IR 05000269/1990021
| ML15224A711 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 07/31/1990 |
| From: | Binoy Desai, Shymlock M, Skinner P, Wert L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML15224A710 | List: |
| References | |
| 50-269-90-21, 50-270-90-21, 50-287-90-21, NUDOCS 9008270205 | |
| Download: ML15224A711 (12) | |
Text
14 REGU UNITED STATES o
NUCLEAR REGULATORY COMMISSION
REGION II
o 101 MARIETTA STREET, ATLANTA, GEORGIA 30323 Report Nos: 50-269/90-21, 50-270/90-21 and 50-287/90-21 Licensee: Duke Power Company P.O. Box 1007 Charlotte, N.C. 28201-1007 Docket Nos.:
50-269, 50-270, 50-287, 74-2 License Nos.:
DPR-38, DPR-47, DPR-55, SNM-2503 Facility Name:
Oconee Nuclear Station Inspection Conducted: June 17 - July 14, 1990 Inspectors:
k'.
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P. H. Skinner, Senior Residebt Inspector Dae'Signed L. D. Wert, Resident Inspdctor Dteigned B. B. Desai, Resident Inspector at Signe Approved b :
.Shy06k,,,Section Chief Date Signed D
vision of Reactor Projects SUMMARY Scope:
This routine, announced inspection involved inspection on-site in the areas of operations, surveillance testing, maintenance activities, spent fuel transfer cask inspection, installation and testing of modifications, and inspection of open item Results: The inspector expressed concern during the review of documentation associated with the welding examinations performed on the spent fuel transfer cask and canister. Numerous errors in interpretation of film indications, radiographic techniques used, and documentation by ENSA's Level III examiners raises serious concerns over the adequacy of the ENSA quality assurance program. Also of concern is the inspection at the ENSA facility conducted by the Duke Power Company Quality Assurance Audit personnel which failed to identify the problems discussed in this report (See paragraph 5).
(III
REPORT DETAILS 1. Persons Contacted Licensee Employees
- B. Barron, Station Manager D. Couch, Keowee Hydrostation 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 D. Hubbard, Performance Engineer E. LeGette, Compliance Engineer 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
- P. Skinner L. Wert
- B. Desai
- Attended exit interview 2. Plant Operations (71707)
a. 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 routinel Discussions were conducted with plant operations, maintenance, chemistry, health physics, instrument and 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 weekends. Some inspections were made during shift change in order to evaluate shift turnover performanc Actions observed were conducted as required by the licensee's 0II
Administrative Procedures. 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 Independent Spent Fuel Storage Facility Units 1, 2 and 3 Electrical Equipment Rooms Units 1, 2 and 3 Cable Spreading Rooms Units 1, 2 and 3 Penetration Rooms Station Yard Zone within the Protected Area Standby Shutdown Facility Units 1, 2 and 3 Spent Fuel Pool Rooms Keowee Hydro Station During the plant tours, ongoing activities, housekeeping, security, equipment status, and radiation control practices were observe Unit 1 operated at 100 percent power until power level was reduced to approximately 70 percent on July 10, 1990 as a result of a low oil level in the Reactor Coolant Pump (RCP) motor. The pump was secured at that time. On July 14, power level was reduced to approximately 20 percent and entry made into the area around the 1A1 RCP to determine the problem. No visible problem was identified. Oil was added to the motor and the pump returned to service. The licensee is monitoring the pump for further signs of oil leakage. The unit returned to 100 percent power on July 1 Unit 2 operated at 100 percent power for the entire reporting perio Unit 3 operated at 100 percent power during this report period except for several days at 85 percent when the dispatcher requested the power reduction for load follo b. SSF-1HP-426 Not Properly Closed On July 2, 1990, while performing PT/1/A/0400/07, SSF (Standby Shutdown Facility) Makeup Pump Performance Test, a control room operator noticed that the level in the LDST had decreased by about four inches from 70" to 66".
Investigation revealed that approximately 120 gallons of primary coolant had leaked past the motor operated SSF letdown isolation valve (HP-426) into the spent fuel poo The leak was immediately secured by closing valves HP-426 and HP-428 (letdown and RC makeup recirculation block), and opening their associated breakers. HP-426 had apparently not properly seated during a previous operation. The unit supervisor confirmed that the leak past HP-426 was stopped by reopening valve HP-428 and verifying
no further reduction in the LOST leve The licensee is investigating the cause of the valve being ope No violations or deviations were identifie. Surveillance Testing (61726)
Surveillance tests were reviewed by the inspectors to verify procedural and performance adequacy. 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 work. 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 The following surveillances were reviewed and witnessed in whole or in part:
IP/O/A/0203/001A BWST Level Instrument Calibration OP/O/A/1106/19 Keowee Operability Verification OP/O/A/1600/010 Operation of the SSF Diesel - Generator MP/O/A/5050/017 Diesels -
SSF -
Operational Inspection and Checks PT/O/A/290/04 Main Steam Valve Movement CP/3/A/2002/01 Primary Chemistry Sampling (Core Flood Tank portion)
No violations or deviations were identifie. Maintenance Activities (62703)
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 trade. 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 The following maintenance activities were reviewed and witnessed in whole or in part:
WR 28776C Borated Water Storage Tank Instrument Heater Failure Computer Alarm Repairs WR 55018A Borated Water Storage Tank Instrument Calibration No violations or deviations were identifie. Spent Fuel Transfer Casks Inspection (55050)
This inspection was performed in response to a request for technical assistance from the Oconee Nuclear Station (ONS) resident to help resolve a concern over a spent fuel transfer cask with code rejectable radiographs and a spent fuel canister with code rejectable radiograph indication The components in question are located at ONS while the radiographs and supporting documents are located at the licensee's corporate offices in Charlotte, North Carolin On July 10, 1990, the inspector met at the corporate offices with the cognizant design engineer and the corporate Level III Radiography examiner to review the radiographs in question, discuss the problem, and ascertain what corrective action(s) were being planne By document review and through discussion with the above individuals, the inspector ascertained that the spent fuel transfer casks and spent fuel canisters were designed by Pacific Nuclear Fuel Services Inc., (PNFS) of San Jose, California and were manufactured by Equipos Nucleares SA (ENSA)
of Spain. To assist in the evaluation of the radiographs in question, PNFS has contracted Pittsburgh Testing Laboratories to provide Level III examiner service In general, the cask is a large cylinder with flanges welded on each en Sufficient space is provided for lead shielding and a grid for placing and securing spent fue The flanges are forgings made of SA-182 type F304N stainless steel material while the cylinder is made of SA-516 GR-70 material between 1.4 to 1.5 inches thick and has an outside diameter of 79 inches and a length of 191.8 inches. The cask was designed by PNFS to be manufactured and nondestructively tested in accordance with ASME Code Section III NC Edition 1983 with 1985 winter addenda. The cask is not a Code N stamp vesse Basically, the cask has a top and a bottom weld joint, identified as weld No. 3102 and weld No. 3103 respectively. A description of weldment fabrication and radiography is as follows:
Top Weld 3102:
This weld was designed as a single full penetration V-groove, open butt joint fabricated with a combination gas tungsten arc welding process (GTAW) for the root and the shielded metal arc (SMAW) process for completing the weldment. Filler metal used was 309/309-L stainless steel wire and electrodes. The weldment was radiographed using a cobalt-60 source and a single wall panoramic technique with a film-side number 30 penetramete Problems Noted:
o Weldment thickness of 1.41 inches requires use of No. 20 film side penetrameter instead of the No. 30 penetrameter used by ENS o Sensitivity achieved was 4T instead of the 2T required by the referencing cod o Cobalt-60 source used was against code recommendation for the weld thickness,Section V paragraph T-24 o Weldment should have been radiographed prior to welding a 0.59 thick radiation shielding plate on the flange inside the cas Failure to do this necessitated adding the plate thickness to the weld joint thickness in violation of Section V, paragraphs T-266.2.1 and. o Shooting technique resulted in poor film quality, i.e., too grainy and unacceptable sensitivity per referencing cod o Reader sheet in conflict with shooting sketch in that, penetrameter location on the reader sheet is shown on the film side, whereas the attached shooting sketch shows the penetrameter to be located on the source side of the wel o Only two (2) of the three (3) penetrameters required by the shooting technique were visible, and required station markers were found to be missing in several locations. This makes traceability between the weld area and the corresponding film unreliabl o The reader sheet and radiographs had been received and approved by the manufacturer's Level III examine o The inspector noted that the licensee's vendor group had audited this area at ENSA facility in Spain and had failed to detect these discrepancie Bottom Weld No. 3103: This weld was designed as a full penetration double U-groove, that was fabricated with the SMAW process. The joint was welded on one side, back-gauged and welded out. Filler metal used was 309-L stainless steel electrodes. The weldment was radiographed using a Kv-400 X-ray machine with single wall source-side number 35 penetramete Problems Noted:
o Reader sheet on radiograph certification showed welding was performed using an automatic machine, which suggests some process other than manual SMAW as indicated by the applicable welding Procedure, No. 4639 WT 203, sheet 1 of o All indications on the subject radiograph had been interpreted by the ENSA Level II examiner as being porosity. This interpretation was in disagreement with that of three Level III examiners and this inspector who interpreted the subject indications as code acceptable linear slag inclusion At the close of this visit the inspector discussed the concerns/
problems with radiographic technique film quality and questionable interpretations by the manufacturer's Level III examiner. Also the inspector stated that because the code required radiograph on weld No. 3102 was unacceptable, an engineering evaluation would have to be performed to demonstrate by calculation that the weld in question has sufficient integrity to perform its intended function. The licensee stated that such an evaluation was being generated by PNSF and would be telefaxed to Region II for review and commen Canisters:
Discussions with the licensee's cognizant design engineer and the Level III examiner disclosed that out of the three canisters on hand, one (No. DSC-1) had been rejected because lack of fusion (LOF) indications were observed in two weld joints. The indications were identified by DPC site QA and verified by the corporate Level III examine The finding has been corroborated by the manufacturer's and the vendor's Level III examiners. These welds will be repaired by DPC pending issuance and approval of a repair procedure. The canisters were designed by Pacific Nuclear Fuel Services in accordance with ASME Code Section III NB, Class 1, Edition 1983 with 1985 Winter Addenda. These are not code stamped vessels. A total of five canisters were manufactured in Spain by Equipos Nucleares SA. The canisters are approximately 67.25 inches in diameter and are 189.75 inches long. They are made of 5/8 inch thick type 304 stainless steel materia No problems were identified with the other two canisters on han Subsequent to the inspection, the inspector reviewed the evaluation and calculations and concludes that the welds in their present condition are satisfactor. Installation and Testing of Modifications (37828) (Unit 1)
The inspectors reviewed portions of several completed Nuclear Station Modification (NSM) packages. Emphasis was placed on reviewing completed packages ensuring proper reviews pursuant to 10 CFR 50.59 and that work was completed and appropriate drawings updated in accordance with the licensee's station directive The inspector reviewed in part NSM 12759, "Modify Controls for PORV circuit"; NSM 12842, "Core Flood Penetration Test; and NSM 12401, "RVLIS Software Change."
No violations or deviations were identifie. Inspection of Open Items (92700) (90712) (92701)
The following items were reviewed using licensee reports, inspection, record review, and discussions with licensee personnel, as appropriate:
a. (Closed) IFI 50-269,270,287/89-03-05: Inaccurate As-built Drawings:
This item addressed a concern identified during the Augmented Inspection Team investigation into the ITA switchgear fire in early 1989. Several electrical equipment layout drawings (cable tray sections) did not correctly specify which electrical cables were located in several cable trays. These inaccuracies adversely affected the ability of the licensee to identify and evaluate safety-related cabling potentially affected by the fire. Numerous individual cable sheet drawings had to be examined. At that time, the licensee committed to initiating a Station Problem Report (SPR)
to mark the cable section drawings "For Information Only," and to annotate on them that the cable sheets are the controlling documen Inspection Reports 50-269, 270,287/89-05 and 90-08 discuss the licensee's corrective actions concerning related cable routing and cable separation issues. Voluntary LER 50-269/89-04:
Deviation from FSAR Cable Separation Criteria due to Design Deficiency, addressed cable issues discovered during repair efforts following the ITA switchgear fire. This LER listed extensive corrective actions which directly address the concerns of IFI 50-269,270,287/89-03-05. These corrective actions are discussed belo Based on this action this item is close b. (Closed) LER 50-269/89-04: Deviation From FSAR Cable Separation Criteria Due to Design Deficiency. This LER was submitted voluntarily and addressed a situation in which two sets of redundant Main Feeder Bus (MFB) control cables (lockout relay cables) were inappropriately routed in the same cable tray. This condition existed on each of the three units and was discovered during repairs to the ITA switchgear in January 1989. Extensive corrective actions concerning cable routing and separation issues were initiated including the following:
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The Unit 1 MFB control cables were rerouted through different cable trays. Nuclear Station Modification (NSM) 2803 was completed on Unit 3 to reroute these cables and is scheduled to be implemented on Unit 2 during the upcoming refueling outag The implementation of NSM 3-2803 was examined by the inspectors in January 1990. (See Inspection Report 50-269,270,287/89-40).
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An inspection was performed by the licensee to determine the scope of the actual cable routing versus cable routing sheets problem. A total of 116 randomly selected cables were walked down. Only two valid errors concerning cable sheets or routing were noted. It was concluded that the confidence level in the
accuracy of the cable routing sheets was good. (Inspection Report 269,270,287/90-08 contains additional details of this cable inspection).
As a result of the cable separation inconsistencies discovered after the ITA fire, a DE study was performed. The study consisted of cable separation evaluations on all cables which could potentially contribute to a loss of both MFBs or more than one Engineered Safeguards Switchgea No cable separation problems which were significant enough to require rerouting were identifie Action to label cable trays to enhance physical verification of cable routes was initiated. Cable trays in the Unit 1 and Unit 2 portions of the Auxiliary Building and Turbine Building have been labele Design Engineering has developed a program to incorporate all Unit 1 and Unit 2 cable sheets into a computer databas (Unit 3 cable sheets are already in such a database). The development of a cable tray inventory database for Unit 1 and Unit 2 has been scheduled to be completed by February 1992 (this effort is in progress). This action will also specifically address the concerns of IFI 50-269,270,287/ 89-03-05:
Inaccurate As-built drawing Based on the extensive corrective actions both completed and scheduled, this LER is close c. (Closed) LER 269/89-09 (Rev. 1): Management Deficiency Resulted in Incorrect TS Which Allowed a Single Breaker Failure to Prevent the Emergency Power Switching Logic (EPSL) from Functioning. This LER addressed a situation where an inadequate TS permitted the electrical plant to be placed in configurations which made EPSL susceptible to a single failure. The condition was discovered through the Design Basis Documentation (DBD) analysis of the EPSL. During investigation of this issue, several procedural deficiencies were discovered which could also adversely affect EPSL operability. The corrective actions include formation of a task force to review and submit necessary revisions to TS 3.7, training of personnel concerning these issues, and the development of specific procedures to remove certain electrical components/busses/breakers from service. A revision to TS 3.7 correcting the deficiencies has been approved by NRR. The TS 3.7 task force is continuing its efforts to rewrite TS completely to enhance the operators use of the TS. This effort includes reformatting the TS into the standard TS format. The licensee is dedicating significant resources toward this difficult tas OP/0/A/1107/11:
Removal and Restoration of Auxiliary Electrical System, has been issued. The use of the guidance in this procedure instead of relying on the operators knowledge of the Oconee EPSL and
auxiliary power systems should help ensure that the systems are maintained in required configuration. While the efforts of the TS 3.7 task force are still in progress, all corrective actions to specifically address the issues in this LER have been complete Since this LER was written the licensee has completed the self-initiated Technical Audit (SITA) and the DBD analysis of the EPSL systems. Based on this action, this LER is close d. (Closed) LER 270/89-04:
Unit Trip Due to Management Deficiency for Poor Housekeeping. This LER concerned a Unit 2 reactor trip caused by a falling object striking a condensate booster pump pressure switch. The object was a piece of threaded stock material about inches long and 2 inches in diameter. It fell through a small uncovered sleeve in the turbine building floor and struck the switc Only two minor discrepancies were noted following the trip. The '2A'
Main Feedwater pump stayed in "auto" after the trip, due to a faulty high speed stop limit switch which was repaired prior to restart and several minor components (not safety-related or of safety significance) stopped running after the tri During the Unit 2 End of Cycle 11 refueling outage the licensee verified through testing that the rapid transfer circuitry and associated relays were functioning properly. Additionally, a survey for open floor sleeves was conducted and all discrepancies identified were repaired to prevent a similar incident. A list was compiled by operations of all instruments and devices which could cause secondary plant accidental unit trips. Currently this list is under review by Maintenance Engineering to determine which instruments will be protected by installation of protective guard Based on these actions completed and underway, this LER is close e. (Closed) LER 270/89-06:
Improper Relay Setting, Due to Design Deficiency, Could Trip 2B Reactor Building Spray (RBS) Pump During LOCA/LOOP Event. This LER addressed a problem discovered by the licensee during the Design Basis Documentation (DBD) study of the 4.16KV electrical system. The original 2B RBS pump motor had been replaced with a spare motor in May 1980. During installation of the replacement motor, the overcurrent protective relaying device setting was not evaluated. Since the starting characteristics for the new RBS motor were different than the original, the relay settings were determined to be inadequate. The 2B RBS pump motor could have tripped during starting under Loss of Coolant Accident (LOCA)/Loss of Offsite Power (LOOP) conditions since its replacement in 198 Corrective actions include:
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DE completed an examination of all inservice and spare safety related motors to determine if the existing relay settings are appropriate. The only discrepancy noted was that the settings on the spare Low Pressure Injection (LPI) Pump motor would have to be adjusted if the spare motor was installed for use. The inspectors verified that the spare motors are located in the number six warehouse and are appropriately tagge Oconee Maintenance Procedure MP/0/A/2000/3: Motor Inspection and Maintenance, is utilized whenever a safety-related motor is replaced. Step 6.5 of this procedure has been added which requires Design Engineering to be notified to ensure relay settings are adequate for the replacement motor. It is expected that this revision will be issued by August 31, 199 This LER is similar to LER 269/87-05 which involved resetting the High Pressure Injection (HPI) Pump overcurrent relays to preclude tripping in certain undervoltage situations. NRR personnel held discussions with the licensee and closely examined these issues. It was concluded that although increasing overcurrent settings to offset undervoltage conditions is not the most desirable solution, the motors will function if called upon to operat Based on the actions taken, this item is close f. (Closed) Violation 50-269,270,287/90-12-01: Failure to Incorporate Design Basis Information into Electrical Relay Procedues. This violation addressed an issue involving control of relay settings. As part of the corrective actions for this violation the licensee is reviewing, in detail, the control process used for setting safety-related protective relay As stated in licensee correspondence dated August 23, 1989, to the NRC, an LER for each finding during the DBD program will be submitted and a signle LER supplement will be submitted at the completion of the DBD program covering final resolution of each LE Based on this review, this item is close g. (Open) Inspector Followup Item 50-269,270,287/89-12-02: Review of Actions Taken Based on Findings of ECCS Valve Functional Evaluation of February 1989. During reviews being conducted by the licensee of ECCS valves on June 27, 1990, the licensee informed the resident inspectors that the post LOCA boron dilution flowpaths for all three units did not appear to meet the single failure criteria. Motor operator valves LP-103, LP-104, and valves LP-1, LP-2, located in the reactor building, comprise in part, the primary and secondary boron dilution flowpaths. Valves LP-104, LP-1, and LP-2 are powered from Motor Control Center (MCC) XS1 and valve LP-103 is powered from the SSF motor control center XSF. A single failure of MCC XS1 would render valves LP-104, LP-1, and LP-2 inoperable, thereby making both the post LOCA boron dilution flowpaths unavailabl After further investigation, on June 28, 1990 the licensee declared the systems conditionally operable based on their capability in the event of the loss of MCC XS1 to take damage control measures within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and make one flowpath available. This action would entail installation of jumping cables to MCC XS2 located within ten feet of MCC XS The boron dilution system is designed to prevent boron from concentrating in the core, and possibly reducing decay heat removal, due to evaporation during a cold leg rupture. Following a cold leg rupture, coolant will enter the vessel through the core flood nozzles into the plenum and flow out the cold leg break. The core will continue to steam through the internal vent valves and out the cold leg break. This steaming will cause boric acid to concentrate in the core due to evaporation of the water. In this event, opening LP-103 and LP-104 or LP-1 and LP-2 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the LOCA establishes a postulated 40gpm flowpath through the core and out the hot leg. This flow ensures the required mixing and precipitation to prevent boron from accumulating on the fuel cladding. The resident inspector will continue to monitor the licensee's permanent corrective action associated with this proble. Exit Interview (30703)
The inspection scope and findings were summarized on July 13, 1990, with those persons indicated in paragraph 1 above. The inspectors described the areas inspected and discussed in detail the inspection findings. The licensee did not identify as proprietary any of the material provided to or reviewed by the inspectors during this inspection.