IR 05000334/1983014
| ML20024F443 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 08/19/1983 |
| From: | Lester Tripp, Troskoski W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20024F441 | List: |
| References | |
| 50-334-83-14, IEB-83-03, IEB-83-3, NUDOCS 8309090387 | |
| Download: ML20024F443 (15) | |
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DCS Nos. 830523 830707
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830310 830729
830607 U. S. NUCLEAR REGULATORY COMMISSION-REGION 1 Report No.
50-334/83-14 Docket No.
50-334
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License No.
D_PR-66 Priority Category C
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Licensee:
Duquesne Light Company
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One Oxford Center 301 Grant Street l
Pittsburgh, PA 15279 i
Facility Name:
Beaver Valley Power Station, Unit 1 Inspection at:
Shippingport, Pennsylvania
Inspection Conducted: July 12 - August 15, 1983
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Inspector:
//& 7
/ d /z/975 W. M. Troskoski, Resident Inspector
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Approved by: 4-
7!8!O
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L*. E. Tr<pp, Chief, Reactor Projects date signed i
Section No. 3A, Reactor Projects Branch 3 Inspection Sumary:
Inspection on July 12 - August 15,1983, (Inspection No. 50-334/
li3-14).
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' Areas Inspected:
Routine inspections by the resident inspector (77 hours8.912037e-4 days <br />0.0214 hours <br />1.273148e-4 weeks <br />2.92985e-5 months <br />)of:
licensee action on previous inspection findings, plant operations, refueling activities, housekeeping, fire protection, radiological controls, physical security, maintenance activities, surveillance testing, in office and onsite licensee ~ event report. followup, IE bulletins, and outage activities.
Results:- One Violation (failure to obtain an eq ipment clearance prior to removing the A river water header from service - detail 3.
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8309090387 830823 PDR ADOCK 05000334 G
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_ DETAILS 1.
Persons Contacted F. Bissert, Manager, Nuclear Support Services J. Carey, Vice President, Nuclear Division M. Coppula, Superintendent of Technical Services R. Druga, Chief Engineer K. Grada, Superintendent of Licensing and Compliance R. Hansen, Maintenance Supervisor J. Indovina, I&C Supervisor T. Jones, Manager, Nuclear Operations J. Kosmal, Radiological Operations Coordinator W. Lacey, Station Su)erintendent V. Linnenbom, Radioclemist J. Lukehart, Security Director L. Schad, Operations Supervisor E. Schnell, Radcon Supervisor
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J. Sieber, Manager, Nuclear Safety and Licensing R. Swiderski, Superintendent of Nuclear Construction N. Tonet, Manager, Nuclear Engineering T. Zyra, Plant Perfomance and Testing Supervisor The inspector also contacted other licensee employees and contractors during this inspection.
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2.
The NRC Outstanding Items (OI) List was reviewed with cognizant licensee personnel.
Items selected by the inspector were subsequently reviewed through discussions with licensee personnel, documentation review and field inspection to determine whether licensee actions specified in the OIs had been satisfactorily completed. The overall status of previously identified inspection findings were reviewed, and planned and completed licensee actions were discussed for those items reported below.
l (Closed) Unresolved Item (83-12-02):
Review Construction Department -
Nuclear (CDN) corrective action associated with the electrical clearance violation that isolated emergency bus 1DF. Since the exit meeting for NRC Inspection Report 50-334/83-12, the licensee has concluded an investigation involving the contractor and craft personnel involved.
It was detemined that a construction sub-foreman mistakenly assigned two electricians a task under Modification Work Package (MWP) 366-61, prior to electrically clearing the 1D10 4160 volt breaker. This was due in part to the amount of ongoing outage work, and a breakdown in turnover communications whereby the sub-foreman mistakenly assumed that that portion of MWP 366-61 had been cleared during the previous day shift.
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To achieve a higher visability of the plant equipment clearance process, the licensee modified construction procedure PPM 4.2, Electrical Clearances, to provide a copy of the permit with the work package. The inspector randomly reviewed several MWP's in progress and verified that this was being carried out. Additionally, to emphasize the seriousness of this incident to all craft personnel, training was conducted to reinforce the necessity of procedure compliance. This was verified by inspector review of associated training roosters and discussions with various craft personnel. This item is closed.
(Closed) Unresolved Item (83-12-03): Verify procedure compliance when CDN perfoms mechanical work under station maintenance surveillance procedures (MSPs). Mechanical preventive maintenance was performed on the No.1 diesel generator by construction personnel using several MSPs (36.XX series) without obtaining the referenced administrative approvals. This was apparently due to the CDN test engineer believing that a properly executed equipment clearance was all that is necessary prior to initiating work. The licensee's representative informed the inspector that strict procedure compliance was expected of all personnel and work groups performing safety related work at BVPS. On July 20, 1983, the inspector observed the mechanical maintenance on the No. 2 diesel generator under MWR 836298. Procedure compliance with the MSP was confimed. This item is closed.
(Closed) Unresolved Item (77-20-11):
Piston setting discrepancy. Since this item was opened, Amendment No. 49 to the BVPS Technical Specification has incorporated srrveillance requirements for snubbers on plant systems (TS 3.7.12). This technical specification specifies testing frequency, visual inspection acceptance criteria, and functional testing requirements for hydraulic and mechanical snubbers. The inspector perfomed a technical review of the following in-service inspection procedures:
4.0, ISI of Steam Generator and Reactor Coolant Pump Support Hydraulic Snubbers, Revision 5.
5.0, ISI of Hydraulic Shock and Sway Suppressors on Piping Code Class I, Revision 6.
8.0, ISI of Hydraulic Shock and Sway Suppressors on Code Class II, III and NRC Safety Related Piping, Revision 7.
15.0, ISI of Mechanical Shock Arrestors on Piping Code
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Class I, II and III, Revision 1.
Acceptance criteria specified in these ISI procedures match that of the technical specification surveillance requirement. A review of test data verified that as found snubber piston settings were being measured and those outside of the operability band are documented in an 0QC general inspection report. _ By acceptance criteria definition, this would include any piston at its end of travel.. The inspector reviewed a sampling of the 0QC general inspection report issued for the ITT Grinnel hydraulic snubbers and verified that QC rejectable items were either corrected by a Maintenance Work Request or are being tracked as an open item. This item is closed.
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l (Closed) Unresolved Item (78-13-02): Verification of snubber piston settings. Sub-section D.K-2 of the ASME Boiler and Pressure Vessel Code,Section XI, IWB-2500 requires the verification of support settings of constant and variable spring type hangers, snubbers, and shock absorbers.
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Previous inspection requirements outlined in ISI Procedures 4.0, 5.0,
and 8.0 did not address this verification in 1978. Since then, the inspector verified that Procedure 15.0, ISI Mechanical Shock Arrestors on Piping Code Class I, II and III, Revision 1, has been implemented for the PSA mechanical snubber as part of the visual inspection program
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required by Technical Specification 3.7.12, and its associated mechanical
snubber function test acceptance criteria. This item is closed.
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(Closed) Unresolved Item (83-12-01):
DLC to evaluate the effect of thick wall pipe members on seismic and flexure design input. A meeting was held on July 27, 1983, between NRC Regional management, NRR structural engineering specialists, the licensee, and their architect-engineer to cvaluate the effects of this issue on Beaver Valley Power Station, Units 1 and 2.
From those discussions, it appears that this issue has generic
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implication applicable to nuclear power plants built by Stone & Webster and other architect-engineers. Further evaluation of this problem has been tasked to NRR for resolution on a national basis. Though DLC is continuing their evaluation, no additional inspection or testing require-
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ments will be imposed upon BVPS.
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(Closed) Unresolved Item (81-28-12): Verify completion of IEB 81-01 actions for uninstalled snubbers. This bulletin required BVPS Unit 1 to )erform a visual examination and manual test of all PSA mechanical snu)bers installed on safet Maintenance Work Requests (y related systems. From a review of completed Series 826502 thru 826507), the inspector verified that the six mechanical snubbers listed in TS Table 3.7-4b, l
Safety Related Mechanical Snubbers, were tested satisfactorily during January, 1982, per CMP No. 1-75-221, PSA Mechanical Snubber Inspection
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and Testing. The results were forwarded to the NRC in a letter dated l
September 13, 1982.
Since those tests were conducted, DLC has submitted Proposed Change Request
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i No. 78 to their operating license, dated November 23, 1982, as revised in
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their additional submittal dated July 14, 1983, adding an additional 31 l
mechanical snubbers to Table 3.7-4b (due to Design Change Packages -253, l-305, and -408). Freedom of movement tests for these new snubbers were
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perfomed under Plant Modification Manual. Procedure 5.
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Plant Operations t
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General I
Inspection tours of the plant areas listed below were conducted during both day and night shifts with respect to Technical
. Specification (TS) compliance,housekeepingandcleanliness, fire protection, radiation control, physical security and plant
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protection, operational and maintenance administrative controls.
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Control Room
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Primary Auxiliary Building
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Turbine Building
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Service Building
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Main Intake Structure
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Main Steam Valve Room
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Purge Duct Room
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East / West Cable Vaults
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Emergency Diesel Generator Rooms
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Containment Building
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Penetration Areas
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Safeguards Areas
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Various Switchgear Rooms / Cable Spreading Room
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Protected Areas
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Acceptance criteria for the above areas include the following:
BVPS FSAR Appendix A, Technical Specifications (TS)
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BVPS Operating Manual, (OH), Chapter 48, Conduct of Operations
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OH 1.48.5, Section D Jumpers and Lifted Leads
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OM 1.48.6, Clearance Procedures
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OM 1.48.8, Records
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OM 1.48.9, Rules of Practice
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OM Chapter 55A, Periodic Checks - Operating Surveillance Tests
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BVPS Maintenance Manual (MM), Chapter 1, Conduct of Maintenance
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10 CFR 50.54(k), Control Room Manning Requirements
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BVPSSite/StationAdministrativeProcedures(SAP)
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BVPS Physical Security Plan (PSP)
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Inspector Judgement
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b.
Operations The inspector toured the Control Room regularly to verify compliance with NRC requirements and facility technical specifications (TS).
i Direct observations of instrumentation, recorder traces and control panels were made for items important to safety.
Included in the reviews are the rod position indicators, nuclear instrumentation systems, radiation monitors, containment pressure and temperature parameters, onsite/offsite emergency power sources, availability of reactor protection systems and proper alignment of engineered safety feature systems. Where an abnormal condition existed (such as out-of-service equipment), adherence to appropriate TS action statements were independently verified. Also, various operation
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logs and records, including completed surveillance tests, equipment l
clearance pennits in progress, status board maintenance and temporary operating procedures were reviewed on a sampling basis for compliance
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in paragraph 3a.
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During the course of the inspection, discussions were conducted with operators concerning reasons for selected annunciators and knowledge of recent changes to procedures, facility configuration and plant conditions. The inspector verified adherence to approved procedures for ongoing activities observed. Shift turnovers were witnessed and staffing requirements confimed. Except as noted below, inspector comments or questions resulting from these daily reviews were acceptably resolved by licensee personnel.
(1) Westinghouse informed DLC on June 2, 1983, of a potential generic problem whereby the k(z) LOCA limit may be violated in the 0-6 foot region of the core.
It is believed that this situation could occur after continuous plant oaeration at less than or equal to 85% themal power for a fuel aurnup of greater than 500 MWD /MTU, followed by a return to full power during the same cycle. Westinghouse is expected to provide the licensee with additional guidance on this matter. The Superintendent of Licensing and Compliance informed the inspector on August'5, 1983, that though the guidance had not yet been provided, DLC was aware of the potential and would develop administrative guidance to limit allowable operating scenerios during the upcoming Cycle 4 (plant restart is scheduled for about mid-September,1983),
(2)PreparationsfortheCycle4CoreReloadingwerereviewed during the week of August 8,1993. The inspector reviewed the following documents for technical adequacy and to verify the the pre-requisites specified in Technical Specification section 3/4.9, Refueling Operations, were met.
Westinghouse Refueling Procedure
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OST 1.47.3, Containment Integrity Checklist for Refueling.
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OST 1.49.3, Refueling Operation Pre-requisites
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All documents were in order. Additionally, on August 15, 1983, the inspector observed Control Room activities during the reload and independently verified the operability of two source range neutron flux monitors,that direct comunications were maintained between the Control Room and personnel at the refueling station.
that one RHR loop was in operation and that the other loop was l
operable, 'and the containment purge and exhaust isolation system was operable. No deficiencies were noted.
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(3) Both reactor plant river water (RW) systems were mistakenly removed from service on August 5, 1983. At the time of the event, there was no fuel in the reactor; consequently, no technical specifications were violated. The cause of the event was due to personnel error that resulted in removing a 24" expansion joint (RE-J-4) on the wrong river water header.
The B river water header had been previously removed from service (EquipmentClearanceNumber 474473) for inservice inspection and hydrostatic testing that was being conducted under BVPS OM Chapter 30, Procedure F.F.
On August 4, 1983, all but four bolts were removed from the A header expansion joint. On August 5,1983, the remaining bolts were removed and the work crew started to wedge out the expansion joint when water started to fill the valve pit (located in solid waste area of the PAB). Construction personnel imediately notified the Control Room. The inspector observed the licensee's response to this from both the Control Room and the PAB. Water from the pit was pumped into PAB sumps for trestment in the liquid waste system.
Through discussions with licensee personnel, a partial walk-down, and review of process prints, the inspector detemined that a unique design feature of the RW systems contributed to the error. After leaving their respective river water aumps, the A and B header crossed each other under the pump louse prior to running in the plant. This condition is not evident in the P&ID reviewed by personnel prior to perfoming the job. This problem had not been previously experienced during other system hydros because personnel performing the tests had been involved in the initial construction and start-up. The unavailability of this expertise and the lack of proper line identification, were the main contributors. The unplanned removal of the A river water header from service on August 5,1983, is a Violation (83-14-01) of Section A.2.2.14, Inspection, Tests and Operating Status, of the BVPS Quality Assurance Program (Appendix A to the FSAR), as implemented by QA Procedure OP-9 and OM Chapter 1.48.6, Clearance Procedures.
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Plant Security / Physical Protection Implementation of the Physical Security Plan was observed in the areas listed in paragraph 3a above with regard to the following:
Protected area barriers were not degraded;
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Isolation zones were clear;
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Persons and packages were checked prior to allowing entry
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into the Protected Area;
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Vehicles were properly searched and vehicle access to the
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t Protected Areas was in accordance with approved procedures;
- - Security access controls to Vital Areas were being maintained and that persons in Vital Areas were properly authorized; Security posts were adequately manned, equipped and security
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personnel were alert and knowledgeable regarding position requirements, and that written procedures were available; and, Adequate lighting maintained.
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The inspector identified no deficiencies.
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Radiation Controls Radiation controls, including posting of radiation areas, the conditions of step-off pads, disposal of protective clothing, completien of Radiation Work Pemits, compliance with Radiation Work Permits, personnel monitoring devices being worn, cleanliness of work areas, radiation control job coverage, area monitor operability (portable and pemanent), area monitor calibration, and personnel frisking procedures were observed on a sampling basis.
(1) A startup operator received an unexpected 1700 mrem dose
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(1780 mrem for the quarter) as determined by processing his TLD badge. The event occurred at about 1:30 p.m.,
August 4, 1983, in the High Radiation Exclusion Area of
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the Solid Waste Building after a resin transfer operation.
Work was being perfomed in one section of the exclusion
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area under a Radiation Work Pemit (RWP) with health physics coverage and radiation monitoring of the 180 mrem /hr zone, when the operator was directed to obtain valve plate information in a sectioned off room containing the resin
. waste hold tank. The tank room had a maximum background level of 350 rem /hr. No prejob survey was conducted, and the HP Technician did not accompany the operator with a radiation dose rate monitoring device, in violation of facility Technical Specifications and the RWP.
The resident inspector was informed of the event'at 3:15 p.m.
August 4, 1983. A region-based radiation protection
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specialist and supervisor were sent to the site to followup
on this event. Their findings will be documented in NRC Inspection Report Number 50-334/83-16.
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e.
Plant Housekeeping and Fire Protection
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Plant housekeeping conditions including general cleanliness conditions and control of material to prevent fire hazards were observed in areas
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listed in paragraph 3a. Maintenance of fire barriers, fire barrier
penetrations, and verification of posted fire watches in these areas was also observed. No inadequacies were observed.
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4.
IE Bulletin No. 83-03: Check Valve Failures in Raw Water Cooling Systems
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of Diesel Generators.
This bulletin was sent to all operating licensees to inform them of numerous
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incidents involving failed check valves in systems important to safety.
Technical Specification testing per Section XI of the ASME Boiler and Pressure Vessel Code and Addenda, is usually done with only forward flow through the check valves, and may not detect an internal failure, unless the valves are either disassembled or flow checked in the reverse direction.
The inspector reviewed the licensee's actions forwarded in their response dated June 7, 1983. The licensee was required to review their pump and valve inservice testing (IST) program to verify that all check valves in the cooling water flow path to the diesel generator heat exchangers were addressed. Nine valves were identified, seven of which were already
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included in the IST program. The inspector independently reviewed Operating Manual Figure No. 30-1, River Water System, and cor.fimed that
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all applicable check valves were identified.
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To confim the integrity of valve internals, the licensee was requested to examine their IST program and make appropriate modifications to provide
either a forward flow and back flow test or provide for valve disassembly
and inspection. Forward flow of valve RW-110 and RW-lll is checked on l
a monthly basis per OST 1.36.1, Diesel Generator No.1 Monthly Test.
RW-ll2 and 113 are also tested per OST 1.36.2 Rather than perform a
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back flow test on those four valves, the licensee has scheduled their
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disassembly and inspection per Corrective Maintenance Procedure 1-75-186,
Repair Mission Check Valve, 2"-48", on a five year frequency. Through discussions with the IE Technical Contact for the bulletin, the inspector determined that this inspection frequency was acceptable because RW-110 thru 113 are constructed of stainless steel and are not susceptible to nomal errosion.
Check Valves RW-57, 58, and 59, are checked with forward and reverse flow per OST 1.30.2, R.P. River Water Pump 1 A test, OST 1.30.3, RW Pump 1B Test, and OST 1.30.6, RW' Pump lC Test, on a monthly frequency.
The two remaining check valves, RW-106 and 107, which were not originally part of the IST. program, are now scheduled to be disassembled and inspected in conjunction with the ISI hydrostatic testing of the river water lines (a 40 month frequency). The inspector observed the disassembly and
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inspection of RW-106 as part of that test.
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Testing is required to be perfomed by the end of the next refueling outage that comenced after April 1,1983. The licensee scheduled these tests for the current outage, with the exception of valves RW-110, 111, 112 and 113. These four valves are scheduled to be replaced
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as part of DCP 369 during the first scheduled outage after January,
1984. Since they were installed in 1980, the scheduled change-out period falls within the acceptable five year inspection frequency.
The inspector reviewed the design concept behind DCP 369 and determined that carbon steel check valves had been initially employed prior to their 1980 failure that was due to chloride corrosion from the Ohio River. The original check valves were replaced with 4" mission check valves of a high grade stainless steel construction. This 1980 replacement was performed on a " temporary basis" without all of the material traceability
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documentation. This work should have been performed by the Maintenance
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Department under a temporary procedure that would be reviewed and approved by the Onsite Safety Committee. However, due to an error, these river water check valves were installed without such a procedure after the receipt of an Engineering Memorandum accepting the use of stainless steel in this application. During 1983,' the inspector has noted increased licensee
emphasis on procedure adherence. Under the current program, such work would be accomplished under a corrective maintenance procedure such as
CMP 1-75-186, Repair of Mission Check Valves, 2"-48", which are reviewed by the Onsite Safety Comittee and approved for use by the Station i
Superintendent. Through discussions with Station Engineering, the inspector determined that the Onsite Safety Comittee had recently
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identified this problem and requested that a safety evaluation be performed prior to plant startup. The licensee's actions on this item are satisfactory.
5.
Surveillance Activities l
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To ascertain that surveillance of safety-related systems or components is being conducted in accordance with license requirements, the inspector observed portions of selected tests to verify that:
a.- The surveillance test procedure confoms to technical specification requirements, b.
Required administrative approvals and tagouts are obtained before initiating the test.
c.
Testing is being accomplished by qualified personnel in i
accordance with an approved test procedure.
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Required test instrumentation is calibrated.
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LCOs are met, f.
The test data are accurate and complete. Selected test result data was independently reviewed to verify accuracy.
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Independently verify the system was properly returned to service.
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Test results meet technical specification requirements and test discrepancies are rectified.
i. The surveillance test was completed at the required frequency.
Portions of the following surveillance tests were witnessed:
1.
BYT 1.3 - 1.1.6, Rod Position Indicator Calibration Procedure, Observed July 18, 1983.
2.
MSP 21.23, P-4851B Steam Line Pressure. Loop 2, Protection Channel III Calibration, Revision 9, performed July 22, 1983.
3.
MSP 6.34, F-436 Reactor Coolant Flow Loop 3, Protection Channel III Calibration, perfomed July 30, 1983.
4.
MSP 21.24, P-486, 1D Steam Line Pressure Loop 2, Protection Channel IV Calibration, perfomed July 30, 1983.
The Maintenance Surveillance Procedures were being run on plant protection instrumentation to verify correct emergency response facility (ERF) tie-ins, by meter and control repaimen under the guidance of CDN test engineers. The inspector noted that a step in MSP 21.23, requiring a disablement of imputs to the Process Variable Computer System and
Safety Parameter Display System computers was not signed off as required.
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This was discussed with the cognizant test engineer who told the inspector that portions of the MSP were not currently applicable to the planned
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scope of testing. Since this deviation from the test procedure was not documented, the inspector discussed it further with CDN Management.
Though the MSPs are being used as guidance by CDN, adherence to approved procedures, whether for routine maintenance or for start-up testing, is required.. The licensee's representative acknowledged the inspector's coments. On July 30, 1983, the inspector reviewed further ERF tie-in work and verified that personnel were strictly complying with the MSPs.
Licensee corrective action is satisfactor.
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6.
Outsge Items
A.
Split Pin Re)lacement - As a result of IE Information Notice 82-29, Control Rod ) rive (CRD) Guide Tube Support Pin Failures at Westing-housePWRs,thelicenseeinitiatedastationmodification(SMRNo.
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777) to replace the CRD split pins with ones of a new design to reduce the likelihood of stress corrosion cracking (SCC). The
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redesigned split pin incorporates a different material and a parabolic radius at the shank to shoulder junction to improve structural inte-grity of the pin. A station engineering review was performed, which
detemined that the proposed modification was safety related but did not constitute a design change. A safety evaluation was performed per 10 CFR 50.59, which was reviewed and accepted by the Onsite SafetyCommittee(BV-OSC-23-83). The inspector observed portions
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of job, including the radiological controls employed when a diver made underwater repairs to the cutting apparatus during the week
of July 18, 1983. Licensee actions were satisfactory.
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B.
Emergency Response Facility (ERF) Tie-In Work - As part of the ERF
tie-in work being performed on various plant instrumentation loops, i
the licensee issued multiple equipment clearances in order to facil-itate cable pulling and temination for design change packages 296/
366. The inspector reviewed the clearance points for equipment
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clearance pemit Nos. 474402 thru 474404 and 474419 at the applicable
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instrument process racks in the Switchgear room. The inspector also reviewed various safeguard relay cabinets to verify that pulled fuses
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were properly tagged (equipment clearance 474551 and 460241). No discrepancies were noted.
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C.
Steam Generator Tube Inspection - The inspector reviewed the licensee's
eddy current inspection activities being perfomed during the current outage to meet Amendment Number 48 of the Technical Specifications L
and the ASME Boiler and Pressure Vessel Code Section XI,1974 Edition
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thru Summer 1975 Addenda.
The inspector reviewed the below listed Babcock & Wilcox Company non-destructive examination procedures for compliance with applicable requirements.
ISI 460, Eddy Current Examination of Steam Generator Tubes, Revision 5.
ISI 24, Personnel Qualification Eddy Current Examination,.
Revision 5.
ISI 80, Preventive Maintenance of Nondestructive Examination'
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i Equipment, Revision 15.
The procedures meet the requirements of ASME Section XI.
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j The requirements for recording of tube wastage equal to or greater than
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20 percent and plugging of tubes equal to or greater than 40 percent is contained in Technical Specification 4.4.5.4.
Because the accuracy of the measuring technique is approximately 10%, the inspector asked the licensee whether this was accounted for in selecting those tubes that would be plugged during this current outage. The licensee's representative indicated that he believed the 40% limit in the TS took into the account the apparent error in the eddy current measuring technique. Through discussions with the Material Engineering Branch, of NRR, the inspector confirmed that the 40% limit did in fact take
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into account measurement error.
On steam generator "C," the contractor, B&W, perfomed eddy current i
examination of 3387 tubes (100% of unplugged tubes, one tube plugged in 1982) using multifrequency eddy current probes. The examination found. ten recordable indications (. 20%). Of the ten, two tubes
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exceeded the plugging limits (3 40%) with thru wail wastage of 46
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and 49 percent. Five of the recordable indications were located 1 to 2" above the top of the tube sheet and appear associated with loose parts found on the secondary side of the steam generator.
One of these parts was 3-1/4 inches long by 1/2-7/8 inches wide which the licensee believes to be a temporary hold down lug inadvert-ently left at the time the feedwater spargers were modified.
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is wedged between tubes and will require removal of three tubes for
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retrieval. Of these three tubes which will be cut and plugged, two contain recordable indications of 39 and 22 percent, while the i
third contained no recordable indication. An additional tube with a recordable indication of 38 percent will be plugged because previous inspections did not reveal any wastage, indicating that it might be
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i degrading at a faster rate than other tubes.
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In summary,on "C" steam generator, six tubes will be plugged, of which five are recordable. Five other recordable indications with tube thinning of 24, 30, 31, 35, 35 were evaluated as being acceptable and will remain in the steam generator unplugged.
On the "B" steam generator, 705 tubes were examined. Two tubes were plugged, one of which had wall thinning of 49 percent at the 14th support plate. The other tube was plugged to allow removal of a
. loose part that is similar to the one found in
"C" steam generator.
Other loose material (weld rod, etc), was found on the secondary
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side which had affected a tube in Row 1, Column 91. A 27 percent l
wall reduction located about 18 inches above the tube sheet had occurred. The part was removed and the tube was left unplugged.
Five other recordable indications, located at either the 13th or 14th support plate, were found with wall reductions of 22, 22, 26, 29 and 36 percent. These tubes were left in service. No inspection of steam generator "A" is planned during this outage. No measurable sludge was on top of the tube sheets as detemined by eddy current and visual examinations.
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The tubes that are to be plugged, will use a 6 inch insert sleeve which is roll expanded and then tested at 680 psi. The.875 rolled plugs, inghouse steam generators as outlined in B&W Document No.
supplied by Babcock & Wilcox were qualified for installation in West 51-1143735-00. Stress analysis for these plugs is contained in B&W Document No. 32-1146560-00. The licensee plans to perform a code hydrostatic test on steam generators B and C after plugging. The inspector found all areas reviewed acceptable.
D.
Snubber Surveillance Testing - Technical Specification 3.7.12, Snubbers, requires a visual inspection of all snubbers used on safety related systems, and functional testing of a representative sample on an 18 month frequency, per the specified acceptance criteria for both hydraulic and mechanical snubbers. Additionally, this technical specification requires that a record of the service life of each snubber be maintained to verify that the indicated service life has not been exceeded or will not be exceeded prior to the next scheduled snubber service life review. During this outage, the licensee initially tested a sample of 10 Grinnell hydraulic snubbers to obtain as found base line data and then rebuilt these 10 snubbers with an ethylene propylene seal material and reinstalled them per Corrective Maintenance Procedures 1-75-48, Installation and/or Removal of ITT Grinnell Hydraulic Shock and Sway Snubbers. The inspector reviewed the work packages for snubbers CH-HSS-311,.RH-HSS-101, RC-HSS-5.
Functional testing was performed per CMPl-75-159, Operating and Maintenance Procedure for ITT Grinnell Snubber Tester, and acceptance
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criteria specified for' determining lock up and bleed rates as a function of plant location and ambient temperature were derived as required by Technical Specification 4.7.12.
The remaining Grinnell snubbers were shipped off-site to Wyle Labs for similar modification and testing. The Bergan Patterson snubbers were also shipped to Wyle Lab for work under DCP 570 for modification to NED-002 specifications.
Maintenance Surveillance Procedure 48.22, Snubber Maintenance and Inspection Administrative Procedure, provides control of the testing, inspection and overhaul of mechanical and hydraulic snubbers as specified in Technical Specification 3.7.12, and serves as a central source to maintain the required data. This MSP tracks snubber manufacturer, associated serial numbers, MWR numbers, seal replace-ment and 0-ring cure dates, as well as the dates of functional testing, visual inspection and replacement parts.
It also documents cognizant management review of.this data. The snubber sampling plan is also specified in this MSP and identifys specific snubbers in each sample lot of ten, for hydraulic snubbers, and sample lot of four for mechanical snubbers. The inspector concluded that the surveillance testing is acceptable.
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In-Office Review of Licensee Event Reports The inspector reviewed LERs submitted to the NRC:R1 office to verify
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that the details of the event were clearly reported, including the accuracy of the description of cause and adequacy of corrective action.
The inspector determined whether further information was required from the licensee, whether generic implications were indicated, and whether the event warranted onsite followup. The following LERs were reviewed:
LER: 83-19/03L 10 of 15 Main Steam Safety Valve se points outside of Technical Specification }- 1%
limit.
LER: 83-20/03L*
Interruption of normal power source to the i
1 DF Emergency Bus.
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- This item was previously reviewed in NRC Inspection Report 50-334/83-12.
Corrective actions are discussed under detail 2 of this report (Unresolved Item 83-12-02).
8.
Unresolved Items Unresolved items are matters about which more information is required to determine whether they are acceptable, items of noncompliance or deviations. No new unresolved items were identified. Followup on several previous unresolved items is discussed in Section'2.
9.
Exit Interview Meetings were held with senior. facility management periodically during the course of this inspection to discuss the inspection scope and findings. A summary of inspection findings was further discussed with the licensee at the conclusion of the report period. No written material was provided to the licensee during the course of this inspection period.
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