IR 05000322/1985042
| ML20141G383 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 12/31/1985 |
| From: | Strosnider J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20141G369 | List: |
| References | |
| 50-322-85-42, IEB-85-001, IEB-85-002, IEB-85-003, IEB-85-1, IEB-85-2, IEB-85-3, NUDOCS 8601100172 | |
| Download: ML20141G383 (17) | |
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U.S. NUCLEAR REGULATORY COMMISSION Region I REPORT NO.
50-322/85-42 DOCKET NO.
50-322 LICENSE NO.
NPF-36 LICENSEE:
Long Island Lighting Company Post Office Box 618 Shoreham Nuclear Power Station Wading River, New York INSPECTION AT:
Wading River, New York INSPECTION CONDUCTED:
November 1-30, 1985 INSPECTORS:
J. A. Berry, Senior Resident Inspector E. L. Conner, Project Engineer, Section IB R
L.
hr eister, Reactor Engineer, Section IB
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APPROVED BY:
/MJ//#f s
R. Strosnider, Chief, Reector Projects Date Section 18, Division of Reactor Projects SUMMARY:
During this inspection period November 1-30, 1985 the inspactors observed maintenance / modification activities related to the neutrcn source outage and environmental qualification of electrical equipment important to safety. A special inspection was conducted the week of November 18, 1985, by a Regional Project Engineer and Reactor Engineer of activities related to modif-ications.
Additionally, the licensee discovered failure of check valves in the HPCI and RCIC systems during this period.
As a result of this inspection, two violations were identified for inadequate procedural controls and failure to adhere to approved procedures.
(See Section 8.0) One item was opened as a result of this inspection, and one item was closed.
This inspection involved 219 hours0.00253 days <br />0.0608 hours <br />3.621032e-4 weeks <br />8.33295e-5 months <br /> of inspection by the Senior Resident Inspector and a Region-based Project Engineer.
8601100172 860102
PDR ADOCM 05000322 G
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DETAILS 1.
Persons Contactedd L. Britt, Nuclear Licensing Division Manager H. Carter, Operation Engineer G. Gisonda, Supervisor, Nuclear Licensing R. Grunseich, Operational Compliance Engineer R. Gutman, Modifications Engineer W. Hunt, Outage Engineer
!.. Lewin, Outage and Modifications Divison Manager B. McCaffrey, Assistant to the Vice President-Nuclear Operations J. Scalice, Operations Division Manager J. Schraitt, Radiological Controls Division Manager C. Seaman, Quality Controls Division Manager W. Steiger, Plant Manager C. Swenson, Modification Engineer D. Terry, Maintenance Division Manager E. Youngling, Manager, Nuclear Engineering Department The inspectors also held discussions with other licensee and contractor personnel during the course of the inspection.
2.
Status of Previous Inspection Items 2.1 (0 pen) 50-322/85-39-03, Part 21 Notification-American Air Filter (Update)
NRC Inspection Report 50-322/85-39 detailed a potential deficiency with Intake Silencers supplied to Transamerica Delaval (TDI) for use on Emergency Diesel Generators.
The deficiency involved the absence of required welds on an internal part of the silencer.
During this inspection period, the licensee completed inspection of two of the three TDI Diesel Generator Intake Silencers. The first, EDG 101, was inspected on November 8, 1985. The inspector accompanied
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a Quality Controls Division inspector on the visual inspection.
Visual inspection verified that the required welds were present on both end caps.
The second diesel, EDG 102 was inspected on November
,9, 1985 by Quality Controls Division personnel.
Required welds were
found to present.
EDG 103 is scheduled to be inspected during the Bus 103 outage in mid-December.
Results of that inspection will be detailed in a future inspection repor _
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2.2 (Closed) 50-322/85-27-01 Modifications This item was opened in NRC Inspection Report 50-322/85-27 to track inspectors' review of modification activities.
The item was opened due to a weakness in the implementation of a station modification involving radiation monitors.
A special inspection by two region-based inspectors was conducted during this inspection period to review the licensee's modification program (see Section 9.0) Based on the results of this inspection, this item is closed.
3.
Review of Facility Operations
~3.1 Plant Status Summary During this inspection period, November 1-30, 1985, the licensee
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continued outage and modification activities related to the neutron source replacement and Equipment Qualification.
Installation of CRD Blade 22-35 was completed on November 1, 1985.
The blade had been replaced with a spare from onsite.
In-core alter-ations were completed on Noven,ber 2,1985, and further core alterations were suspended, in accordance with the Technical Specifications, to allow modification activities on the Reactor Building Stanc'by Ventilation System to begin.
During inspection of the HPCI turbine exhaust valve (E41*M0'/-044),
on November 4, 1985, it was discovered that the two upstream check valves had failed.
(See Section 7.0 for further details).
In addition to Equipment Qualification work during the inspection period, the licensee also conducted maintenance on the CRD Air System, Service Water System, Main Generator, Fire Detection, Diesel Generators, Main Steam Isolation Valves, and Suppression Pool Modifications.
As of the November.30, 1985 deadline for all equipment. qualification to be complete, the licensee had completed 19 of the 25 required modifications.
(See Section 10.0 for further details on Equipment Qualification).
I 3.2' Operational Safety Verification The inspector toured the control room daily to verify proper shift manning, use of and adherence to approved procedures, and compliance with Technical Specification Limiting Conditions for Operation.
Control panel instrumentation and recorder traces were observed and
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the status of annunciators was reviewed. Nuclear instrumentation and l
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reactor protection system status were examined.
Radiation monitoring instrumentation, including in plant Area Radiation monitors and effluent monitors were verified to be within allowable limits, and observed for indications of trends.
Electrical distribution panels were examined for verification of proper lineups of backup and emerg-ency electrical power sources as required by the Technical Specifica-tion.
The inspector reviewed Watch Engineer and Nuclear Station Operator logs for adequacy of review by oncoming watchstanders, and for proper entries. A periodic review of Night Orders, Maintenance Work Requests, the Technical Specification LCO Log, and other control room logs and records was made.
Shift turnovers were observed on a periodic basis.
The inspector also observed and reviewed the adequacy of access con-trols to the Main Control Room, and verified that no loitering by unauthroized personnel in the Control Room Area was permitted. The inspector observed the conduct of Shift personnel to ensure adherence to Shoreham Procedures 21.001.01, " Shift Operations" and 21.004.01,
" Main Control Room - Conduct for Personnel".
No unacceptable conditions were identified.
3.3 Plant and Site Tours The inspector conducted periodic tours of accessible areas of plant and site throughout the inspection period.
These included:
the Turbine and Reactor Buildings, the Rad Waste Building, the Control Building, the Screenwell Structure, the Fire Pump House, the Security Building, and the Colt Diesel Generator Building.
During these tours, the following specific items were evaluated:
Fire Equipment - Operability and evidence of periodic inspection
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of fire suppression equipment; Housekeeping - Maintenance of required cleanliness levels;
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Equipment Preservation - Maintenance of special precautionary measures for installed equipment, as applicable;
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QA/QC Surveillance - Pertinent activities were being surveilled on a sampling basis by qualified QA/QC personnel; Component Tagging - Implementation of appropriate equipment
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tagging for safety, equipment protection, and jurisdiction; Personnel adherence to Radiological Controlled Area rules,
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including proper Personnel frisking upon RCA exit;
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Access control to the Protected Area, including search
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activities, escorting and badging, and vehicle access control; Integrity of the Protected Area boundary.
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No unacceptable conditions were identified.
3.4 Administrative Matters 3.4.1 Design Review Committee (DRC)
The inspector attended a portion of one Design Review Committee meeting on November 27, 1985. The Design Review Committee's function is to perform a review of Design Input Packages and Design Output Packages to ensure that the packages have been thoroughly assembled and are adequate to proceed to the next step in the modification program. The committee consists of representatives of the Plant Staff, Nuclear Engineering Department, and Nuclear Operations Support Department.
No unacceptable conditions were identified.
- 4.
Licensee Reports 4.1 In Office Review of Licensee Event Reports The inspector reviewed Licensee Event Reports (LERs) submitted to the NRC to verify that details were clearly reported, including accuracy of the cause description and adequacy of corrective action.
The inspector determined whether further information was required from the licensee, whether generic implications were involved, and whether the event warranted onsite follow-up.
The following LERs were reviewed:
LER NUMBER TITLE 85-048
'B' RBSVS Initiation due to Technical error 85-049 LLRT Exceeds Allowable Tech. Spec. limits for MSIV's and Penetration X-17 85-50 RBSVS/CRAC "B" side initiation due to Technician error
- 85-51 HPCI Check Valve malfunction No unacceptable conditions were identified.
- Further discussed in Section.
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5.
Monthly Surveillance and Maintenance Observation 5.1 Surveillance Activities The inspector observed the performance of various surveillance tests to verify that; the surveillance procedure conformed to technical specification requirements, administrative approvals and tagging requirements were reviewed and approved prior to test initiation, testing was accomplished by qualified personnel, current approved procedures were used, test instrumentation was currently calibrated, limiting conditions for operation were met, test data was accurately and completely recorded, removal and restoration of affected components was properly accomplished, and tests were completed within the required Technical Specification frequency.
No unacceptable conditions were identified.
5.2 Maintenance Activities The inspector observed the conduct of various maintenance activities throughout the inspection period. During this observation, the inspector verified that: maintenance activities were conducted within the requirements of the plant's administrative procedures and technical specifications, proper radiological controls were implemented and observed, proper safety precautions were observed, and that activities which have the potential to impact plant operations are properly coordinated with the control room.
No unacceptable conditions were identified.
6.
Review and Followup of I&E Notices, Bulletins and Generic Letters 6.1 I&E Notices The inspector reviewed notices issued by the Office of Inspection and Enforcement during the inspection period.
Review was to determine; if the subject of the notice was applicable to the Shoreham Nuclear Power Station, and if followup of the licensee's action was required by the inspector.
Eight I&E Information Notices were received and reviewed by the inspector during this inspection period. One notice (85-58, Supple-ment 1) did not apply to Boiling Water Reactors. The licensee's actions on the other notices will be followed up in future inspection reports as appropriat E
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The I&E Notices received and reviewed during this inspection period were:
IE Notice 85-83:
Potential Failures of General Electric TK-2
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Test Blacks.
IE Notice 85-84:
Inadequate testing of Main Steam Isolation
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Valves IE Notice 85-85:
System Interaction Event Resulting in Reactor
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System Safety Relief Valve Opening following a Fire Protection Deluge System Malfunction IE Notice 85-86:
Lightning Strikes at Nuclear Power Generating
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Stations IE Notice 85-88:
Licensee Control of Contracted Services Pro-
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viding Training IE Notice 85-89:
Potential Loss of Solid State Instrumentation
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Following Failure of Control Room Cooling
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IE Notice 85-90:
Use of Sealing Compounds in an Operating System
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IE Notice 85-58, Supplement 1:
Failure of General Electric Type
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AK-2-25 Reactor Trip Breaker 6.2 I&E Bulletins Three I&E Bulletins were received and reviewed by the inspector during this inspection period. They were:
IE Bulletin 85-01:
Steam Binding of Auxiliary Feedwater Pumps
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IE Bulletin 85-02: Undervoltage Trip Attachments of Westinghouse
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08-50 Type Reactor Trip Breakers.
IE Bulletin 85-03: Motor Operated Valve Common Mode Failures
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During Plant Transients Due To Improper Switch SettingsBulletins 85-01 and 85-02 do not apply to the Shoreham Nuclear Power Station. The licensee's response to Bulletin 85-03 will be reviewed upon submittal of the written report required by the Bulleti.
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7.
HPCI/RCIC Check Valve Failure On Monday, November 4,1985, during a maintenance inspection of the High Pressure Coolant Injection System (HPCI) Turbine Steam Discharge Valve (E41*MOV-44), it was discovered that both upstream check valves in the steam discharge line (E41*18V-021 and E41*18V-022) had come apart. A bolt from check valve 022 was discovered in the valve guide of MOV-44.
Further investigation discovered that the flapper from valve 022 was lodged in the piping just upstream of the MOV. Maintenance personnel inspected check valve 18V-022, and discovered the flapper from check valve 021 lodged just upstream in the piping.
The plant was shutdown at the time the problem was discovered.
HPCI had last been operated on September 25, 1985.
The valves which failed are 18"-150 lb Class, Carbon Steel Swing Check Valves manufactured by the Anchor-Darling Valve Company. The valves are considered Primary Containment Isolation Valves, and are located outside the Primary Containment.
The valves are constructed such that a disc (or flapper) is attached to a hinge which rotates on a hinge pin. The hinge pin is supported in place by a hinge support piece which attaches to the valve bonnet with two capscrews.
These capscrews are approximately 1 1/2 inches in length and 3/4 inch in diameter.
The capscrews thread into the valve bonnet.
The bonnet and disc assembly are bolted to the check valve body.
Inspection indicated that both capscrews on each valve became disengaged from the bonnet, thereby allowing the disc and hinge support assembly to come free from the bonnet and move down the steam exhaust piping.
Initial inspection recovered only one of the four capscrews and one of the hinge pins and hinge pin supports.
It was suspected that the other three bolts were trapped in the steam line sparger in the suppression pool. On November 8, suppression pool level was lowered to just above the sparger, and scaffolding was erected to disassemble the sparger and look for the three missing bolts. On November 9, maintenance personnel disassembled the sparger.
The bolts were not discovered. The holes in the sparger walls are approximately 1" in diameter, and the bolt head is approximately 3/4" in diameter, so the bolts must have made their way out of the sparger and into the suppression pool itself.
Damage to, and potential failure of, HPCI and RCIC turbine exhaust swing check valves has been a subject of concern for some time.
I&E Information Notice 82-26 and General Electric Application Information Document No. 56 both addressed this subject, and recommended that certain actions be taken to minimize the possibility of valve damage and/or failure. These recom-mendations included:
1) certain starting and operating precautions, 2)
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that the check valve, exhaust line vacuum breaker, and exhaust line sparger f
be designed in accordance with the GE system design specifications, 3)
that the check valve be located as close to the containment as possible, and 4) that the turbine exhaust check valve internals should be visually inspected on a routine schedule such as at every refueling outage.
The licensee's response to the recommendations in this report were the subject
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of NRC Inspection Report 50-322/83-29.
The inspector noted in that reoort
.that the licensee's= actions provided acceptable corrective action.
In addition to the corrective action taken in response to I&E Information Notice 82-26, and GE AID No. 56, the licensee purchased new type lift check valves for the HPCI and RCIC systems, which are not subject to the damage / failures, of the swing check valves.
These check valves were scheduled to be installed at the first refueling outage.
In July of 1985, it was decided by the licensee to replace the RCIC check valves during the scheduled September source outage, instead of waiting
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The basis for this decision was that during STP-14 RCIC testing, excessive check valve slamming was noted.
The potential for RCIC check valve damage, due to this situation, caused the licensee to accelerate the schedule for RCIC valve replacement. HPCI check valve replacement was not rescheduled due to the fact that no problems were noted with the valves during STP-15 HPCI testing.
The RCIC check valves were removed from the system for replacement, and on November 13, 1985, upon disassembly in the machine shop, it was discovered that one of the two valves had failed.
The failure mechanism in this case was the shearing of three bolts which connect the disc assembly to the valve body. These valves were 8 inch Velan swing check valves.
Inspection of the material also disclosed surface cracking at the bend of the hinge a m.
The three bolts had been lockwired.
The holes drilled in the bolt head appeared to be too large, and went into
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the shaft of the bolt. Additionally, the disc had become disassembled from the hinge arm, and the nut and washer were missing.
The licensee is evaluating the HPCI check valve problem. The RCIC check valves are being replaced. As part of the investigation into the failure of the HPCI valves, the licensee is conducting an inspection of all Anchor-Darling swing check valves. The results of these inspections, and the licensee's analysis of the valve failures will be detailed in the next inspection report. Until completion of that effort, this is designated i
unresolved item 50-322/85-42-01.
8.
Residual Heat Removal Pump 'B' Suction Valve Leak On November 7, 1985, during maintenance on the RHR Pump 'B' Suction Valve (MOV318),approximately 3500-4500 gallons of Suppression Pool water was spflied into the Reactor Building.
The water spilled from the unbolted bonnet of the valve during its repair for Local Leak Rate Test Failure.
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The reason for the spill was that the suppression pool was not drained
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down far enough below the level of the line penetration. This was not obvious during tagging of the system prior to maintenance because the l
valve was tagged closed during draining of the line.
On November 1,1985, the plant's night orders, issued for the weekend of November 2-3, directed plant operators to lower suppression pool levels approximately three feet for work on the RCIC suction valve from the suppression pool. The level was lowered to -38 inches, which, while approximately 8 inches below the RCIC suction, was still about 2 inches
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above the bottom of the RHR suction.
RCIC valve maintenance was completed l
without incident. On November 5, 1985 Station Equipment Clearance Permit 85-11-38 was generated to disassemble and repair the RHR Suction Valve
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IE11*MOV-031B. As part of the tagging order for this SECP, the line between the pump suction and discharge valves was drained, however, since MOV-0318 was tagged closed during this drain, there was no indication that Suppression Pool Level was still approximately 2 inches above the bottom of the suction line. There were no instructions in the SECP, MWR or Tagging l
Order to adjust suppression pool level.
On November 7, 1985 maintenance personnel unbolted the valve bonnet and l
unseated the valve disc. At this time water began to flow from the body to bonnet joint. MOV-0318 is installed in an upside down configuration, so the valve disc had acted like a dam when the bonnet was initially r
unbolted.
It was not until the disc was unseated that water from the suppression pool began to spill into the Reactor Building.
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Efforts by the control room and maintenance personnel to correct the condition were successful approximately one hour after the leak began.
l The control room began immediate pump down of the suppression pool to -50 inches when informed of the leak.
The direct cause of this incident was the suppression pool level being too high. This is a result of inadequate procedural controls on maintenance activities involving the suppression pool. This lack of procedural controls is a violation of NRC requirements as detailed in the Notice of Violation transmitted with this report (50-322/85-42-02). As part of the investiga-tion of the incident, the inspector reviewed the SECP generated for the valve work.
This review disclosed that the SECP had not been completed in accordance with the requirements of Station Procedure SP12.011.01, Station Equipment Clearance Permits. The procedure requires that Step 8 of Section 1 be completed by the originator of the SECP. This step had not been completed on the SECP prepared for the RHR Suction Valve work. Subsequent review by the inspector determined that approximately 30 active SECP's had not been completed in accordance with the procedure. This is a violation of NRC requirements as detailed in the Notice of Violation transmitted
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with this report (50-322/85-42-03).
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O-9.
Modification Process Review
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During this report period, a special inspection was performed to review
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- selected modifications that had been installed or were in the process'of being: installed during the source-replacement outage at Shoreham.
From
- the more than fifty station modifications being performed during the outage,=the inspectors selected, the following eight (8) for review:
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' SM85-015 - Replace Minimum Flow Orifices in Rad Waste to Reduce Corrosion
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of Pipe
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SM85-066 - TDI Fuel Oil Transfer Mod to Supply Colt Fuel Oil
--SM85-065-ReplacebondenserOff-GasRadiationMonitorwithPulseType
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SM85-088 - Relocate HPCI Control Panel
- SM85-091s-HPCI Turbine S'tartup Mod to Reduce Potential of Overspeed Trip
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SM85-096 - RCIC Turbine Eihaust Check Valve Replacement
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. SM85-104 - Relocate Dryiell Temperature Elements -
9il: Identified Strong Phints
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The inspectors found that the following areas of the licensee's i
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modification program were go'od:
The approved procedures used to control station modifications
had sufficient detail and were easy to understand and use.
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'The Modification Engineer interface with the Cognizant Site
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Engineer (CSE) was very good, providing meaningful review of
Design Output Packages (DOP's) knowledgeable observations of the
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modification work, improvement in plant communications related to modifications program.'
Quality Control Department.*s, independence to select their' hold
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points _ (not preselected) andfn the use of different inspectors for package, work, and completed job QCD reviews.
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The extensive and competentspackage review by the Review of s:
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Operations Committee (ROC)',
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l 9.2 Identified Weak Points The inspectors identified _the following weak points where work on the licensee's part is needed to prevent equipment modification problems, incorrect installations, or violations of codes / standards.
Incorrect control of danger tags on RCIC check valve assembly
in the machine shop (see Detail 9.4.5 ).
Inability to locate two (2) randomly-selected completed
modification packages in the SR-2 permanent record storage (see Detail 9.8).
Inattention to program details such as approval of pressure
test data sheet by the preparer, and failure to closeout an expired fire protection permit.
In addition to the above identified weak points, the inspectors expressed a concern regarding the marked-up drawings, showing recent modifications, not being available in the TSC and the EOF prior to issuance of the permanent drawing changes on aperture cards.
.9.3 Review of Modifcation Programs The inspector reviewed the implementation of procedures and controls for Station Modifications including:
PD-NE-01 (Rev. 3) LILC0 Nuclear Organization Interim Management
Control Program for Station Modification.
NED3.01 (Rev. 3) Review and approval of Design Innut/ Output
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Packages for the Interim Station Modification Program.
SP12.010.02 (Rev. 7) Station Modification Activities.
- Modification Engineering Administrative Guidelines and
Directives.
These procedures identify the management responsibilities for modif-ications; the flow path from the proposed modification (EEAR) through the review chain to the engineering process (DIP /DOP); reviews by the design review committee, modification engineering (ME), QCD, ROC and NRB (if required); scheduling issuance of Maintenance Work Requests (MWR's), and implementation of the modification; the acceptance review by ME, an independent reviewer and QCD; return to service and docu-mentation of procedure and drawing update, adequate training and final i
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QCD inspection; final completed package retention in plant files.
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Comparison'of this program with ANSI N45.2.11 (1974), Reg. Guide 1.64,
.IS Section 6.5 and FSAR 17.2.3 regulations / commitments reveals no
.d'screpancies.
It was noted that the Emergency Modification Request,
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and/or Design Output Packages (EDOP), which eliminate or postpone certain reviews are used,~in many cases, like an expedited package not one resulting from any real emergancy.
The inspector recommended this minor issue be clarified in the procedures and in practice.
'9.4 Audit of Modification Packages
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The following modification packages we e reviewed in detail.
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9f.1 SM85-15 Replace Minimum Flow Orifices in Rad Waste to Reduce Corrosion of Pipe Anumberbfpipingleakshaveoccurredjustdownstreamof flow reducing orifices in horizontal runs of piping in
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waste collector and floor drain collector tanks.
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' modification is to reroute the recirculating piping so
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that the orifices-are in the vertical run of pipe, the use of multiple ori.fices, and the change from carbon steel to stainless steel pipe in the area of the orifices.
The inspector's review of this package and the physical pip.ing changeout in the rad waste building revealed no deficiencies.
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9.4.2 SM85-066 TDI Fuel Oil System Mods For Colt Diesel Tie-In
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The packas,e was instituted to simplify the piping tie-in of the Colt Diesel Generators at some future date. With the completion of this modification, all that will be needed to complete the piping will be a flanged spoolpiece. At the time.of the inspection, this modification had been per-formed on 2 of the 3 diesel fuel oil systems.
Several minor-deficiencies were noted:
a Pressure Test Data Sheet l'(SPF31.011.01-1 Rev. 5) was approved for the supervisor by the same person who prepared it; a fire protection permit
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for the welding was4not signed as work completed, nor had t
it been signed by the Responsible Supervisor as having cleared the area, in spite of the work hs.ving been completed and the permit having expi'ed r
t 9.4.3 SM85-088 Relocate HPCI Control Panel
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This p'ackage was initiated in order to move the turbine governor' electronics to an area of lower dose and dose rate to' ensure its survival in the event of a Design Basis Event. At first look the planned modification was extremely I
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confusing due to the number of revisions to the procedure (7) and the Engineering and Design Coordination Report (12).
An in-depth review of the package and all related change documentation revealed that the changes were required to account for field conditions (inaccessibility of cables, rebar interference, non-availability of certain tools for splicing wire). This points out the need for engineering personnel to more closely associate with field personnel during the initial phases of design work.
It also demon-
.strates the knowledge and expertise of the field personnel and the close working relationship between field and engineering once the work is commenced.
9.4.4 SM85-91, HPCI Turbine Startup Control Modification The subject modification was requested by a GE engireer in July during HPCI testing and as a result of modification made at other plants designed by GE (i.e., Limerick, Peach Bottom). The problem, according to Engineering Fvaluation and Assistance Request (EEAR)85-188 by the GE engineer, was the quick start of the turbine leads to low pump suction pressure trips or turbine overspeed trips. The fix was to:
1) adjust the ramp generator / signal converter idle voltage to keep control valves initially partially closed; and, 2)
add a hydraulic bypass, with a check valve, around the EG-R hydraulic actuator. This modification was awaiting the arrival of qualified tubing at the time of the inspection.
The inspector had no problems with the modification package noting that the CSE had prepared in advance the Station Procedure Change Notice forms for STP-15 and SP47.202.01.
9.4.5 SM85-096,'RCIC Turbine Exhaust Check Valve Replacement This package was initiated to replace the swing check valves in the RCIC Turbine Exhaust Line with a new design lift check valve. This replacement is necessary due to the poor performance which swing' check valves have historically exhibited in this service. The work was not complete at the time of the inspection. The only deficiency noted was that " Hold-off" tags were not removed from valves on the piping spool prior to its renioval from the plant.
This is still under investigation by the resident inspector and plant staff.
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9.5 Quality Assurance Review All Station Modification packages are sent to Quality Control Division (QCD, formerly Operations Quality Assurance) for review and approval.
The review is performed in accordance with Section 4.3 of Quality Assurance Procedure 3.2.
The QCD reviewer chooses and marks hold points and inspection points on the procedure. Any related comments are submitted on a comments control form and resolved prior to the procedure being issued to the field. A traveler is added by the QCD reviewer for any witness which is required by a referenced procedure (for instance a welded or bolted joint). After completion, the signed-off procedure is returned to QCD for final review and closure. Quality Assurance Procedure 3.2 includes check-lists for the reviews to aid the reviewers and document the reviews.
The inspector found the QCD reviews of the modification process acceptable.
9.6 Procedure Update Review The cognizant site engineer (CSE) is responsible for the preparation of station procedure change notices (SPCN's) to address the modifica-tion and submitting them to the responsible section head.
The modif-ication package remains open until the CSE has confirmed that all SPCN's related to the modification have been processed and the appro-priate procedures have been changed. The inspector selected two recently completed but still open modifications, SM85-083 (Alternate Compressed Air for PASS) and SM85-100 (Annunciator Windows) to con-firm SPCN's were issued.
In both cases, the SPCN's had been closed.
The inspector confirmed that the appropriate procedures, SP73.040.02, EPIP 2.9 and 2.11 for SM85-083 and SP23.121.01, ARP 1120 and 1121 for SM85-100, had been revised. The procedure revisions had been issued within one month of the modification completion date. There was some confusion about the proper location of the control room master copy of alarm procedures.
In some cases, the master and therefore the procedure changes were kept at the panels.
In other cases, the changes were put in the procedure copy at the back of the control' room. The licensee said this would be resolved.
9.7 Drawing Update Review SP12.010.02 specifies that the CSE coordinate revisions to the drawings by submitting marked-up copies to the responsible department head and to records management (SR-2) who sends -the master marked-up drawing to the control room for use until the drawing is up-dated.
The responsible department head orders the drawing revision from S&W.
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16 Drawing revision backlogs are currently running two to four months.
The inspector confirmed that control room Drawing No. MFSK-44A for SM85-066-(TDI Fuel Oil Transfer mod to Supply Colt Fuel 011) were properly marked-up (red lined). No problems were noted.
However, the inspector was concerned about the lack of up-to-date drawing information in the TSC and the EOF. At both of these locations, along with all other site locations, no attempt is made to keep the drawings (aperture cards) updated prior to issuance of permanent drawing changes.
This issue was discussed at the exit meeting with the licensee.
9.8 Record Retention Two station modification packages were chosen at random from the computer printout of completed modifications. A check was made in Record Management (SR-2) for the SM packages.
The two packages selected were 81-014:1E41*MOV-032 LRT Valve Addition, and 83-111:
Modify Fuel Pool Cooling Pumps. These two packages were shown as having been completed in January and September of 1983, respectively.
In neither case were the records available from SR-2.. The microfiche index did not list them as being in the microfilm library.
Further investigation revealed that the work had been completed in both cases.
The leak rate test connection for IE41*MOV-032 was installed using the ASME XI Repair Rework program under Station Procedure 15.001.01 and the SM had been cancelled as being unnecessary. The replacement of fuel pool cooling pump bolting under SM83-111 had been completed and the drawings and technical manual were updated.
It appears that the package was not sent for microfilming due to a clerical error.
Personnel in Modification Engineering committed to checking the SR-2 files against their computer generated list and correcting all deficiencies found.
9.9 Modification Feedback The licensee has established a feedback program to improve the modifications process.
The CSE is responsible to evaluate the design packages including any revisions, the procurement of required parts, the installation and testing, the return to service and the training of operators. Although the inspector never physically reviewed any feedback files, he was able to determine that the process is worthwhile by his conversations with licensee personnel involved with modifications.
10.
Environmental Qualification of Electrical Equipment The requirement for the environmental qualification of all electrical equipment important to safety at Nuclear Power Plants is established in 10 CFR 50.49. -The regulation established a deadline of November 30, 1985 for the completion of all qualification. Additionally, the Shoreham 5%
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Power License, NPF-36, Item 2.C.8, requires as a license condition that
" Prior to_ November 30, 1985 the licensee shall environmentally qualify all electrical equipment according to the provisions of 10CFR50.49". A discussion of the licensee's. status in the area of Environmental-Qualification was given in NRC Inspection Report 50-322/85-39.
As of the November 30, 1985 deadline the licensee had completed work on 19 of the 25 modifications required to comply with the regulation.
Of the i:
six remaining items, two had been granted extensions to the November 30, 1985 deadline by the Commission (see letter Samuel J. Chilk to John D.
Leonard, November 14,1985). These extensions were for the Raymond Actuators in the Reactor Building Standby Ventilation System and for the Hydrogen Recombiners. The remaining four modifications not completed by the deadline were:
Replacement of flow transmitters in the MSIV Leakage Control System.
- Provide a new power supply for the feeder trip circuits in the Low
Pressure _ Coolant Injection System Motor-Generator Sets.
Replace High Range Area Radiation Monitoring System In-Containment
Cable Assemblies.
Modification of a Low Range Accident Radiation Monitoring Panel.
- The licensee made notification to the NRC on November 4, 1985 that they were in non-compliance with license condition 2.C.8.
At the time of the non-compliance, all electrical equipment not environmentally qualified had been removed from the plant.
The plant was in cold shutdown at the time.
11. Unresolved Items Areas for which more information is required to determine acceptability are considered unresolved.
Unresolved items are discussed in Section 7.
12. Management Meetings At periodic intervals during the course of this inspection, meetings were held with licensee management to discuss the scope and findings of this inspection.
Based on NRC Region I review of this report, and discussions with licensee representatives, it was determined that this report does not contain information subject to 10 CFR 2.790 restrictions.
The inspectors also attended entrance and exit interviews for inspections conducted by region-based inspectors during the period.
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