IR 05000395/1994019
| ML20024J283 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 09/14/1994 |
| From: | Cantrell F, Thomas Farnholtz, Haag R, Kleinsorge W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20024J275 | List: |
| References | |
| 50-395-94-19, NUDOCS 9410130152 | |
| Download: ML20024J283 (16) | |
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pn atco UNITED STATES f
q'o, NUCLEAR REGULATORY COMMISSION
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'e REGloN 11
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o 101 MARIETTA STREET, N.W., SUITE 2900 j
ATLANTA, GEORGIA 30323-0199 rn c
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Report No.:
50-395/94-19
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Licensee:
South Carolina Electric & Gas Company Columbia, SC 29218 Docket No.:
50-395 License No.:
NPF-12 facility Name:
Virgil C. Summer Nuclear Station Inspection Conducted:
August 1-31, 1994 Inspectors:
b av b 4(
9// 9/f u If."C. Haag, Senior,84sitlent Inspector Date Signed
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W. /P. K16insorge, ef16nal Inspector Date Si ned
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Approved by:
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oyd S. Cantrell, C h i
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Date S'igned
Reactor Projects Section IB
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Division of Reactor Projects SUMMARY i
Scope:
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This routine inspection was conducted by the resident inspectors onsite in the areas of operational safety verification; followup on previous operation i
findings; maintenance observations; surveillance observations; and plant
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support activities.
Selected tours were conducted on backshift or weekends.
l These tours were conducted on August 20 and 22, 1994.
I Results:
(Summarized by SALP functional area)
Operations i
A violation was identified concerning mispositioned valves resulting in an
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unplanned waste gas decay tank release (paragraph 3.a.4).
The licensee
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installed a nitrogen-16 monitor in the "B" main steam line. The overall
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material condition of the service water system is considered good.
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i 9410130152 940914 PDR ADOCK 05000395 i
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Maintenance and Surveillance An unresolved item was identified concerning the surveillance requirements for
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the service water screen wash supply valves (paragraph 4.b.1).
An inspector followup item was identified concerning the verification of service water pond temperature and level (paragraph 4.b.2).
An improvement was noted in the
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method used to test snubbers.
Three of five main steam safety valves required adjustment.
Overall maintenance and surveillance activities were effective.
Plant Support i
The emergency response facilities are considered improved and very adequate.
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A severe weather incident illustrated the need to review plans for personnel evacuation.
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REPORT DETAILS 1.
Persons Contacted Licensee Employees W. Baehr, Manager, Health Physics
- M. Browne, Manager, Design Engineering
- L. Faltus, Acting Manager, Chemistry
- M. Fowlkes, Manager, Nuclear Licensing & Operating Experience
- S. Furstenberg, Manager, Maintenance Services
- S. Hunt, Manager, Quality Systems D. Lavigne, General Manager, Nuclear Safety
- J. Nesbitt, Manager, Technical Services
- K. Nettles, General Manager, Station Support H. O'Quinn, Manager, Nuclear Protection Services
- J. Proper, Supervisor, Nuclear Licensing & Operating Experience
- W. Stuart, Supervisor, Mechanical Design Engineering
- G. Taylor, General Manager, Nuclear Plant Operations
- R. Waselus, Manager, System & Component Engineering
- B. Williams, Manager, Operations
- G. Williams, Acting Associate Manager, Operations Other licensee employees contacted included engineers, technicians, operators, mechanics, security force members, and office personnel.
NRC Personnel
- F.
Cantrell, Section Chief, Division of Reactor Projects
- R. Haag, Senior Resident Inspector
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- T. Farnholtz, Resident Inspector i
- Attended exit interview Acronyms and initialisms used throughout this report are listed in the i
last paragraph.
2.
Plant Status and Activities a.
The plant operated at or near 90 percent power until August 4, 1994, when reactor power was allowed to begin to drift down.
The purpose of reducing power in this way was to control steam generator tube leakage and to allow steady state operation until the end of the operating cycle without coasting down. The power decrease was halted on or about August 12, 1994, when reactor power reached 80 percent and remained at that level for the remainder of the inspection period.
During the inspection period, the primary to secondary leakage in the "B" steam generator remained steady at approximately 70 gallons per day.
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Other NRC inspections or meetings:
During July 19-21 and August 15-18, 1994, a regional inspection I
was performed in the area of security (NRC Inspection Report i
No. 395/94-18).
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On August 31, 1994, Jim Hufham, Regional Emergency Preparedness
Coordinator, was onsite to tour the emergency response f
feilities and meet with the licensee personnel.
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Operations a.
Plant Operations (71707)
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(1)
Plant Tour and Observations The inspectors conducted daily inspections in the following areas: control room staffing, access, and operator behavior; operator adherence to approved
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procedures, TS, and limiting conditions for operations; i'
status of control room annunciators and instrumentation; and review of control room operator logs, operating
orders, plant deviation reports, tagout logs, equipment l
out of service log, and tags on components to verify
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compliance with approved procedures.
Routinely, the
inspectors attended the operations shift turnover l
meetings.
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The inspectors conducted weekly inspections for i
operability verification of selected ESF systems by i
valve alignment, breaker positions, condition of
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equipment or component (s), and operability of l
instrumentation and support items essential to system
actuation or performance. The diesel fuel oil
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transfer and spent fuel pool cooling systems were included in these inspections.
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Plant tours included observation of general plant /
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equipment conditions, control of activities in progress, plant housekeeping conditions / cleanliness, and missile hazards.
Reactor coolant system leak i
rates were reviewed to ensure that detected or i
suspected leakage from the system was recorded,
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investigated, and evaluated; and that appropriate
actions were taken if required.
Selected tours were
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conducted on backshifts or weekends. During one of the backshift tnurs, the inspectors accompanied the upper auxiliary building operator while the trend logs were taken. Trend logs are taken weekly and they monitor additional parameters other than those recorded on the normal logs which are taken each shift.
Recently the trend log sheets were revised
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such that each page of the log will have the entire month's data. This allows an easy comparison of the logs to previous values so that the operator can more readily identify problem areas.
(2)
ESF System Walkdown The inspectors verified the operability of an ESF system by performing a walkdown of the accessible portions of the SW system.
The inspectors confirmed that the licensee's system line-up procedures matched plant drawings and the as-built configuration. The inspectors looked for equipment conditions and items that might degrade performance (hangers and supports were operable, housekeeping, etc.). The inspectors verified that valves, including instrumentation isolation valves, were in the proper position, power was available, and valves were locked as appropriate.
The inspectors compared both local and remote position indications. The overall material condition of the SW system was good.
Several minor items such as three valves missing the new style equipment tags and one SW valve not listed on the valve lineup sheet were identified. Also, the questioned the basis for the SW screen wash header isolation valves, XVT31 A, B, and C being listed as " throttled" on the valve lineup sheet. The valves appeared to be near full open.
The inspectors noted that a recent modification which removed the screen wash pumps could have effected the desired position for these valves.
The licensee stated that they would review
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the required position for these isolation valves and correct the other identified items.
(3)
Steam Generator Tube Leakage
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The inspectors continued to monitor the primary to secondary leakrate in the "B" steam generator. At the beginning of this inspection period, the leakrate was approximately 65 to 70 gallons per day (GPD). The leakrate remained at that level throughout the entire inspection period.
Early in the inspection period, the licensee installed an N-16 monitor system on the "B" main steam line. The purpose of the N-16 monitor is to give early indication of a change in steam generator tube leakrate. The detector monitoring system is capaH e of indicating actual primary to secondary leakrate, if it is setup in accordance with the manufacturer's recommendations.
Since the licensee installed the detector system while the plant was at power, many of the manufacturer's recommendations could not be followed. As a result, the licensee is not using the detector to give
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an absolute value of actual leakrate, but rather to indicate trends in the leakrate. The inspectors have reviewed the installation and use of the N-16 detector and concluded that it does provide useful information in
addition to steam generator sampling and the plant
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condenser off-gas monitor.
(4)
Unplanned Release of Waste Gas Decay Tank i
On August 12, 1994, the licensee was performing a planned release of waste gas decay tank (WGDT) "G".
During the release it was noted that the inservice WGDT "A" was also being released as indicated by decreasing pressure in the tank. The contents of both tanks were being monitored for high radiation levels during the release and no setpoints i
were exceeded. The licensee stopped the release of both
"A" and "G" tanks and performed a valve lineup for the
waste gas system. Two valves, XVD07881-WG and XVD07883A-
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WG, were found to be open rather than closed as required by the Waste Gas Processing System Operating Procedure,
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S0P-119. These two mispositioned valves account for the simultaneous release of the "A" and "G" WGDTs.
The licensee performed an investigation to determine the circumstances behind the mispositioning of these two valves.
Based on trend logs and computer graphs, it was revealed that on May 16, 1994, the contents of WGDT "C" was pumped to WGDT "G" in preparation for a planned
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maintenance task. This transfer was to be accomplished in accordance with SOP-119 which requires that the two valves
in question, XVD07881-WG and XVD07883A-WG, be open during the transfer and closed following completion of the transfer.
The licensee believes the valves were opened for the transfer of waste gas from WGDT "C" to WGDT-."G",
but not closed as required by S0P-119. The failure to
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close valves XVD07881-WG and XVD07883A-WG when required by S0P-119 is identified as Violation 395/94-19-01, Failure
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to comply with a System Operating Procedure.
b.
Followup - Operations (92901)
(0 pen)
IFI 92-08-01: Unincorporated A0P comments awaiting A0P upgrade.
During a 1992 NRC inspection it was identified that all of the A0P comments from an earlier 1990 E0P inspection had not been resolved j
by the licensee.
Based on a review of the unincorporated A0P comments and the licensee's effort to resolve the E0P inspection items, the 1992 inspection characterized these remaining items as
" minor".
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Currently, the licensee is near completion of an A0P upgrade program.
Draft versions of the upgraded AOPs have been completed and validation of the procedures must be completed before the revised A0Ps are issued. The expected completion date is March 1, 1995. Of the five unincorporated items, the inspectors verified that two items have been resolved by previous procedure changes.
The licensee stated that the remaining three, general type, items will be resolved by the upgraded A0Ps.
Pending issuance of the upgraded A0Ps, the items identified by this IFI will be reviewed.
4.
Maintenance a.
Maintenance Observation (62703)
Station maintenance activities for the safety-related systems and components listed below were observed to ascertain that they were conducted in accordance with approved procedures, regulatory guides, and industry codes or standards and in conformance with TS.
The following items were considered during this review: that limiting conditions for operation were met while components or systems were removed from service, approvals were obtained prior to initiating the work, activities were accomplished using approved procedures and were inspected as applicable, functional testing and/or calibrations were performed prior to returning components or systems to service, activities were accomplished by qualified personnel, parts and materials used were properly certified, and radiological and fire prevention controls were implemented.
Work
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requests were reviewed to determine the status of outstanding jobs and to ensure that priority was assigned to safety-related equipment maintenance that may affect system performance. The following maintenance activities were observed:
(1)
Oil change in the "B" chiller unit compressor (PMTS P0179721).
The inspectors verified that the proper type and quantity of oil was used.
No deviations were noted.
(2)
Replacement of inboard and cutboard seals and bearings on the "B" spent fuel pool cooling pump (MWR 94T3230). This was the same maintenance which was performed on the "A" pump and consisted of installing bearings and seals of a different design than those removed.
In July, the "A" pump had experienced high vibrations following this maintenance and had to be reworked.
No such vibration conditions existed following the
"B" pump maintenance.
In addition, the suction and discharge bolted flange gaskets were replaced.
(3)
Repack the "B" service water pump (MWR 9403835). During disassembly, one of the two packing gland studs broke and required replacement.
The licensee wrote a nonconformance
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notice (#4975), replaced the broken stud, and completed repacking the pump.
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(4)
Torque check on the "A" RHR pump casing nuts (MWR I
94M3114). The licensee has noted a small amount of leakage from the area of the casing joint on this pump j
under certain conditions.
This check was done to ensure
that the nuts which hold the casing of this pump together
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were tightened to the proper torque and were not the cause of the leakage.
No detectable movement of the nuts was noted during the torque check indicating that the nuts had not loosened. The licensee will continue to monitor the
"A" RHR pump for leakage.
The observed maintenance tasks were performed in a satisfactory manner.
No discrepancies were noted.
b.
Surveillance Observation (61726)
f The inspectors observed surveillance activities of safety related systems and components listed below to ascertain that these
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activities were conducted in accordance with license requirements.
The inspectors verified that required administrative approvals were obtained prior to initiating the test, testing was accomplished by qualified personnel in accordance with an approved test procedure, test instrumentation was calibrated, and limiting conditions for operation were met.
Upon completion of the test, the inspectors verified that test results conformed with technical specification
and procedure requirements, any deficiencies identified during the
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testing were properly reviewed and resolved and the systems were properly returned to service. Specifically, the inspectors witnessed / reviewed portions of the following test activities:
(1)
Monthly SW system valve lineup verification (STP 123.001).
TS 4.7.4.a requires verification that each SW valve servicing safety-related equipment, which is not locked or sealed in it's position, be verified to be in its correct position.
In addition, TS Bases 3/4.7.4 addresses the operability of the SW system and the need for the system to perform it's function during normal and accident conditions. When the inspectors reviewed STP 123.003 it was discovered that the valves which service the SW pump motor upper and lower bearing coolers and the screen wash for the traveling screens were not included in the STP.
The licensee noted that the flows for the bearing coolers are verified to be in the acceptable range every other day by the operator when completing their required logs. The licensee viewed the flow verification as an acceptable
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means for performing the monthly valve position i
verification. The inspectors recognized that the flow i
verifications are direct indications of water being
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supplied to the bearing cooler and that the flow verifications were performed more frequently than monthly.
Therefore, it appears that the licensee is meeting the intent of the TS surveillance requirement.
Yet based on the wording in the TS, the licensee does not verify that these valves are in their required, predetermined positions. The licensee stated that they would review their method for meeting this TS requirement to ensure strict compliance with TS.
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The licensee's basis for not performing the position verification on the screen wash supply valves to the traveling screens was that screen wash is not actuated on an SI signal and not required for accident mitigation.
However, screen wash is routinely used during normal operation of the traveling screens and would be available during an acc.ident since the MOVs are powered from safety-related buses.
In addition, the traveling screens are classified as safety-related equipment. A further review of SW screen wash surveillance requirements is required to make a final determination on this issue. This item is identified as an Unresolved Item, URI 395/94-19-02, Traveling screen wash supply valves position verification.
(2)
Daily verification of service water pond temperature and level as required by TS 4.7.5.
The operator logs (0AP 106.001) monitor these parameters every shift. The
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acceptance criteria in the logs are the same values as specified in TSs, i.e., a minimum water level of 415 feet and a maximum temperature of 95'F.
The inspectors questioned the licensee on the instrument inaccuracies associated with monitoring these parameters and how these inaccuracies are accounted for to ensure that the TS limits are not exceeded.
The licensee's response was that sufficient margin exist between the TS limits and the SW pond design such that instrument inaccuracies are not required to be considered.
The inspectors reviewed the DBD and the FSAR and determined that this margin could not be quantified based on these documents. A further i
review of SW instrumentation and design calculations is required for this determination.
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Additional inspector followup of this issue is identified as IFI 395/94-19-03, Verification of service water pond temperature and level.
(3)
Functional test of RHR system snubber as required by TS 4.7.7.e (STP 803.003). The snubber was bench tested on a new test machine recently acquired by the licensee. The inspectors observed both the drag and acceleration tests which were within the specified acceptance criteria.
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method of testing snubbers using an automated test machine is an improvement over previously observed tests which utilized a spring scale to perform the functional test.
(4)
Fire service system flow test (STP 128.021). This test is done every three years to verify sufficient flow through the fire main header and to detect significant blockage.
i Evaluation of the results of this test revealed no significant areas of degradation.
(5)
Reactor building personnel access hatch test (STP 215.001A).
The test was satisfactory with no deviations noted. The inspectors did note that one of the test personnel performing the test seemed confused about the proper way in which PC's are to be removed when exiting the potentially contaminated area. The licensee will consider the use of signs or postings to assist personnel in this regard.
(6)
Solid state protection system train
"A" actuation test and slave relay test (STP 345.037).
No deviations were noted.
(7)
Main steam line code safety valve ASME Section XI test (STP 401.002). All five safety valves in the "A" main steam line were tested. Two of the safety valves lifted within one percent of their setpoint. The other three lifted within three percent and were reset to within one percent of their setpoints.
An unresolved item was identified concerning the surveillance requirements for the service water screen wash supply valves. An inspectors followup item was identified concerning the verification of service water pond temperature and level. An improvement was noted in the method used to test snubbers. Three of five main steam safety valves required adjustment.
5.
Plant Support Plant Support Activities (71750)
a.
During inspection activities and tours of the plant, the inspectors routinely observed aspects of plant support in the areas of radiological controls, physical security, and fire protection.
The level of radiological protection controls applied to work activities observed was commensurate with the difficulty and risk associated with the task. Aspects of the fire protection program that were examined included transient fire loads, fire brigade readiness, and fire watch patrols.
Effective implementation of the physical security program continued to be demonstrated during inspectors observations of:
security badge control; search and inspection of packages, personnel and vehicles; tours and compensatory posting of
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security officers; and control of protected and vital area barriers.
b.
The inspectors toured the new Emergency Operating Facility (EOF).
The E0F is located on the first floor of the nuclear training center; however, additional, permanently dedicated rooms are now being used to support the main EOF functions. The new EOF command center will no longer require setup prior to activation, since all the support equipment is permanently located / stored in the command center. Dose assessment and engineering support also have dedicated work areas that will only be used for E0F activities.
The close proximity of these work areas to the command center should enhance EOF communications. Overall, the inspectors viewed the E0F changes as a positive improvement to the licensee's emergency response capabilities.
On August 31, 1994, the Region II Emergency Response Coordinator visited the site at the request of the licensee.
The purpose of the visit was to review the emergency response facilities
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including the Emergency Operations Facility (E0F), the Joint Information Center (JIC), the Technical Support Center (TSC), and the Operations Support Center (OSC).
The specific area reviewed
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was the arrangements for accommodating the NRC initial site team i
in the event of an incident.
The Emergency Response Coordinator i
concluded that the facilities are fully adequate and functional for NRC personnel responding to an incident.
The number and location of desks assigned to the NRC and FTS-2000 telephones which would be used by NRC personnel are adequate to allow the NRC and the licensee to work together following the arrival of l
the initial site team.
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c.
On August 16, 1994, tornadoes were experienced in the general area of the site and surrounding counties.
During the day, what appeared to be a developing tornado was spotted south of the plant and moving north toward the plant.
The licensee had j
already implemented the severe weather procedure to protect the i
plant. Word spread quickly throughout the plant of the approaching storm and personnel began looking for a safe place to
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take shelter. Approximately 200 people located outside the protected area seemed confused as to where to take shelter. The shops, warehouses, and office buildings outside the protected
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area would not be considered substantial enough to withstand the forces associated with a tornado.
No other buildings outside protected area were available to shelter these outside workers.
Those personnel outside the protected area moved toward the
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security building which is also not considered a substantial building.
The General Manager, Nuclear Plant Operations, exercised his authority undar 10 CFR 50.54(x) to allow all personnel, many of which were not badged, to enter the protected area and move to the control building which is a concrete
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structure that would likely survive a tornado.
Following passage
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of the tornado the licensee took prompt actions to reestablish
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protective area access control. The inspectors verified license's actions. The inspectors reviewed this action and the licensee's security measures following passage of the tornado and concluded they were appropriate and reasonable under the circumstances.
The storm passed just west of the plant and did not cause any property damage or personnel injuries. This event highlighted the need for pre-planning personnel evacuation in the event of
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severe weather. The licensee is developing procedures that will notify personnel of approaching severe weather and evacuate them
to safe shelter.
(Additional supportive information is included in safeguard's NRC Inspection Report No. 50-395/94-18)
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6.
Review of Special Procedures for Welding and Nondestructive Examination
(55050, 57050, 57060, 57070 and 57090)
The licensee's procedures for welding and Nondestructive Examination
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(NDE) are contained in the Bechtel Power Corporation (BPC) Special i
Process Manual (SPM). The applicable codes for S/G installation is ASME B&PV Code Section III Subsections NB, NC and ND, 1971 Edition with i
Addenda though the Summer 1973 (71S73) for piping and AWS DI.1 1970-1974 for structures and D1.1 the latest for temporary welds.
To evaluate the SGRP welding program the inspectors reviewed documents
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and observed work activities as indicated below:
The inspectors reviewed the below listed documents.
Observations were compared to the applicable procedures, Sections III, V, and IX of the ASME B&PV Code and AWS Dl.l.
Documents Reviewed Identification Revision Title i
BPC WD-1
Bechtel Welding Standard Documen-2/24/94 tation.of Welds j
Bechtel Nondestructive 11/11/93 Standard Magnetic Particle Exam-ination l
Bechtel Nondestructive 4/26/93 Standard Radiographic Examination
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Specification BPC PT(SR)-ASME III
Bechtel Nondestructive 11/11/93 Standard Liquid Penetrant Examination BPC VE AWS
Bechtel Nondestructive 2/1/93 Standard Visual AWS
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Documents Reviewed Identification Revision Title BPC VE ASME III-NF
Bechtel Nondestructive 2/1/93 Standard Visual ASME III NF BPC VE ASME III
Bechtel Nondestructive 4/26/93 Standard Visual ASME III BPC EP-1
Bechtel Guidelines for 2/5/93 Locating Welds in Austenitic Stainless Steel and Ferritic Steel Components Using Chemical Etching.
BPC PHT-1
Bechtel Welding Standard Postweld 2/2/93 Heat treatment of Field Welds Collectively the above documents form part of the SPM.
Relative to the
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review of the SPM, the inspectors noted the following:
l BPC Procedure Nos BPC MT ASME, RT ASME, BPC PT(SR)-ASME III, BPC
VE ASME III-NF, and BPC VE ASME III do not address visual acuity requirements.
BPC indicated that visual acuity requirements for ASME Level I and II examiners, are contained in BPC NEPQ-2.
BPC
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NEPQ-2, paragraph 5.1 adequately address the visual acuity requirements.
It should however, be noted that paragraph 5.3 of BPC NEPQ-2 indicates that visual acuity examinations may be
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administered by (in addition to medical professionals and NDE
Level III examiners) individuals designated by the NDE Level III
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examiner. The document is silent as to qualifications of the individual designated by the NDE Level III Examiner.
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BPC Procedures MT ASME paragraph 5.3.1 d) specifies the use of a
black (ultraviolet) light meter to assure that minimum code required ultraviolet light intensity is achieved for examination.
No mention is made concerning the calibration of the ultraviolet i
light meter to assure accuracy of the light intensity readings.
ASME Code Section V, Article 7, Paragraph T-733 requires AC and
DC Yokes to be calibrated by lifting 10 and 40 pound weights respectively. BPC Procedure MT ASME, paragraph 4.5.3 c) states:
"Each weight shall be weighed with a scale from a reputable manufacturer and stenciled with the nominal weight prior to the first use..."
No reference is made to traceable standard
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weights used to verify the accuracy of the " scale from a reputable manufacturer" which is subsequently used to assure that the 10 and 40 pound weights are actually 10 and 40 pounds respec-tively.
This lack of assurance of scale accuracy, conflicts with the notion of " calibration".
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BPC Procedure MT ASME, paragraph 7.1.4, states:
"If coatings are
left on the part in the area being examined, it must be demon-strated on a case-by-case basis that indications can be detected through the maximum coating thickness applied, and approved by an
M&QS NDE Level III prior to use."
No mention is made of the Authorized Nuclear Inspectors (ANI). The lack of involvement of
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the ANI conflicts with ASME B&PV Code Section III, paragraph NB-5112 and Section V paragraph T-150 which require examination procedures to be demonstrated to the satisfaction of the ANI.
BPC Procedure PT(SR)-ASME III, paragraph 8.6.3 states in part:
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... incomplete bond between base material and cladding may
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produce similar indications which are nonrelevant to the
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detection of unacceptable discontinuities." The clarification of lack of bond (fusion) between cladding and the base material as a nonrelevant indication did not appear in the ASME B&PV Code Section III until the 1977 Edition.
BPC Procedure PT(SR)-ASME III allows Liquid Penetrant
examination at temperatures as low as 40*F.
ASME B&PV Code Section V, Article 6, Paragraph T-632.1 (d) requires a qualification demonstration for examinations below 60*F.
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Procedure PT(SR)ASME III included Qualification Records supporting temperatures down to 40*F.
The Qualification Records make no mention of the ANI.
The lack of involvement of the ANI conflicts with ASME B&PV Code Section III, paragraph NB-5112 and Section V paragraph T-150, which require examination procedures
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to be demonstrated to the satisfaction of the ANI.
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BPC Procedure PHT-1 does not address the method for identifying
and documenting areas where thermocouples are attached to components as specified in ASME B&PV Code Section III paragraphs NB-4231.2 (e) and NC-4110. The Code requires that temporary attachment welds be marked prior to removal to assure that the
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weld removal locations, after removal, are suitably _ examined.
BPC indicated that they would revise Procedure PHT-1 to address identification of temporary attachment areas for thermocouples.
BPC Procedure VE AWS does not provide illumination criteria and
angle of inspection (i.e., 30' perpendicular to the surface).
BPC indicated that they would revise Procedure VE AWS to include illumination criteria and angle of inspection criteria.
BPC Procedure VE ASME III does not address personnel
qual 4fication other than visual acuity.
BPC indicated that Procedure VE ASME III will be revised to address personnel qualification and clarification of examination performance.
The inspectors reviewed the below listed Welding Procedure Specifications (WPS)s and the associated Procedure Qualification Records (PQR)s. Observations were compared with ASME B&PV Code Section IX.
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Welding Procedure Specifications Examined
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WPS PQR P4(Gl),P3(G3)-AT-Lh(CVN) R1 1060 P3,P1,-A-Lh R0 061, 660 Pl-AT-Lh R1 695 696 Pl-A-Lh R0 9, 695 and, 696 Welding Procedure Specifications were properly qualified with supporting PQRs consistent with the Code.
Within the areas examined no violations or deviations were identified.
6.
Exit Interview The inspectors met with licensee representatives (denoted in paragraph 1) at the conclusion of the inspection on September 2,1994.
During this meeting, the inspectors summarized the scope and findings of the inspection as they are detailed in this report. The licensee representatives acknowledged the inspector's comments and did not identify as proprietary any of the materials provided to or reviewed by the inspectors during this inspection. No dissenting comments from the licensee were received.
Item Number Description and Reference 94-19-01 NOV - Failure to comply with a System Operating Procedure.
94-19-02 URI - Traveling screen wash supply valves position verification.
94-19-03 IFI - Verification of service water pond temperature and level.
7.
Acronyms and Initialisms ANI Authorized Nuclear Inspector AOP Abnormal Operating Procedure ASME American Society of Mechanical Engineers AWS American Welding Society B&PV Boiler and Pressure Vessel BPC Bechtel Power Corporation DBD Design Basis Document E0F Emergency Operating Facility E0P Emergency Operating Procedure ESF Engineered Safety Feature FSAR Final Safety Analysis Report GPD Gallons Per Day
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IFI Inspector Followup Item
LER Licensee Event Report
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MOV Motor Operated Valve MWR Maintenance Work Request N
Nitrogen NDE Nondestructive Examination NRC Nuclear Regulatory Commission i
NOV Notice of Violation NRR Nuclear Reactor Regulation 0AP Operations Administrative Procedure PMTS Preventive Maintenance Task Sheet PQR Procedure Qualitication Record PT Liquid Penetrant Testing RCS Reactor Coolant System RHR Residual Heat Removal RT Radiological Testing
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RWP Radiation Work Permit S/G Steam Generator SGRP Steam Generator Replacement Project SI Safety Injection 50P System Operating Procedure SPM Special Process Manual
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SPR Special Report STP Surveillance Test Procedure SW Service Water TS Technical Specification URI Unresolved Item VE Visual Examination WGDT Waste Gas Decay Tank WPS Welding Procedure Specification
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