN process which had not previously received a FORC review have now been reviewed by DE personnel and PORC.
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LtR NUMBERtel Waterford Steam viaR ag4;p,*6 ;egy; Electric Station Unit 3 Tixt tr m-. . w. -mic i-= assamm A contributing cause of this event was inadequate training for design and-system engineers. This explains why a design criteria was overlooked in the evaluation process by several engineers from both PE and N0E Departments.
A schedule of recurring lectures focusing on the general design criteria for the plant and other topics related to the design modification process was not implemented until recently. The information of FSAR section 3.2, " Classification of Structures, Components and Systems", has been incorporated into the recurring
- trainit.g programs of the PE and - Desig.i Engineering (DE) ' (formerly N0E) depart-ments. '1 raining also has been cone' acted on the 10CFR50.59 evaluation process for site perconnel vi o ordinarily perform these evaluations. This training will assist in prevenring recurrences of incorrect procedural valve lineups.
CC-8062 and CC-80310B were closed on Nrch 27, 1989. PE personnel revised their response to PEIR 6088) :teting the valves should be normally closed. DRN M8900525 has been issued to correct design drawing LOU 1564 G160 to show both -
valves as normally closed. Chemistry procedures CE-3-900 and CE-3-905 have been revised to show correct valve locations and verification of position and restoration. OP-02-003 has been changed to show correct valve locations and positions, and OP-100-009, " Control of Valves and Breakers " has been changed to include both valves on the locked valve list.
A final engineering evaluation by DE personnel has shown there would be at least a 1.53 hour6.134259e-4 days <br />0.0147 hours <br />8.763227e-5 weeks <br />2.01665e-5 months <br /> margin from a postulated break in the NNS piping with CC-8062 and CC-80310B open. This would have allowed Operators sufficient time to detect and isolate a leak. Thus, this event did not threaten the health or safety of the public or plant personnel.
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SIMILAR EVENTS A review of previous Licensee Event Reports (LERs) reveals two similar events. LER 87-029 described an inadequate design change which connected two non-safety control room emergency DC lighting panels (EIIS Identifier FH-PL) to two safety related (class 1E) DC power distribution panels (PDPs) (EIIS Identifier EJ-PL), each through a single circuit breaker. Additionally, a non-lE telephone system cabinet (EIIS Identifier FI-TEL) was connected to a lE PDP through a single circuit breaker. These designs were not in compliance with Regulatory Guide (RG) 1.75 committed to in the FSAR.
The root cause of LER 87-029 was classified as cognitive personnel error. A review of documentation associated with the event reveals a series of personnel errors by various engineers in developing, implementing, and reviewing these design changes.
Another series of personnel errors involving a station modification was described in LER 88-026. Tubing supports for the Boric Acid Makeup Tank (BAMT) level indicator (EIIS Identifier CB-TK-LI) were discovered not to have been installed as seismically qualified during construction closeout, so an evaluation wa.s performed to justify the non scismic installation. This evaluation was not taken into account when the BAMT Technical Specifications were changrd to allow a lower boric acid concentration.- This would not allow the BAMTs to satisfy their design requirement with the lower boric acid concentration assuming a postulated break in the non-seismic tubing.
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PACILITY NAME (18 DoCKit NUMBER (2) LIR NUMSER (4) PA06 (3)
Waterford Steam _
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O ll 0 18 OF 0 l8 inxi ta === m a w. me eneane mm ansa enns Programmatic _ enhancements were identified in the design change program and nonconformance review process as corrective action for the above two events.
It is believed that this action combined with-improvements in the recurring-training program should prevent similar recurrences of this type of event.
PLANT CONTACT G.M. Davis, Event Analysis, Reporting & Response Manager, 504/464-3153
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