IR 05000206/1990020
| ML20043F460 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 05/23/1990 |
| From: | Johnson P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20043F456 | List: |
| References | |
| 50-206-90-20, 50-361-90-20, 50-362-90-20, NUDOCS 9006150032 | |
| Download: ML20043F460 (10) | |
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U. S. NUCLEAR REGULATORY COMMISSION
REGION V
Report Nos.
50-206/90-20,50-361/90-20,50-362/90-20 j
Docket Nos.
50-206, 50-361, 50-362 License Nos.
DPR-13 NPF-10, NPF-15 Licensee:
Southern California Edison Company Irvine Operations Center 23 Parker Street Irvine, California 92718 Facility Name: San Onofre Units 1, 2 and 3 Inspection at: San Onofre Site, San Clemente, California Inspection conducted: April 30 through May 4, 1990
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Inspector:
J.F.Burdoin, Pro {ectInspector
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[ A P14 7 5/23/90 Approved by:
P.HfJohnson, Chief Date Signed Reactor Projects Sectb' n 3 Sununary:
Inspection on April 30 through May 4, 1990 (Report Nos. 50-206/90-20, 50-361/90-20, 50-362/90-20).
Areas Inspected:
An unannounced routine inspection by one regional inspector of various vital areas, equipment in the plant, resolution of enforcement items, followup items, licensee event reports, and compliance with the ATWS Rule,10 CFR 50.62.
Inspection Procedures Nos. 30703, 71707, and 92700, 92701, 92702 and TI 2500/20 were used as guidance during the inspection.
Results:
General Conclusions and Specific Findings:
The licensee's actions taken to correct deficiencies resulting from inspection findings appeared thorough, timely, and properly documented.
Significant Safety Matters:
None Summary of Violations or Deviations: None Open Items Suninary:
Four open items were closed; one item was examined and left open.
9006150032 900530 PDR ADOCK 05000206 O
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DETAILS 1.
Persons Contacted Southern California Edison Company
- R. Bridenbecker, Vice President Station Manager J. Anderson, Coordinating Supervisor Unit 1
- R. Baker, Project Engineer, Units 2/3
- C. Brandt, Quality Assurance Engineer
- D. Brevig, Onsite Nuclear Licensing Supervisor
- A. Brough, Engineering Supervisor
- L. Cash, Manager, Maintenance
- R. Clark, Station Technical C. Diamond, NEDO, Mechanical Engineer G. Gartland, Station Technical. NSS$
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H. McShane. NEDO, Mechanical Engineer H. Moran, NEDO, Materials Control Supervisor
- H. Morgan, Station Manager
- S. Morris, Licensing Engineer R. Ornelas. Licensing Supervisor, Unit 1
- J. Patterson, Assistant Manager, Maintenance, Units 2/3
- D. Peacor, Manager, Station Emergency Preparation
- N. Quigley, Station Technical J. Rainsberry, Licensing Supervisor, Units 2/3 J. Robertson, Compliance Engineer A. Schramm, Superintendent, Unit 1
- P. Shaffer, Compliance Supervisor
- K. Slagle Deputy Station Manager
- R. Waldo, Assistant Manager, Station Technical
- D. Werntz, Licensing Engineer San Diego Gas and Electric Company l
- R. Erickson, Senior Engineer
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The inspector also talked with other lionsee personnel during the course L
of the inspection.
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- Attended the exit meeting on May 4, 1990.
In addition, the NRC resident inspectors attended the exit meeting.
I 2.
Area Inspection (71707)
An independent inspection was conducted in the Units 1, 2 and 3 Control and Auxiliary Buildings. The inspector examined areas and equipment for debris, potential hazards, oil and water leakage, and equipment condition; e.g., oil level, valve position, and electrical connection
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configuration and cleanliness.
The equipment and areas inspectec inch ded:
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F a.
4160V Switchgear Room, Buses IC and 20, and 480V Bus No. 1
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b.
4BOV Switchgear Room 2. Buses 2 and 3 c.
125V Battery Rooms 1 and 2 d.
Main Feedwater Pumps FWS G-3A/3B
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e.
Auxiliary Feedwater Pumps Trains A and B
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Diesel Generator Rooms 1 and 2 g.
Designated Shutdown Diesel Generator h.
Safety Injection valves HV 851 A&B, 853 A&B and 854 A&B.
Unit 2 a.
4160V/480V Switchgear Rooms. Trains A and B b.
D-G Rooms 2G002/2G003, Trains A and B t
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Cable Spreading Room d.
Safety Equipment Rooms. Trains A and B
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Unit 3
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a.
4160V/480V Switchgear Rooms. Trains A and B b.
D-G Rooms 3G002/3G003 Trains A and B c.
Cable Spreading Room i
Those minor housekeeping items found during the walkdown of the above areas were brought to the attention of the licensee and were corrected.
Housekeeping and equipment status appeared to be acceptable.
No violations or deviations were identified.
3.
Followup of Previously Identified Items (92701)
a.
(Closed) 50-206/89-07-01. Steam Generator Yube Inspection Surveillance Interval.
l This item concerns the licensee's request to NRR to conduct a steam generator tube inspection (initially scheduled for March 7,1990)
during the outage beginning June 30, 1990.
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On January 2,1990, the NRC issued an " Order Confirming Licensee Commitments on Full-Term Operating License Open Items." The inspector examined this order in detail. The order confirmed the licensee's request to conduct the upcoming steam generator tube inspection during the Cycle XI refueling outage commencing June 30, 1990, rather than by March 7,1990. This item is closed.
b.
(Closed) 50-206/87-31-01, comparison Of Proposed T.S. To Standard T.S. Awaiting NRR Review This item involved a concern whereby the licensee was operating the facility in accordance with proposed fire protection technical specifications which were at the time under review by the NRC staff.
These proposed f, ire protection technical specifications were
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developed as a result of modifications made to the facility to satisfy the provisions of Appendix R.
This item was retained as an open item pending NRR's review as to the adequacy of the licensee's
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proposed fire protection T.S.
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The review of the licensee's proposed fire protection T.S. by the J
NRR staff was completed and approved under amendment No.131 to the Unit 1 license, issued November 15, 1989. The inspector reviewed
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the licensing amendment and examined T.S. Sections 3.14-1 thru 3-14-26 and Sections 4.15-1 thru 4.15-12 in detail, whicn contain the limiting conditions for operation and surveillance sections for fire protection.
The inspector also examined the Unit I control room copy of the T.S.
and audited the following surveillances to verify that they were
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being performed:
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501-12.2-4, " Fire Water Level Check"
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S01-12.3-18. " Fire System Valve Alignment and Operability Check"
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The control room technical specifications had been revised on November 27, 1989 to include amendment 131, and those surveillances sampled were being done on time and in accordance with the requirements of the T.S.
This item is closed, c.
(Closed) 50-361/362/89-16-06, LPSI Pump Seal Leakaoe This item involves various aspects of the licensee's evaluation relatedtoareportedevent(seeUnit3LER89-08-LO)ofan unexpected amount of leakage from a LPSI pump seal.
On June 29, 1989, while perfonning an Inservice Test (IST) of LPSI pump 3P-015, a seal leek of 1400 cc/ min was measured during pump operation. No leakage was observed while the pump was idle. The Units-2/3 UFSAR specified 500 cc/ min. as a realistic amount of leakage that would occur from an ESF system pump seal under a " gross seal failure." This event was reviewed by an NRC inspection team
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during an inspection on July 10 through 21, 1989, detailing the SCE investigation and evaluation of the event in Region V Inspection l
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Report 50-206/361/362/89-16 dated August 29, 1989.
Abstract of Event (LER 89-08-L2)
At 0606 on 6/27/89, with Unit 3 at 75% power, Low Pressure Safety t
l Injection (LPSI) pump 3P015wasremovedfromserviceforpreventive
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maintenance (PM). At 0855 on 6/29/89, upon completion of the PM, an inservice test was conducted on 3P015. Excessive mechanical seal leakage was observed necessitating seal replacement. Since seal replacement would require more time than remained in the Technical Specification Action Statement, at 0630 on 6/30/89, a unit shutdown was initiated.
In accordance with procedures, an Unusual Event (UE)
l was declared. At 0745 on 6/30/89, the UE was terminated. At 1134,
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the unit entered Mode 3 and at 1830, entered Mode 4.
The mechanical seal was replaced and the pump was returned to service on 7/8/89.
Investigation has determined that the seal failure was caused by oil in contact with the seal o-ring. The source of the oil is unknown, however, it is believed to be from either an overfilling of the lower motor bearing oil reservoir or leakage from the threaded connection between the lower bearing oil gauge fill tube and the lower bearing cartridge.
During a review of this event, discrepancies between the Updated Final Safety Analysis Report (UFSAR) and the configuration of the Unit 3 LPSI pump seal leakoff lines were identified. These seal leak-off lines have been modified to conform to the UFSAR description.
Cause of The Event:
Investigation had determined that the excessive seal leakage was caused by failure of a seal o-ring. During disassembly of the pump, oil was evident on the external surfaces of the mechanical seal gland and in the vicinity of the Ethylene Propylene Rubber (EPR)
o-ring. Oil is known to cause rapid swelling of EPR o-rings. This
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in turn resulted in catastrophic failure of the carbon insert, leading to the observed seal leakage.
The licensee's investigation identified two probable sources for oil found in the vicinity of the EPR o-ring. The first source of oil could have been an overfilling of the lower motor bearing oil reservoir resulting in the overflow of the oil down the pump shaft into the o-ring.
The investigation has been unable to definitely determine when, or if, the lower motor oil reservoir was overfilled. Overfilling could
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have occurred following the removal of a oil sample (as part of the PM work being performed on the pump) on 6/27/89. Records show that no oil had been added to the pump between the last IST on 3/25/89
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and perfonnance of the PM.
The second probable source of oil found in the vicinity of the
o-ring was discovered on 7/21/89. A minor amount of oil leakage was observed at the lower motor bearing housing. The source of this oil is believed to be the threaded connection between the lower bearing oil gauge fill tube and the lower bearing cartridge, Corrective Actions:
To prevent recurrence, the following actions have been taken:
(1) The mechanical seal was replaced and the pump returned to service on 7/8/89.
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(2) This event has been reviewed with appropriate personnel with emphasis on the need to avoid overfilling oil reservoirs when adding oil to pumps.
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(3) As a design enhancement, oil deflectors have been installed on
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all Units 2 and 3 LPSI and conteiranent spr6y (CS) pumps. The
oil deflectors, which are located on the pump shaft, will prevent oil from entering the mechanical seal if leakage froni or overfill of the lower motor bearing occurs. Additionally, the oil deflector should prevent water from shaft seal leakage
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from spraying and damaging the lower motor bearing.
(4) AppropriateMaintenanceOrder(s)andProcedure(s)usedfor
adding oil to LPSI pumps and pumps which are similar in design
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to the LPSI pump (i.e., single-stage, vertical, centrifugal),
have been revised to incorporate precautions to avoid overf1111ng oil reservoirs.
The inspector in following up this open item discussed with the following licensee personnel the above corrective actions, as outlined in LER 89-08 Revision 2; cognizant engineer, assistant l
manager-station technical services, and licensing supervisor, Units 2/3.
The inspector learned from these discussions that a number of telephone conference calls on the issue took place betwesn the licensee and the NRR project manager for Units 2/3 following the issuance of the original LER on July 31, 1989. All aspects of the causes and effects of this event and those issues raised and described in Region Y Inspection Report 50-361/362/89-16 were resolved.
The inspector verified the above corrective actions by inspecting in the field the installed oil deflectors on the LPSI and CS sumps of both trains for Unit 2.
The inspector also inspected in tie field the modifications to Unit 2 LPSI pump seal leakoff lines for both l
l trains. The field changes appeared to conform to the committed l
corrective actions.
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The inspector examined revisions incorporated into Operations Procedure 50123-0-9 under TCN 0-21, issued 9/20/89.
Precautions to
avoid overfilling oil reservoirs were included in Sections 4.12 and
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6.4.2.4.
The inspector examined the changes to Section 6 and 15 of the FSAR, which were initiated as a result of this LER. The revisions to the operating procedure and the changes to the FSAR appeared to be adequate.
The inspector also coordinated the review of this open item with the
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Region V Radiation Specialist and the NRR Project Mt. nager assigned to SONGS Units 2 and 3 for their concurrence in closing this item.
This item is closed.
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d.
(Closed) 50-206/89-07-04 CCW Flow To RHR Heat Exchanger Temperature Control Valve. Single Failure The licensee identified that during certain failure scenarios, the component cooling water (CCW) throttle valves (TCV-601A & B) for the residual heat removal (RHR)/ letdown heat exchangers could fail open and allow the CCW pumps to exceed run-out limits and divert flow
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from the recirculation heat exchangers.
The licensee's resolution of this problem was to isolate CCW to one train of RHR heat exchangers and to limit the maximum flow to the
in-service RHR heat exchanger to 500 gpm by limiting the stroke on
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valves TCV-601A & B (PFC 89-008). The licensee contracted Stone and
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Webster Engineering Cor) oration to determine the minimum CCW system flow requirements for tie in-service RHR heat exchanger (to verify that 500 gpm was adequate), and to determine whether the proposed resolution would satisfy RHR heat removal requirements and resolve CCW pump run-out concerns. The resident inspector reviewed the Stone and Webster calculations and the safety evaluation contained in the PFC. Certain concerns remained at that time to be followed i
up during a future inspection. These concerns grew out of discussions between the resident inspector and the licensee on the subjects of the Stone and Webster calculations and the Safety Evaluation in the PFC. The concerns were:
(1) A clear definition of the design bases, including system parameter requirements, for the RHR and CCW systems was not i
included in the analysis. The licensee's safety evaluation did not address the effect of reducing CCW flow through the RHR heat exchangers on these design bases.
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The licensee stated that the safety evaluation would be revised to reflect these results and to more clearly state the design bases of the system. This item remained open pending licensee action to properly document and address the system design basis in the safety evaluation.
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(2) The accuracies of the analyses, including consideration for assumptions that were made, were not documented.
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(3) The analysis (Stone and Webster Calculation 18872-NP(B)-002)
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l stated that the RHR heat exchanger inlet temperature could
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reach 550 degrees F during a load rejection, but the design
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I data sheet stated that the RHR heat exchanger was only designed l
for 400 degrees F.
This condition was not evaluated.
I (4) The analysis concluded that the onset of cavitation could occur in the operating pump during post-LOCA recirculation operation, and it was judged that this condition would be acceptable for approximately 30 days without resulting in pump failure. The licensee tested the CCW pump under worst-case flow conditions for several hours and confirmed that the pump did not exhibit any fluctuations in flow, discharge pressure nr running current, although the cognizant engineer noted minor
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indications of void collapse noises at the pump discharge.
It was noted that the licensee had not obtained vendor concurrence with the conclusion that long term pump operation with onset of cavitatiun was acceptable.
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For additional information on the above concerns, see Region V Inspection Report 50-206/89-07 dated June 6, 1989.
The inspector examined the following documents to verify the
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licensee's commitments and corrective actions to address the above concerns.
(a) PFC 1-89-008 Install stem travel limiting collars on
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SI-CCW-TCV-601A and 601B valves.
(b) NCR 501-P-7005 Revision 2, dated 6/19/89.
(c) Safety Evaluation, dated 6/8/89 included in NCR 501-P-7005.
(d) Attachment to Safety Evaluation "CCW/RHR Design Basis Requirements", revised 6/23/89.
(e) Memorandum for file, dated 5/10/89, and signed by the NEDO mechanical project engineer.
(f) Calculation No. D0-3180, dated 5/8/89, "RHR Hx (E-21A) Subject to a low flow and high temperature conditions."
Thedocumentsidentifiedinitems(b)and(c)addressedthefollowup concernsofitem(1)above. The concerns of item (2) above are
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addressed in documents (b), (c) and (d). The concerns of items (3)
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and (4) above are addressed in documents (a) through (f). After having first examined the above documents and correlating them with the concerns, each of the concerns identified above were reviewed by the inspector with the licensee's project mechanical engineer to resolve any questions. This item is closed.
4.
(0 pen) Compliance with ATWS Rule,10 CFR 50.62 (Temporary Instruction 2500/20 Revision 1)
The purpose of this inspection was to assess the licensee's efforts to comply with the Anticipated Transient Without Scram (ATWS) rule, 10 CFR 50.62.
Specifically, the inspector intended to update the status of procurement, installation and testing of ATWS equipment for the three plants. Two previous Region V inspection reports that addressed portions of ATWS are 50-206/361/362/88-02 and 50-206/362/88-17 with 50-361/88-16.
Unit 1 The modifications to be made on Unit 1 to satisfy the requirements of the Anticipated Transients Without Scram (ATWS) Rule, 10 CFR 50.62 are:
(a) Provide a diverse system for initiating auxiliary feedwate _
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(b) Provide a diverse system for tripping the main turbine.
(c) A TMI upgrade required that a dedicated safe shutdown train of auxiliary feedwater be provided. A followup of this item was inspectedalongwithItem(a).
Item (a)and(c)wereaccomplishedduringthecycle10refuelingoutage which was completed in September 1989.
Item (b) is to be installed during the cycle 11 refueling outage scheduled to commence June 30, 1990.
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Unit 2/3 The modifications to Units 2 and 3 to satisfy the requirements of the ATWS rule are:
(a) Provide diverse systems for initiating reactor trip and main turbine trip.
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(b)
Provide a diverse system for initiating auxiliary feedwater.
Item (a) was accomplished on Unit 2 during the Cycle 5 refueling outage completed in December 1989. These modifications are scheduled to be completed on Unit 3 during the present (Cycle 5) refueling outage which commenced April 13, 1990.
The design for Item (b) is still being reviewed by NRR and is presently
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scheduled for installation on Unit 2 during the Cycle 6 refueling scheduled for the Spring of 1991. The installation of this item in Unit 3 will be accomplished no later than one month after Unit 2 startup
following the Cycle 6 outage. This modification may iequire a special
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outage for Unit 3 imediately after Unit 2 returns to service.
The inspector examined the following documents to evaluate the licensee QA controls that apply to the procurement and installation of ATWS equipment:
Unit 1 P.O.8K010116, Two, Solenoid valves for diverse turbine trip system,
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Non-Safety)related (addressed in the licensee's QA Manual
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Section 8B.
i P.O.441207-4-0601-01, One motor driven auxiliary feedwater pump.
QAP N10.02, Rev. 9. Receiving Inspection Log, dated 2/7/86.
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Report 441207, FieldReceivingReport(FLUOREngineers,Inc.),
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dated 2/7/86.
Units 2L3 l
P.O. SPP00113. Two, complete anticipated transient without scram / diverse SCRAM systems.
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Receiving Inspection Data Report (RIDR), RSO 2069-88, dated
7/22/88.
P.O. SPP00137. Eight, Rosemount Model 1154 Pressure Transmitters...
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ATWS/ Diverse Reactor Trip Systems.
RIDR, RS0-3178-87, dated 12-22-87.
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P.O. 6J00013, Eight, Controllers....ATWS/ Diverse Reactor trip systems.
RIDR, RS0-0985-90, dated 4/23/90.
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RIDR, RS0-1067-90, dated 4/30/90.
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Proposed Facility Change (PFC) 2-87-6553, Add a backup reactor trip system that is independent and diverse from the existing reactor trip system (Unit 2).
The review of the above documents attested to the requirements of quality assuranca and quality control in the manufacturing, delivery and receiving, and installation of safety related equipment in meeting the require m ts of the ATWS rule.
This item will remain open for inspection in the field of the completed installations referenced above and for future inspections of portions of ATWS which remain to be completed.
No violation or deviations were identified.
5.
Exit Meeting An exit meeting was held with members of the licensee's staff (see paragraph 1) on May 4, 1990. The specific concerns addressed in this report were discussed with the licensee during this meeting and were acknowledged by the licensee.
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