IR 05000293/1980025
| ML20003A553 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 09/25/1980 |
| From: | Jerrica Johnson, Martin T, Roberts K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20003A545 | List: |
| References | |
| 50-293-80-25, NUDOCS 8102040272 | |
| Download: ML20003A553 (15) | |
Text
O U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT Region I Report No. 50-293/80-25 Docket No. 50-293 License No. DPR-35 Priority Category C
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Licensee:
Boston Edison Company 800 Bovlston Street
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B~uston, Massachusetts 02199 Facility Name:
Pilorim Nuclear Power Station, Unit 1 Inspection at:
Plymouth, Massachusetts Inspection conducted:
July 1-31, 1980 Inspectors:
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oberts Resident r ector
'date signed
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P MM J. Johnson Senior R cent Inspector date signed
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Approved by:
gV A I J I; T. MartinT Gief, Reactor Projects date signed Section No. 3, RO&NS Branch Inspection Summary:
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Inspection on July 1-31,.1980 (Report No. 50-293/80-25)
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Areas Inspected:
Rc. tine, unannounced inspection of plant onerations including an operational safety verification, surveillance and maintenance l
observation, a review of periodic reports, an in-office review of licensee I
I event reports, a review of license applications for training staff, a review of licensee actions in response to IE Bulletins, and a review of temperature effects on reactor vessel instrumentation.
The inspection involved 148 inspector-hours onsite by two resident inspectors.
Results:
One item of noncompliance was identified (Infraction - failure to follow surveillance and shift turnover procedures which specify valve positions, Paragraph 2).
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DETAILS 1.
Persons Contacted E. Cobb, Chief Operating Engineer R. Cox, Jr.. ISI Supervisor R. Machon, Assistant Station Manager C. Mathis, Methods, Training, and Compliance Group Leader J. McEachern, Securty Supervisor P. McGuire, Station Manager T. McLaughlin, Senior Compliance Engineer W. Olsen, Senior Nuclear Training Specialist P. Smith, Chief Technical Engineer R. Smith, Chemistry Supervisor D. Sukanek, Acting Chief Maintenance Engineer P. Willard, I&C Engineer The inspectors also interviewed other licensee personnel including members of the Health Physics, Operations, Security, Maintenance, Chemistry and Training staffs.
2.
Operational Safety Verification a.
Scope and Acceptance Criteria The inspector observed control room operations, reviewed applicable logs, and conducted discussions with control room operators during the month of July 1980.
The inspector verified the operability of selected emergency systems and verified proper return to service of affected components.
Tours of the screen house, reactor building, turbine building, and site perimeter were conducted to observe pl nt equipment conditions, including potential fire hazards, fluid leaks, excessive vibration, and to verify that maintenance requests had been initiated for equipment in need of maintenance.
The inspector verified by direct observation and interview that the physical
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security procedures were being implemented in accordance with the i
station security plan.
The inspector observed plant housekeeping and cleanliness conditions and verified implementation of radiation protection controls.
The inspector walked down the accessible portions of the standby liquid control system to verify operabilit These reviews and observations were conducted to verify that facility operations were in conformance with the requirements established under Technical Specifications, 10 CFR, and Administrative Procedures.
b.
Findings 1.
While conducting a daily operational safety verification on July 18, 1980, the inspector found the inboard Reactor Core Isolation Cooling (RCIC) suction valve from the torus (1301-MOV-25) indicating open in the control room.
The valve is required to be closed in accordance with the system valve line-up checklist 2.22A-4, Rev. 3.
This was brought to the attention of the watch engineer, who ordered that valve to be closed.
The inspector reviewed the Shift Turnover Checklist (OPER-38) accepted by the 0000-0800 shift on July 18, 1980 and noted that the checklist did not indicate that the valve was out of position.
Further inspection determined that 1301-MOV-25 had been left open after completing a valve operability surveillance (ST-8.5.5.1, Rev. 6) at approximately 1800 hours0.0208 days <br />0.5 hours <br />0.00298 weeks <br />6.849e-4 months <br /> on July 17, 1980.
The surveillance test specified that valve 1301-MOV-25 be returned to the closed position.
The failure to return 1301-MOV-25 to the closed position as required by the surveillance procedure and the failure to recognize that the valve was out of position during shift turnover is considered to be an item of noncompliance at the infraction level.
(50-293/80-25-01)
2.
While conducting a review of the Station Operating Log, it became apparent that the log lacked sufficient detail to adequately reflect the status of the facility.
The inspector expressed his concern to the Chief Operating Engineer who concurred and issued firm instructions that the quality of the log must improve.
The inspector has no further comments, but will monitor the quality of the log.
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Events 1.
Off Gas Radiation Monitor
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At 1913 hours0.0221 days <br />0.531 hours <br />0.00316 weeks <br />7.278965e-4 months <br /> on July 29, 1980, the NRC Duty Officer was
notified that the plant had commenced a controlled shutdown l
due to the loss of both off gas monitors (Technical
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Specification 3.8.8.5).
At 2200 hours0.0255 days <br />0.611 hours <br />0.00364 weeks <br />8.371e-4 months <br /> on July 29, 1980, l
off gas monitors were returned to service and the controlled shutdown was terminated.
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Evaluation of the event by the inspector revealed the following:
The monitors were reading lower than they were 2 days before
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I when at the same power level.
The licensee recognized this
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and isolated the monitors to inspect and clear the sample
lines of moisture.
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Moisture was not found to be the cause of the low readings.
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High condenser air in-leakage was noted.
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A grab sample of the off-gas was taken and analyzed.
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The sample flow rate was increased such that the monitor
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reading correlated with the grab sample analysir.
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During subsequent investigation of the monitor system, it
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was detemined that the two way purge valve was leaking air into the sample line, diluting the sample. The leakage was eliminated and the sample flow rate was returned to normal.
No abnormal release rates were identified in the inspector's
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i review of the stack gas monitor recorder traces for the period of the outage of the off-gas monitors.
i The inspector had no further questions at this time.
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2.
Containment Atmospheric Monitor
i The licensee discovered on July 16, 1980, that the containment i
atmospheric monitor (C-19) was not operable, and had not been operable since approximately 1500 hours0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.7075e-4 months <br /> on July 1, 1980.
The atmospheric monitor had been inadvertently isolated on July 1,1980, while perfoming maintenance on the containment isolation
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valves serving this monitor as well as the oxygen monitor. The
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facility staff did not recognize that C-19 was isolated until a
radio-chemistry technician _noted a disparity between the existing Iodine cartridge readings and past Iodine cartridge readings.
The C-19 high and low flow alams were not functioning due to a flange leak on the monitor.
The licensee imediately returned the monitor to service and corrected the flange leak.
(Refer to
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LER 80-029 for further amplification.) The licensee performed the following corrective actions:
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Caution tagged the oxygen analyzer (and also C-19) isolation
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valve control switches located in the control room to warn j
operators that this also isolates C-19.
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S Initiated review of requirements for containment atmospheric
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sampling and the practicability of the current technical specifications.
Initiated proposal for a new atmospheric monitoring system
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completely separated from the oxygen analyzer system.
Initiated proposal for installation of a status board which
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will clearly identify Technical Specification related systems which are inoperable or are having maintenance perfonned on them.
The insp1ctor reviewed drywell sump collection records and verified no abnon.al reactor coolant system leakage had occurred during the period of loss of C-19.
The inspector had no further questions at this time.
3.
Trip of Emergency Diesel Generator At 0440 hours0.00509 days <br />0.122 hours <br />7.275132e-4 weeks <br />1.6742e-4 months <br /> on July 17, 1980, the "B" Emergency Diesel Generator (EDG) tripped while conducting a routine operability surveillance.
The NRC was ininediately notified via the ENS that the facility was operating under the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> action statement of Technical Specifications. This was a conservative decision as the facility was operating under Technical Specification 3.9.B.3 which specifies a 7 day time limit.
The EDG trip was due to a conservative setpoint drift on the High Jacket Cooling Water Temperature Switch. This switch is bypassed when the EDG is given an emergency start signal.
The temperature switch was removed from the circuit, and the "B" EDG was started and run for one hour in accordance with Surveillance Procedure 8.9.1, Rev. 9.
The "B" EDG was then declared operable and the NRC notified via the ENS. The temperature switch was subsequently recalibrated, re-installed, and the "B" EDG again run for one hour.
The inspector verified that the "B" EDG was demonstrated ooerable, that no abnonnal temperature existed, and that the temperature switch was not in the trip circuit during emergency starts.
The inspector verified that the proper maintenance documents were completed and that the switch had been recalibrated.
The inspector had no further questions.
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THIS PAGE, CONTAINING 10 CFR 2.790 INFORMATION, NOT FOR PUBLIC DISCLOSURE, IS INTENTIONALLY LEFT BLANK.
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5.
Safety Relief Valve (SRV) Failure On July 25, 1980, while attempting to conduct testr that would evaluate Mark I containment modifications, the 'D' SRV failed to open when the operator initiated an open demand. The 'B'
and 'C'
SRVs were immediately tested and operated properly.
(The 'A' was not tested at this time as it is physically located next to a safety valve). The reactor was manually scrammed to conduct IE Bulletin 80-17 testing. The 'A' SRV operated correctly at 160 psig and the Reactor was brought to cold shutdown.
The drywell was entered on July 25, 1980, and gross air leakage was detected flowing through the 'D' SRV 3-way solenoid valve and minor leakage through the 'A' SRV 3-way solonoid valve. The solenoid assemblies of the 'D'
and 'A' SRVs were removed for detailed onsite inspection by a Target Rock Corporation field representative. The inspection indicated that the 'D' SRV solenoid plunger had been secured to the stem with excessive amounts of
" Loc-tite". The " Loc-tite" had spilled into the clearance between the solenoid plunger 0.D. and the bonnet I.D., thus causing the components to adhere to each other.
Inspection of the
'A'
SRV solenoid assembly indicated no abnormalities.
Following replacement of the solenoid assemblies, 'A' and 'D'
SRV's were reassembled, the reactor taken critical and pressured to 160 psig on July 26, 1980.
The safety relief valves were then tested with the following results:
'D'
Failed on first demand to open.
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'A'
Operated on first attempt.
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Operated on second attempt.
'D'
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Operated on third attempt.
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The licensee declared all SRVs operable and continued the plant start-up to conduct the automatic scram test required by IEB 80-17.
The inspector. expressed concerns about the cause of the failure of 'D' SRV to operate on the first attempt. The licensee was unable to determine the cause of this anomaly and stated that repeated satisfactory performance demonstrated operability. The cause of the 'D' SRV to open on its first attempt on July 26, 1980, is unresolved pending further investigation by the licensee and review by the NRC (50-293/80-25-02).
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6.
Surveillance Observations The inspector observed Technical Specifications required surveillance testing on the Rod Block Monitor System (Channel B) in order to verify that testing was performed in accordance with procedures, that test instrumentation was calibrated, and that limiting conditions for operation were met. The inspector also reviewed the qualifications of the personnel conducting the surveillance.
The inspector also witnessed portions of the following test activities:
" Core Spray Pump Operability and Flow Rate Test"
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"LCPI Subsystem Operability Surveillance Test"
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"A.D.S. Subsystem Manual Opening cf Relief Valves"
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No items of noncompliance were identified.
7.
Maintenance Observation The licensee approved Plant Design Change Request (PDCR) No. 80-48 on July 25, 1980. This modification involved the installation of a vacuum breaker on both the east and west side Scram Discharge Volumes (SDV's).
The inspector reviewed selected portions of the maintenance activities involved with this modification to determine whether they were conducted in accordance with approved procedures, and in accordance.vith the Technical Specifications.
The following items were reviewed:
PDCR 80-48 "CRD Scram Discharge Header Vacuum Breaker Installation"
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Safety Evaluation No. 1026, July 24, 1980.
Maintenance Request No.80-204.
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No items of noncompliance were identified.
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Review of Periodic and Special Reports Periodic reports submitted by the licensee pursuant to T.S. 6.9.A were reviewed by the inspector.
This review included the following considerations:
the report includes the information required to be reported by NRC requirements; determination whether any infomation in the report should be classified as an abnomal occurrence; and the validity of the reported infomation. The monthly reports for September, 1979 through May, 1980 were reviewed by the inspector.
No items of noncompliance were identified.
9.
Licensee Event Reports (LER's)
The inspector reviewed the following LER's to verify that the details of events were clearly reported, including the accuracy of the description of the cause and adequacy of corrective action.
The inspector detemined whether further infomation was required and whether generic implications were involved. The inspector also verified that the reporting requirements of the Technical Specifications had been met and that appropriate corrective action had been taken.
Loss of Offsite Power due to lightening 79-33
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-79-34 Reactor Pressure Permissive pressure
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switches out of calibration.
Broken drive belt on air sampler due to 79-35
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misalignment Air sampling system motor overheated 79-40
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79-44 Condensation in air sampling system due to
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steam leak 79-45 Drywell to suppression chamber d/p less
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than 1.50 psid 79-46 Organization change adding Assistant Station
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Manager 79-47 Fire brigade was reduced to less than required number
'A' EDG auxiliary trip relay would not clear 79-49
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79-51 Air sampling motor tripped due to inadvertent
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de-energization 79-52 Fire alarm due to moisture from a steam
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leak 79-53 Condenser Delta T exceeded 320F
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79-54 Shutdown transformer trip due to shorting
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of secondary phases
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LER's 79-19, 79-28 and 79-50 required followup reports. The licensee stated the following:
that the followup for LER 79-28 was included in LER No. 79-
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that a revised 79-19 report would be issued correcting the
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description of which valve was throttled, and that the analysis had not been completed for the followup on
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LER 79-50.
These items will be reviewed in a future routine inspection. The inspector also stated his concerns with the administrative format of LER's in general.
Items discussed including the following:
the use of "see attachment" on the LER form
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the date of the LER form vs. the date of transmitting cover
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letter
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the use of referencing similar LER's The licensee acknowledged the inspector's comments and stated that more emphasis would be placed in these areas. The inspector had no further questions.
10.
License Application Submittal for Licensee Training Staff Personnel On March 28, 1980, a letter was sent to all licensees requesting that SR0 license applications for staff instructors (who conduct training on systems, integrated response, transients, and simulator courses) be subnitted by August 1, 1980. The inspector questioned the licensea on July 10, 1980, concerning the status of these license applications. The licensee stated that only those staff members who currently hold SR0 licenses would conduct this training and that a request for SR0 license applications would not be subnitted. The inspector had no further questions.
11.
IE Bulletin Followup The inspector reviewed the licensee's actions and responses to the following IE Bulletins.
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a.
Loss of Non-Class 1-E Instrumentation and Control Power System Bus During Operation.
Item 2 of this Bulletin required a review of emergency procedures to ensure that specified items were included upon loss of power to each class 1-E and non-class 1-E bus. The licensee.was further requested to describe any proposed design modifications or administrative controls to be implemented resulting from these procedures and the schedule for implementing these changes.
The licensee stated in the February 5, 1980, response that there were no procedures for loss of busses Y3 or Y4 and that there were some deficiencies in the existing procedures. No plans for proposed changes were described as required by the Bulletin.
The inspector questioned the licensee concerning his plans. The licensee stated that an additional response to this Bulletin would be provided which would address the following items:
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plans for making changes to existing procedures plans for providing additional procedures for both previously
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existing equipment and newly installed alternate shutdown control panels a proposed schedule for completing these items.
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This Bulletin remains open pending a review of the licensee's additional response and the actions stated therein.
b.
IEB 80-15 Possible Loss of Emergency Notification System
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(ENS) with Loss of Offsite Power On June 30, 1980, the licensee determined that the ENS audio circuits did not have an uninterruptable power supply for emergency backup power. On July 31, 1980 the licensee completed installation of an uninterruptable power supply to the ENS audio circuits. This supply derives its power from the security system uninterruptable source.
This Bulletin remains open pending review of the licensee's forthcoming response and of other actions required in the Bulletin.
c.
IEB 80-17 Failure of 76 of 185 Control Rods to Fully-
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Insert During a Scram at a BWR
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The inspector reviewed the licensees actions in response to the Bulletin (as supplemented). This review included discussions with licensee personnel, a review of records, and witnessing of tests being performed.
The " paragraphs" listed below pertain to the paragraphs in IEB 80-17.
Paragraph 1.
The licensee's actions as described in a response
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dated July 10, 1980, were verified by a review of the station operations log and a review of inspection data sheets which describe the UT examination performed on July 5,1980. The licensee's original attempt to utilize the GE recommended UT procedure resulted in a question pertaining to gain adjustment.
These questions were resolved and the licensee reverified that no significant amount of water remained in the SOV and associated piping.
Paragraphs 3,5 The inspector met with the ISI supervisor and the
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Level III UT examiner on July 11, 1980 to witness one of the daily tests to determine whether any water remained in the SDV and associated piping.
The inspector witnessed the UT test on a mock-up piece of piping and then observed the conduct of the test in the Reactor Building on the SDV piping.
The inspector questioned the licensee on the method of documenting the test results. The licensee acknowledged the inspectors comments and stated that a revision to the method of documenting test results would be made.
The inspector had no further questions.
Paragraph 4 The licensee's actions as described in a response
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dated July 21, 1980, were verified by a review of station procedures and spot checks with selected operators.
Procedure 5.3.2, " Inability to Shutdown with Control Rods", was revised on July 16, 1980, and incorporated the requirements of IEB 80-17, items 4a through d.
Spot checks with selected operators were performed to verify preliminary training of the Brown's Ferry occurrence.
Verification j
of complete training will be performed in a future inspection.
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Paragraph 2 The licensee performed the manual and automatic
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scrams on July 25 and 26, 1980 respectively.
The results of these tests are described in the licensee's August 1, 1980 response.
The inspector reviewed a draft of Temporary Procedure 80-70 "PNPS Testing for IEB 80-17 and SRV Discharge", prior to the tests and observed satisfactory performance of the automatic scram test on July 26, 1980.
A detailed review of the data from these scram tests as described in the licensee's August 1, 1980, response will be performed in a future inspection.
Because of a problem discovered at another BWR site, the inspector was requested to verify the voltage rating on the backup scram solenoid valves.
On July 22, 1980, the inspector, accompanied by a licensee representative, inspected the installed solenoid valves and verified the following data:
Backup Scram Solenoid Valve Model No.
Nameplate Rating SV302-19A WPLBX831636 115V (d.c.)
SV302-19B WPLBX831636 115V(d.c.)
The inspector verified that an alternate vent path open to the Reactor Building atmosphere was provided within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of telephone notification on July 19, 1980, as described in the licensee's July 28, 1980 response to IEB 80-17, Supplement 2.
This Bulletin remains open pending completion of additional actions by the licensee and verification by the inspector.
12.
Reactor Vessel Level Instrumeg ii,un l
On July 29, 1980, the licensee informed the inspector of information pertaining j
to temperature effects on Reactor Vessel Level Instrumentation.
i Due to previous concerns with temperature effects on RV level instrumentation l
(expressed by both the NRC and GE) the licensee instrumented the RV Yarway l
level column reference leg (during the Jan to May 1980 refueling outage) to be able to determine actual reference leg temperature during normal full power operations.
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The licensee stated that the data received subsequent to the startup in May, 1980, showed that the average full power reference leg temperature was approximately 392?F.
A review was performed by BECO Nuclear Engineering Department personnel and also a representative from the vendor (Yarway Co.).
Station personnel were informed on July 24, 1980, of the recommendation of the NED to recalibrate the zero (of the level instruments that utilize this reference leg). This would shift the actual RV level up approximately 5 inches (hot) or approximately 3.5 inches (cold).
On July 25, 1980, the Station approved changes to the calibration procedures to effect this 5 inch change.
The licensee commenced recalibration of the applicable level instrumentation on July 25, 1980, but had not completed all instruments by the time the reactor was to be started up to conduct the automatice scram test (per IEB 80-17)onJuly 26, 1980. The licensee reviewed the existing instrumentation trip set points and the data pertaining to temperature effects on level measurements and determined that the instrumentation setpoints specified by the Technical Specifications would be met if the Yarway column reference leg temperature did not exceed approximately 315?F.
The licensee informed the inspector on July 29, 1980, that the reference leg temperature had exceeded 315?F during the reactar startup on July 26, 1980, but that the completion of recalibration of all applicable instrumentation was performed within the time limits specified in the action statement of T.S. Table 3.2.B, note 1.
The inspector questioned the licensee concerning these items and in particular about the reporting requirements of T.S. 6.9.
The licensee then stated that, upon further review of the data, the recommendation to recalibrate the level instruments was conservative and that these events did not constitute a reportable occurrence per the Technical Specifications.
The inspector acknowledged the licensee's statements and stated that this area was unresolved pending a complete review of the applicable documents and actions surrounding this event (50-293/80-25-03).
13. Unresolved Items Areas for which more information is required to determine acceptability are considered unresolved.
Unresolved items are contained in Paragraph's 5 and 12 of this report.
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14.
Exit Interview At periodic intervals during the course of the inspection, meetings were held with senior facility management to discuss the inspection scope and findings.
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