IR 05000269/1990033
| ML15224A742 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 12/11/1990 |
| From: | Binoy Desai, Poertner W, Shymlock M, Skinner P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML15224A740 | List: |
| References | |
| 50-269-90-33, 50-270-90-33, 50-287-90-33, GL-88-17, NUDOCS 9012280283 | |
| Download: ML15224A742 (10) | |
Text
6pa REGq UNITED STATES NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTA STREET, NX ATLANTA, GEORGIA 30323 Report Nos.:
50-269/90-33, 50-270/90-33, 50-287/90-33 Licensee: Duke Power Company P. 0. Box 1007 Charlotte, NC 28201-1007 Docket Nos.:
50-269, 50-270, 50-287, 72-4 License Nos.:
DPR-38, DPR-47, DPR-55, SNM-2503 Facility Name:
Oconee Nuclear Station Inspection Cond ted:
October 28 - December 1, 1990 Inspector:
e-ae P.H Skinner,,/ e ior Resident Inspector Date Signed B. B.saiei t Inspec or Dae igned
W. Poer er, sident I Date Signed Approved by:
M. B. Shymlock, Sect'
hief ate Signed Division of Reactor Pro*ects SUMMARY Scope:
This routine, announced inspection involved inspection on-site in the areas of operations, surveillance testing, maintenance activities, reduced inventory activities and inspection of open item Results:
Two violations were identified, one for improper operation of a valve which resulted in running a safety-related pump without proper recirculation flow (paragraph 2.f.) and a second violation for an inadequate procedure used during testing of a Building Spray pump (paragraph 3.b.).
Improvements in watchstanders attention to detail were also noted while Unit 2 was in a reduced inventory conditio PDR ADOCK 05000269 G
REPORT DETAILS 1. Persons Contacted Licensee Employees
- B. Barron, Station Manager
- C. Baldwin, Quality Assurance D. Couch, Keowee Hydrostation Manager
- T. Coutu, Operations Manager
- T. Curtis, Compliance Manager
- J. Davis, Technical Services Superintendent D. Deatherage, Operations Support Manager
- B. Dolan, Design Engineering Manager, Oconee Site Office
- W. Foster, Maintenance Superintendent T. Glenn, Engineering Supervisor D. Hubbard, Performance Engineer
- E. LeGette, Compliance Engineer C. Little, Instrument and Electrical Manager H. Lowery, Chairman, Oconee Safety Review Group B. Millsap, Maintenance Engineer D. Powell, Station Services Superintendent G. Rothenberger, Integrated Scheduling Superintendent R. Sweigart, Operations Superintendent Other licensee employees contacted included technicians, operators, mechanics, security force members, and staff engineer NRC Resident Inspectors:
- Skinner
- W. Poertner B. Desai
- Attended exit intervie. Plant Operations (71707) The inspectors reviewed plant operations throughout the reporting period to verify conformance with regulatory requirements, Technical Specifications (TS), and administrative controls. Control room logs, shift turnover records, temporary modification log and equipment removal and restoration records were reviewed routinely. Discussions were conducted with plant operations, maintenance, chemistry, health physics, instrument & electrical (I&E), and performance personne Activities within the control rooms were monitored on an almost daily basi Inspections were conducted on day and on night shifts, during weekdays and on weekend Some inspections were made during shift change in order to evaluate shift turnover performanc Actions observed were conducted as required by the Licensee's Administrative Procedure The complement of licensed personnel on each shift inspected met or exceeded the requirements of T Operators were responsive to plant annunciator alarms and were cognizant of plant condition Plant tours were taken throughout the reporting period on a routine basis. The areas toured included the following:
Turbine Building Auxiliary Building CCW Intake Structure Units 1, 2 and 3 Electrical Equipment Rooms Units 1, 2 and 3 Cable Spreading Rooms Units 1, 2 and 3 Penetration Rooms Units 1, 2 and 3 Spent Fuel Pool Rooms Unit 2 Containment Station Yard Zone within the Protected Area Standby Shutdown Facility Keowee Hydro Station During the plant tours, ongoing activities, housekeeping, security, equipment status, and radiation control practices were observe Unit 1 operated at power for the entire reporting perio Unit 2 commenced this reporting period at 100 percent power operatio On November 8, the unit was shutdown for repairs to the
"A" steam generator (see paragraph 2.b.).
The unit returned to power operation on November 17 and continued operation for the remainder of this reporting perio Unit 3 commenced this reporting period operating at 100 percen On November 13, the unit was tripped due to a rod control system electrical malfunction (see paragraph 2.d.).
The unit was returned to power on November 14 and remained at power for the remainder of the report perio b. Unit 2 Primary to Secondary Leak in the A Steam Generator On November 7, 1990, with Unit 2 operating at 100 percent power, the operators identified a gradual increasing count rate on the condensate system air ejector radiation monitor (RIA-40).
Calculations were performed indicating a leak of.008 gpm leaking from the primary system to the steam generators (SGs).
As the level on RIA-40 continued to increase, the calculated leak rate also increase Samples taken on the SGs and radiation monitoring of the main steam lines identified the leak to be in the "A" S On
November 8, when the leakage reached approximately 0.1 gpm, the licensee reduced power from 100 percent to 60 percent while monitoring the leakage rat At approximately 12:15 p.m.,
the licensee commenced a controlled shutdown to identify and correct the source of the leakag The tube was subsequently identified to be number 77-31 at the 15th tube support plate area. This tube had been inspected in the most recent outage on Unit 2 using the bobbin coil method and no flaws were identified. The licensee performed rotating pancake probe inspections on 228 additional tubes in the vicinity of the flawed tube and identified three additional tubes that met the plugging criteri C. Reduced Inventory Activities (71707)
As a result of the Unit 2 Steam Generator tube leak, the unit was placed in a reduced inventory statu The inspectors completed the actions required by the Midloop/Reduced Inventory Activities Checklist (see L. Reyes memorandum dated April 11, 1990).
Generic Letter (GL) 88-17 and TI 2515/101 were reviewe The inspectors observed that operators displayed a high level of attention when operating in this conditio OP/2/A/1103/11, Draining and Nitrogen Purging of the Reactor Coolant System, provides the controls during this perio These controls were discussed with shift management to assure that all aspects of control were understood by personnel assigned responsibilities for these action The inspectors noted that an ultrasonic level instrument (temporary)
was installed on a cold leg and operable although level was not reduced to within the indicating range of this instrument Since this outage was of limited duration to correct the SG A tube problem, most normal systems remained available for operation, if neede d. Unit 3 Reactor Trip (93702)
On November 13, 1990, at 10:56 p.m., Unit 3 was manually tripped from approximately 60 percent powe The unit had been operating at 100 percent power when all group seven rods dropped into the core. The dropped group caused a rapid reduction in power to approximately 60 percent and the operators manually tripped the reactor prior to automatic actuation by the Reactor Protection System. The post trip response was norma Investigation by I and E technicians identified a failure of the rod programmer for group seven in the Control Rod Drive System. The programmer was replaced, functionally tested and the reactor returned to 100 percent power on November 14, at 11:54 Notification was made to the NRC in accordance with 10 CFR 50.72 (b)(2)(ii) at 12:15 a.m. on November 1 e. High Pressure Injection Unanalyzed Condition On November 20, 1990, as a result of a Self Initiated Technical Audit (SITA),
the licensee determined that a loss of coolant accident (LOCA)
on a High Pressure Injection (HPI) System injection line resulted in an unanalyzed conditio Based on the present TS, two HPI pumps are required to be operable when RCS temperature is greater than 350 degrees F and reactor power is less than 60 percen The most limiting single failure during a LOCA event could result in one HPI pump injecting into two cold leg If a break is postulated in one of the RCS cold legs approximately 50 percent of the HPI flow was assumed to be lost out the break and the remaining HPI flow was assumed to enter the RCS through the intact cold le A 50/50 flow split was assumed to occur since both injection lines would be exposed to RCS pressur If the break is postulated to be in one of the two HPI injection lines between the cold leg nozzle and a HPI check valve, this injection line would be exposed to atmospheric pressure and would result in less than the assumed 50 percent flow reaching the RC The licensee determined that two HPI pumps injecting through two trains exceeds the HPI flow requirements for a HPI line break from full powe The TS requirements for operation above 60 percent full power require three HPI pumps to be operable and the licensee determined that the requirements for operation above 60 percent power bound all conditions for Emergency Core Cooling System (ECCS)
operabilit Based on this review, the licensee committed to apply the TS requirements for operation greater than 60 percent power to all cases where HPI is required to be operable. The NRC was notified of this condition at approximately 4:00 p.m. on November 20, 1990, in accordance with 10 CFR 50.72(b)(1)(ii)(A). At the time of the event, all three units were at 100 percent powe The inspectors reviewed the licensee's short term corrective actions for this item and found them acceptabl The inspectors will continue to monitor the licensee actions associated with this proble f. TDEFW Pump Operation with Improper Valve Lineup At 0200 on November 1, 1990, valve 2FDW-89, the turbine driven emergency feedwater (TDEFW) pump minimum flow recirculation orifice block valve, was found locked closed by the license Procedure OP/2/A/1106/06, Emergency Feedwater System, requires this valve to be locked ope The valve was opened and locked and action was initiated to determine the cause of the valve being shut and the effect of the valve being shut on the operability of the TDEFW pum The licensee determined that the valve was closed on October 20, 1990, for maintenance activities on the system and was not re-opened during performance of the system valve lineup as required by the procedur During the period, October 20 to November 1, the licensee conducted two performance tests with 2FDW-89 unknowingly close A review by the licensee determined that on each of the two tests the TDEFW pump was operated with the valve shu After five minutes other
valves in the system were opened to establish flo The licensee performed the pump operability performance test (PT/2/A/600/12)
after 2FDW-89 was properly aligned and the pump met operability requirements of the procedur No degradation of performance was noted when compared to past performance data including the data obtained with 2FDW-89 shu The licensee also determined that during a loss of main feedwater event, it would take approximately four minutes for the steam generator water level to drop to the point where the EFW system would inject into the steam generator The licensee's engineering analysis determined that the TDEFW pump was operable for the time that 2FDW-89 was out of its normal alignment since it would have operated under the same conditions if it. had been automatically actuated following a loss of main feedwater event and the pump did not sustain any detrimental effect The inspectors followed the licensee's corrective actions and reviewed the pump testing data and the licensee's engineering evaluation for pump operabilit The licensee's engineering evaluation did not address the long term effects on pump operability of operating for extended periods with 2FDW-89 shu The inspectors also expressed concern that 2FDW-89 had been independently verified as being locked open when in fact the valve had actually been locked shu This is identified as Violation 50-270/90-33-01:
Failure to Maintain the Emergency Feedwater System in accordance with OP/2/A/1106/06, Emergency Feedwater Syste One violation was identifie.
Surveillance Testing (61726)
a. Surveillance tests were reviewed by the inspectors to verify procedural and performance adequac The completed tests reviewed were examined for necessary test prerequisites, instructions, acceptance criteria, technical content, authorization to begin work, data collection, independent verification where required, handling of deficiencies noted, and review of completed wor The tests witnessed, in whole or in part, were inspected to determine that approved procedures were available, test equipment was calibrated, prerequisites were met, tests were conducted according to procedure, test results were acceptable and systems restoration was complete Surveillances reviewed and witnessed in whole or in part:
PT/1/A/0204/07 Reactor Building Spray System Performance Test IP/01A/0310/14A Engineered Safeguards System Analog Channel A on Line Calibration PT/O/A/0110/15 Control Room Pressurization PT/3/A/0600/12 Turbine Driven Emergency Feedwater Pump Performance Test IP O/A/0301/03 S SR-IR Channel Test IP 3/A/305/3 NI & RPS Channel Calibration and Functional Test IP O/A/310/12B ES On-line Low Pressure Injection Channel 3 PT 3/A/600/10 RCS Leakage Calculation b. Review of Reactor Building Spray (RBS) Pump Testing On November 16, 1990, during review of Performance Test (PT)/E/A/0204/07, Reactor Building Spray System Performance Test, performed on the 1A RBS pump as a post maintenance test, the inspector determined that the procedure did not specify the required flow band as required by the licensee's definition of ASME Section XI pump testing requirement The licensees program allows the reference flow value to vary according to the accuracy of the flow instrumen The flow band established in the procedure was based on an instrument accuracy of 3 percent; however, the instrument is actually calibrated to 1.5 percen The flow band specified in the procedure was 1230-1350 gpm whereas the correct flow band per the licenseeas guidance for pump testing should have been 1260-1320 gp The flow value recorded in the procedure was 1230 gpm which was within the procedural guidance of the procedure but outside the required flow band per the licensees requirements for pump testin Review of the pump differential pressure data by the inspector indicated that there was not an operability concern with the pum The licensee had been aware of this procedure problem as a result of a previous inquiry made by the inspectors regarding flow bands for pump testing (See Inspection Report 50-269,270,287/89-30) and had subsequently corrected the Unit 3 procedures to reduce the allowable fIow ban When this item was identified to the licensee, the performance test on the 1A RBS pump was satisfactorily repeated using the adjusted flow band requirement Discussions with the performance engineer determined that the Unit 1 procedure had not been changed prior to the post maintenance test since it was considered a programmatic proble This is identified as Violation 50-269/90-33-01:
Failure of Procedure PT//A/0204/07 to Adequately Incorporate the Licensees Requirements for ASME rSection X I Pump Testin. Maintenance Activities (62703)
a. Maintenance activities were observed and/or reviewed during the reporting period to verify that work was performed by qualified personnel and that approved procedures in use adequately described work that was not within the skill of the trad Activities, procedures, and work requests were examined to verify; proper authorization to begin work, provisions for fire, cleanliness, and exposure control, proper return of equipment to service, and that limiting conditions for operation were me Maintenance reviewed and witnessed in whole or in part:
WR 525563 Test manual override circuitry for R.C. make-up pump on U-2 WR 30621 Ground strap not terminated WR 93812C Check thermals on 2LPSW-9 to verify correct setting WR 30784C Turbine rm sump level gage does not work properly WR 30745C Repair U-3 SFP filtered exhaust Fan #2 MTR. BK WR 519373 Investigate/repair loss of output on CRD Group 6 programmer WR 50527 Low voltage and output on DCSF Normal Battery Charger Issue WR 57378E Lubricate 2LPSW 16 WR 57380E Lubricate 2LPSW 18 WR 93797C Hydrogen Recombiner Discharge Flange Incore Tube Assembly Crack On November 8, 1990, Unit 2 was shutdown due to a primary to secondary leak in steam generator A (see paragraph 2.b).
Upon entry into containment a tour was conducted to assess material condition This is a routine tour conducted following normal shutdown During this tour, operations personnel noted two incore instrument tubes that had an accumulation of boron on the tubes. A work request was generated to investigate the source of this problem. On November 12, 1990, while in a cold condition, an I and E technician investigated this problem. One tube cleaned up with no problem, but on the second tube (L-2) after cleaning off the boron, the technician identified a questionable indication on the source tub Further investigation identified a circumferential indication approximately 1/2" in length that contained a through wall crac It was located above the closure assembly to guide tube weld in the heat affected portion of the join As a result of this finding, the licensee performed
penetrant testing (PT)
of all 52 incore closure assemblies in the incore instrumentation tan On instrument tube L-2, several other indications were identified but were not through wall crack An additional tube (E-7)
was found to have an indication in a weld at a tube support plat This weld defect was not on a primary pressure boundary and was ground out in accordance with licensee maintenance procedures. The closure assembly on L-2 was cut off and capped. The assembly was sent to Babcock and Wilcox (B&W) for destructive testing and analysi An analysis performed by the licensee indicated that the loss of this incore instrumentation would have no effect on the operation of the uni All other indications on L-2 that were not part of the section removed were ground out and determined to be acceptabl Preliminary reports from B&W indicate that the material had forging inclusions in the tube material and the through wall crack was the result of low stress-high cycle fatigu Additional analysis is still being performe No violations or deviations were identifie.
Inspection of Open Items (92700)(92701)
The following open items were reviewed using licensee reports, inspection, record review, and discussions with licensee personnel, as appropriate:
(Closed) LER 50-269/90-12:
Potential Overload Conditions May Result in Inadequate On-Site Emergency Power Source During a LOCA/LOOP Event Due to Design Deficienc This LER was submitted in correspondence dated August 29, 199 It identified a postulated condition that could allow the Emergency Power Source to be overloaded due to closure of the Keowee generator breakers prior to tripping of reactor coolant pumps (RCPs)
during a loss of offsite power (LOOP). Modification 52855, Breaker Logic Modification to Prevent Keowee Overload, was completed which added time delays to the Keowee generator breakers to allow for RCPs to be de-energized prior to closing. The inspectors reviewed this modification and the testing associated with the change Based on this review, this item is close.
Exit Interview (30703)
The inspection scope and findings were summarized on December 3, 1990, with those persons indicated in paragraph 1 abov The inspectors described the areas inspected and discussed in detail the inspection finding The licensee did not identify as proprietary any of the material provided to or reviewed by the inspectors during this inspectio Item Number Description/Reference Paragraph 269/90-33-01 Violation - Failure to adequately incorporate requirements for pump testing into procedure PT/1/A/0204/07, paragraph /90-33-01 Violation -
Failure to maintain the Emergency Feedwater System in accordance with OP/2/A/1106/06, paragraph 2.f.