IR 05000321/1989010
| ML20246K774 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 07/05/1989 |
| From: | Herdt A, Menning J, Randy Musser NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20246K749 | List: |
| References | |
| 50-321-89-10, 50-366-89-10, NUDOCS 8907180261 | |
| Download: ML20246K774 (16) | |
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- NUCLEAR REGULATORY COMMISSION
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101 MARIETT A STREET, N.W.
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Report Nos.: 50-321/89-10 and~50-366/89-10-Licensee: l Georgia' Power; Company,
P.O.. Box 1295-Birmingham, AL 35201
.. Docket Nos.:- 50-321 and 50-366
i License Nos'.: DPR-57 and NPF-5
- Facility Name:. Hatch 1 and 2 Inspection Dates: Nay 27. - June 23, 1989'
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Inspection at Hatch :ite near Baxley, Georgia'
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f Inspectors:
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.JohnE. yenning, Senior}esidentInspector Date Engned
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00hf Rand &l'A. Musser, Resident' Inspector-D&te' Signed f/
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Approved by:
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Alan R. Herdt, Chief, Reactor Projects Branch 3 IIate Signed
. Division of Reactor Projects SUMMARY-Scope:
This routine inspection was conducted at the site in the areas of Operational Safety Verification, Maintenance Observations, Surveillance Testing: 0bserva-tions, Calibration, Surveillance Procedures and Records, and Review of Three Mile Island Items.
Results:
Two non-cited violations were. identified during this inspection. The first NCV w:s for failure to have a SLC system valve locked as required. The second NCV-related to an inadequate ADS LSFT. Two unresolved items were also identified.
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The first URI related to installation of non-qualified circuit breakers in safety-related applications. The second URI related to the inadvertent loss of logic power for drywell floor drain and equipment drain pumps.
No specific strengths or weaknesses of licensee' programs were identified based Lon the inspectors' findings and observations in the areas inspected.
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REPORT DETAILS 1.
Persons Contacted l
Licensee Employees
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C. Coggin, Training and Emergency Preparedness Manager
- D. Davis, Manager General Support
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J. Fitzsimmons, Nuclear Security Manager
- P. Fornel, Maintenance Manager
- 0. Fraser, Site Quality Assurance Manager
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- G. Goode, Acting Engineering Manager M. Googe, Outages and Planning Manager-
- W. Kirkley, Acting Health Physics and Chemistry Manager
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- J. Lewis, Acting Operations Manager
- C. Moore, Assistant General Manager - Plant Support H. Nix, General Manager - Nuclear Plant
- H. Sumner, Assistant General Manager - Plant Operations
- S. Tipps, Nuclear Safety and Compliance Manager Other licensee employees contacted included technicians, operators, mechanics, security force members, and office personnel.
NRC Resident Inspectors
- J. Menning
- R. Musser NRC management on site during inspection period:
L. Crocker, Project Manager, Project Directorate 11-3, NRR F. Jape, Chief, Quality Performance Section, Region II D. Matthews, Director, Project Directorate 11-3, NRR E. Merschoff, Deputy Director, Division of Reactor Safety, Region II
- Attended exit interview Acronyms and initialisms used throughout this report are listed in the last paragraph.
2.
Operational Safety Verification (71707) Units 1 and 2 The inspectors kept themselves informed on a daily basis of the overall plant status and any significant safety matters related to plant operations.
Daily discussions were held with plant management and various members of the plant operating staff. The inspectors made frequent visits to the control room.
Observations included control room manning, access control, operator professionalism and attentiveness, adherence to
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procedures, adherence to limiting conditions for operation, instrument readings, recorder traces, annunciator alarms, operability of nuclear instrumentation and reactor protection system channels, availability of power sources, and operability of the Safety Parameter Display system.
These observations also included log book entries, tags and clearances on equipment, temporary alterations in effect, ECCS system lineups, containment integrity, reactor mode switch position, conformance with technical specification safety limits, daily surveillance, plant chemistry, scram discharge volume valve positions, and rod movement controls. This inspection activity involved numerous informal discussions with operators and their supervisors.
The operability of selected safety-related systems was confirmed on essentially a weekly basis.
These confirmations involved verification of proper valve and control switch positioning, proper circuit breaker and fuse alignment, and operability of related instrumentation and support systems.
Major components were also inspected for leakage, proper lubrication, cooling water supply, and general condition.
On May 30, 1989, the inspector confirmed the operability of the Unit 1 emergency diesel ' generators.
Proper switch, valve, and breaker alignments were confirmed using Data Packages 1, 2, and 3, respectively, in procedure 34S0-R43-001-15.
On June 13, 1989, the inspector confirmed the operability of the Unit 2
"B" RHR loop.
Proper switch, breaker, and valve lineups were confirmed using Attachments 1, 2, and 3, respectively, to procedure 3450-E11-010-25.
On June 20, 1989, the operability of the Unit 1 SLC system was confirmed.
Proper switch, breaker, and valve positions were verified using Attachments 1, 2, and 3 to procedure 3450-C41-003-15.
On June 20, 1989, while walking down the Unit 1 SLC system, the inspector observed that valve 1C41-F024 was closed and valve 1041-F025 was locked closed.
These two valves function as isolation valves in the SLC pump discharge drain line. The SLC system valve lineup sheets (Attachment 3 to procedure 3450-C41-003-1S) require that IC41-F024 be locked closed and IC41-F025 be closed. This discrepancy was brought to the attention of the Unit 1 Shift Supervisor, who subsequently had the lock removed from 1C41-F025 and placed on IC41-F024.
The licensee's corrective action was confirmed by the inspector on June 21, 1989.
This failure to have the i
proper valve locked, although an isolated event, is a violation of Technical Specification 6.8.1.a.
Technical Specification 6.8.1.a requires that written procedures be implemented covering the activities recommended in Appendix A of Regulatory Guide 1.33, Revision 2, February 1978.
Section 4 of Appendix A of Regd.atr,cy Guide 1.33 recommends procedures for operation of the SLC system.
Houeve', this violation meets the criteria specified in Section V of the NRC Enforcement Policy for not issuing a Notice of Violation and, therefore, is not being cited.
This matter, identified as NCV 321/89-10-01, is considered to be closed.
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General plant tours were conducted on at least a weekly basis. Portions
- of the control building, diesel generator building, intake structure, turbine building, reactor building, and outside areas were toured.
Observations included general plant / equipment conditions, fire hazards, fire alarms, fire extinguishing equipment, emergency lighting, fire barriers, emergency equipment, control of ignition sources and flammable materials, and control of maintenance / surveillance activities in progress.
Radiation protection controls, implementation of the physical security program, housekeeping conditions / cleanliness, control of missile hazards, and instrumentation. and alarms in the main control room were also observed.
In the area of housekeeping, several discrepancies were observed by the inspector.
On June 15, 1989, while performing a tour of the ' Unit l'
Reactor Building, the inspector noted various articles lying atop the torus.
The items included a hand truck, cleaning implements, and loose lagging material.
This was brought to the attention of the Unit 1 Shift Supervisor.
The inspectors observed selected operations shift turnover briefings to confirm that all necessary information concerning the status of plant systems was being addressed.
Each briefing was conducted by the oncoming OSOS.
The inspectors noted that each OSOS discussed existing plent problems, activities that were anticipated for the shift, and any rew standing orders or management directives.
Radiological and industrial safety were generally stressed.
The STAS discussed any recent procedure revisions that impacted on the attendees.
The inspectors attended shift turnover briefings on the following dates and shifts: June 4, 1989-Day, June 4, 1989-Evening, June 7, 1989-Day, and June 20, 1989-Day.
Several safety-related equipment clearances that were active were reviewed to confirm that they were properly prepared and placed.
Involved circuit'
breakers, switches, and valves were walked down to verify that clearance tags were in place and legible and that equipment was properly positioned.
Equipment clearance program requirements are specified in licensee procedure 30AC-0PS-001-0S, " Control of Equipment Clearances and Tags." On May 31, 1989, Unit 2 equipment clearance 2-89-525 was walked down. This clearance was placed to support maintenance on the "B" Loop of the RHR system.
On June 12, 1989, Unit 1 equipment clearance 1-89-340 was walked down.
This clearance was placed to isolate Drywell Cooling Fan 1T47-B007t.-2 following detection of an electrical ground.
Implementation of the licensee's sampling program was reviewed by the inspector.
This review involved observation of sampling activities (reactor coolant and tank sampling) and chemistry surveillance.
Related records were also reviewed.
During this inspection period, the inspector monitored the following activities.
On June 13, 1989, the inspector observed the auto start functional test of the main stack Kaman system in accordance with procedure 62CI-0PS-005-05.
On June 22, 1989, the inspector observed the performance of a source check of the RBCCW and
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Service Water Radiation Monitors in accordance with proce-dure 62CI-CAL-008-0.
The licensee's deficiency control system was reviewed to verify that the system is functioning as intended.
Licensee procedure t cMGR-004-OS,
" Deficiency Control System," establishes requirements and.
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for the preparation, processing, review, and disposition or deficiency reporting documents..This procedure applies to all deficiencies affecting equipment, procedures, or personnel.
Deficiencies are reported on Deficiency Cards.
On May 31, 1989, the inspector reviewed DCs that had been generated the previous day. The inspector verified that DCs had been prepared as required by-the controlling procedure, and that several deficiencies that were noted in the Shift Supervisors' logs had been documented on DCs.
More specifically, the inspector verified that DC 1-89-2441 had been prepared to document the unanticipated tripping of the "1A" station service air compressor.
It was also noted that DCs 2-89-1566 and 1567 had been generated to document the failure of both Condensate ' Demineralized Body Feed system pumps to produce flow.
On June 12, 1989, the inspector also reviewed recently prepared DCs and verified that problems observed in the-plant had been properly documented.
The inspector observed that DC 1-89-2599 had been generated to document high readings on Reactor Building Ventilation - Exhaust Radiation Monitor 1011-K609B.
The inspector also noted that DC 1-89-2610 had been prepared to document the presence of water in the motor oil of PSW pump IP41-C001D.
Selected portions of the containment isolation lineup were reviewed to confirm that the lineup was correct. The review involved verification of proper valve positioning, verification that motor and air-operated valves were not mechanically blocked and that power was available (unless blocking or power removal was required), and inspection of piping upstream of the valves for leakage or leakage paths.
On May 31, 1989, the inspector reviewed the following Unit I containment isolation valves:
IP33-F005, IP33-F010, IP33-F013, IP70-F066, 1P70-F067, 1T48-F104, IT48-F118A,1T48-F1188, IT48-F319,1T48,F334A,1T48-F334B, IT48-F335A, and IT48-F335B. On June 12, 1989, the inspector reviewed the following Unit 2 containment isolation valves: 2E11-F024B, 2E11-F027B, 2E51-F001, 2E51-F105, 2G11-F003, 2G11-F020, 2P33-F012, 2P51-F513, 2P64-F045, 2T49-F004A and B, and 2T49-F005A and B.
During the inspection period, the inspector performed an audit of the drawings used in the E0F.
This set of plant drawings is normally maintained in the simulator building library by the licensee's Training Department but can be easily transported to the EOF when necessary.
(It should be noted that the E0F is located in the simulator building just down the hall from the library). As is the case with Control Room and TSC drawings, the E0F drawings are maintained on aperture cards.
The inspector chose 20 critical drawings at random (10 from each unit) for evaluation.
Each drawing was examined for legibility and confirmed to be the most current, approved revision.
It was concluded that drawings designated for use in the E0F are adequate and readily available for use.
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On June 7,1989, the inspector reviewed the status of main control room annunciators.
The. annunciator control logs were examined to determine which annunciators were being controlled in accordance with plant procedure 30AC-0PS-009-0S, Control Room Instrumentation.
All applicable illuminated, inoperable, problem, deactivated, and nuisance annunciators that met the criteria of 30AC-0PS-009-0S were properly documented in the annunciator control log.
At the time of the review, 30 Unit 1 and 18 Unit 2 annunciators were being controlled by the guidelines set forth in 30AC-0PS-009-0S.
The inspector walked down the front panels of each unit and verified that all applicable annunciators were properly identified.
The Unit 1 front panels had four annunciators with the cards pulled (all of which were non-safety related), while the Unit 2 front panels had eight annunciators with the cards pulled (two of which were safety-related).
It appears that the licensee's aggressive program in annunciator control, set forth during the Operational Upgrade Effort, is still being maintained.
At 1700 on June 8, 1989, Unit 2 entered LCOs for having the "B" RHR loop torus suction valve (2E11-F004B) and the "B" drywell cooling return air fan (2T47-C002B) inoperable.
The licensee had been conducting plant walkdowns in-response to NRC Bulletin 88-10 and discovered that two circuit breakers powering the valve and fan were non-qualified.
Preliminary investigation revealed that these circuit breakers had been supplied as non-qualified equipment.
The subject circuit breakers are Westinghouse Model HFB 3110ML.
The Unit 2 LCOs were cleared at approximately 0140 on June 9, 1989, following replacement of the breakers with Q circuit breakers and the completion of required operability testing.
The licensee has formed an event review team to determine how the non-qualified circuit breakers were installed in safety-related applications.
Pending completion of the licensee's event review and additional review by the inspectors, this matter will be tracked as URI 366/89-10-02.
At 0913 on June 14, 1989, Unit 2 entered Technical Specification Action Statement 3.5.2.b, which requires the unit to be in hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> due to ADS being declared inoperable.
This matter was reported to the NRC pursuant to 10 CFR 50.72 at 0940 on June 14, 1989.
While performing the annual review of the ADS LSFT (42SV-B21-003-25), the ADS system engineer determined that an inadequate functional test had been performed on ADS following the implementation of DCR 88-320.
More specifically, it was determined that the LSFT (performed 12/88) did not ensure that ADS Logic Train "B" would be powered by Division 11 power ("B" station batteries / backup power supply) following the interruption of Division I power ("A" station batteries / normal power supply). The LSFT of 12/88 only isolated Division I power from ADS Logic Train "A"; therefore, it failed to test the "B" Logic Train's ability to function (be powered by Division 11 power) with the loss of Division I power.
Following entry into the LCO, the licensee prepared special purpose proce-dure 42SP-061489-PI-1-2S, which functionally tested power transfer from Division I to Division 11 power in ADS Logic Train "B".
As observed by the inspector, the power transfer capability was proven thus indicating
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that the ADS system was operable with redundant power supplies.
The licensee exited the LC0 at 1620 on June 14, 1989.
Root cause of the event has been determined to be inadequate preparation and review of the LSFT used for the implementation of DCR 88-320.
More specifically, the licensee modified an existing LSFT rather-than develop a new LSFT or special purpose procedure to ensure proper implementation of the DCR.
To prevent recurrence, the licensee has established a policy which states that LSFTs will not be acceptable in the future as functional tests for design changes.
Additionally, the involved individuals were counselled and training on this event for the engineering staff will be provided with the issuance of a departmental directive.
Criterion XI, Test Control, of 10 CFR Part 50, Appendix B, requires that a test program be established to assure that all testing required to demon-strate that systems will perform satisfactorily in service is identified and performed.
The failure to perform an adequate functional test following the implementation of DCR 88-320 is violation of Criterion XI.
However, this violation meets the criteria specified in Section V of the NRC Enforcement Policy for not issuing a Notice of Violation and, there-fore, is not being cited. This matter, identified as NCV 366/89-10-03,is considered to be closed.
At 1020 on June 14,1989, Unit 2 entered Technical Specification Action Statement 3.4.3.1.c, requiring the unit to be in hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> due to logic power being lost for both drywell equipment drain pumps as well as both drywell floor drain pumps.
This condition was discovered (at approximately 1305) by the licensee during the performance of technical specification surveillance checks in accordance with proce-dure 34SV-SUV-019-25.
The technical specification surveillance checks require the operation of one floor drain and one. equipment drain pump every four hours so that the floor drain and equipment drain leakage rates can be calculated.
Upon discovering the inoperability of the floor drain and equipment drain pumps, the operators determined that logic power for the pumps had been lost due to MCC 2R24-5017 being tagged out for i
maintenance (cleaning).
The Shift Supervisor promptly instructed maintenance personnel to stop work so the the MCC could be reenergized.
The MCC was reenergized at 1353.
Five minutes later at 1358, the LC0 was terminated when operability of the pumps was proven.
The licensee is currently reviewing this event.
Pending completion of the licensee's review and additional review by the insna-tors, this matter will be tracked as URI 366/89-10-04.
Within the areas ir.spected, two non-cited violations and two URIs were identified.
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3.
Maintenance Observations (62703) Unit 1 During the report period, the inspectors observed selected maintenance activities.
The observations included a review of the work documents for adequacy, adherence to procedure, proper tagouts, adherence to technical specifications, radiological controls, observation of all or part of the
- actual work and/or retesting in progress, specified retest requirements,
'and adherence. to the appropriate quality controls.
The primary maintenance observations during this month are summarized below:
Maintenance Activity Date.
a.
Troubleshooting of the IP51-C001A 05/30/89 H.P. Service Air Compressor b.
Repair and Calibration of the 1B SRM 06/07/89 (1C51-K606B) per MWO 1-89-2679 and procedure 57CP-C51-001-1 c.
Relocation of the LPCI Inverter 06/09/89 Disconnect Switch, 1R26-M027 in accordance with MWO 1-89-1576, procedure 42SP-032389-QC-1-15, and DCR 88-146.
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Repair and Calibration of Recorder 06/22/89 1T47-R611 in accordance with MW0 1-89-2818 and procedure 57CP-CAL-162-15 No violations or deviations were identified.
4.
Surveillance Testing Observations (61726) Unit 1 The inspectors observed the performance of selected surveillance.
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observation included a review of the procedure for technical adequacy, conformance to technical specifications, verification of test instrument calibration, observation of all or part of the actual surveillance, removal from service and return to service of the system or components affected, and review of the data for acceptability based upon the acceptance criteria. The primary surveillance testing observations during this month are summarized below:
Surveillance Testing Activity Date a.
Condenser Vacuum Instrument FT&C 06/02/89 per procedure 575V-B21-005-15 i
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HPCI Operability due to RCIC being 06/13/89 i
out service for preventive maintenance
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l per procedure 34SV-E41-002-1S.
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Surveillance Testing Activity Date'
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P.HR Valve 0perability and Quarterly 06/16/89 IST-per procedure 34SV-E11-002-IS
- On June 2, 1989, while-observing a Condenser Vacuum Instrument FT&C in
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accordance with procedure. 57SV-821-005-IS, the-inspector noted _ labeling
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discrepancies with eight. valves manipulated during the surveillance. The test valvess and isolation valves for:each of the. four condenser vacuum
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switches were not labeled as. specified in the procedure.
Specifically, F
. the vacuum. switch isolation valves were labeled as IB21-N056A,.IV-1; IB21-N056B,.IV-2; IB21-N0560. IV-3; and'1821-N0560,;IV-4; 'while,the procedure specified 1821-N056A-IV-1, 1B21-N056B-IV-1,-1821-N056C-IV-1, and
.1B21-N056D-IV-1, respectively.
Likewise, the vacuum switch-test valves.
Lwere labeled:as IB21-N056A, TV-1; IB21-N056B, TV-2;. 1821-N056C, TV-3; and 1821-N056D, TV-4; while the procedure 'specified 1821-N056A-TV-1,.
1821-N056B-TV-1, 1821-N056C-TV-1,-and 1B21-N056D-TV-1, respectively.
The inspector was informed. by the I&C Superintendent that the labeling.
discrepancies would be corrected.
-While.following up on the' surveillance (Condenser Vacuum Instrument FT&C),
the inspector discovered that the most current version of proce-
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dure 575V-B21-005-1S was not used when performed on June: 2, 1989.
(Revision 1, led 1, was used in lieu of Revision 1, Ed 2). The inspector brought this to the attention of the I&C Superintendent, who. initiated a plant' deficiency card.
The inspector did,1however, verify.that the changes made in the new revision were only editorial in nature, and thus did not affect the the technical aspects of the surveillance.
On June 16, 1989, while observing the RHR Valve Operability surveillance in accordance with 345V-E11-002-15, 'the inspector noted some minor procedural discrepancies.
Specifically, the procedure instructs the operator to "take" a valve specific control switch to the appropriate position for a " seal-in" type valve or for a "non seal-in" type valve to.
"take and hold" the specific control switch until the valve reaches the desired position.
In several cases involving "non-seal-in" type valves (step 7.2.1.1 - valve 1E11-F040, step 7.2.2.13 - valve 1E11-F091B, step 7.2.2.57 - valve 1E11-F003B, and step 7.3.1.45 - valve 1E11-F003A),
the procedure only instructs the operator to "take" the valve control switch to open/ closed in lieu of ' instructing the operator to "take and hold" the control switch to open/ closed.
In the case of step 7.2.2.1 (valve 1E11-F049), the procedure unnecessarily instructs the operator to
"take and hold" the control switch in the open position, as this valve is of the " seal-in" type. These procedural discrepancies were brought to the attention of the Operations Superintendent.
No violations or deviations were identified.
5.
Calibrathn (56700) Units 1 and 2 During this inspection period, the inspector reviewed the licensee's technical specification calibration program to determine if the program is being implemented in accordance with technical specification requirements.
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The licensee's program for the calibration of compments associated with safety-related systems or functions, but not specifically addressed in the technical specifications, was also examined.
Licensee procedure 90AC-0AP-001-05, " Test and Surveillance Program," establishes _ requirements for, implementation of the technical specification surveillance program at Plant Hatch. Licensee procedure 51GM-CAL-003-OS, " Calibration Program for LC0/B0P Instrumentation," describes-the methods by which non-technical
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specification instrument calibrations are. controlled.
This inspection effort involved reviews of the licensee's surveillance data bases, completed surveillance documentation, and surveillance procedures.
Selected ongoing functional test and calibration activities were also witnessed.
The inspector initially reviewed the licensee's technical specification surveillance data base to verify that calibrations have been performed at the required frequencies.
More specifically, tiie inspector reviewed selected technical specification calibration requirements associated with reactor protection, ECCS, PCIS, reactivity control, plant auxiliary, reactor coolant, primary containment, and electrical distribution systems.
In the cases of calibrations required on a "once per operating cycle" or
"once per 18 months" basis, the inspector confirmed that the last two calibrations had been performed at the required frequency.
In the cases of calibrations required on an "every 3 months" or "once per 92 days" basis, the inspector confirmed that the last four calibrations had been performed at the required frequency. The specific technical specification calibration requirements reviewed by the inspector are listed below along with brief descriptions of the associated trip functions or equipment.
Calibration Requirement Associated Trip Function / Equipment Unit 1 TS Table 4.1-1, Item 4 Reactor Vessel-High Pressure--RPS Unit 1 TS Table-4.1-1, Item 7.a Scram Discharge Volume High Level--RPS Unit 1 TS Table 4.1-1, Item 11 Turbine Control Valve Fast Closure--RPS Unit 1 TS Table 4.2-2, Item 2 Drywell Pressure - HPCI Initiation Unit 1 TS Table 4.2-4, Item 1 Reactor Vessel Level - ADS Initiation Unit 1 TS Table 4.2-1, Item 3 High Drywell Pressure - Isolation Unit 1 TS Table 4,2-1, Item 6 Main Steam Line Flow - Isolation Unit 1 TS Section 4.6.H.2.b Low Low Set Function of SRVs Unit 1 TS Table 4.2-11. Item 11 Hydrogen and Oxygen Analyzers Unit 1 TS Section 4.9.D.1(b)
RPS Protective Instrumentation Unit 2 TS Table 4.3.1-1, Item 4 Reactor Vessel Level - RPS Unit 2 TS Table 4.3.1-1, Item 6 Main Steam Line Radiation - RPS
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Unit 2 TS Table 4.3.1-1, Item 9 Turbine Stop Valve Closure - RPS Unit 2 TS Table 4.3.3-1, Item 2.a High Drywell Pressure - LPCI Initiation Unit 2 TS Table 4.3.2-1, Main Steam Line Low Pressure--Isolation
Item 1.c.2
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Unit 2 TS Section 4.1.3.5.b.2 Scram Accumulator Pressure Detectors l
Unit 2 TS Section 4.4.2.1.a.2 SRV Tail-Pipe Pressure Switches j
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The inspector reviewed completed surveillance documentation for selected calibrations required by the technical specifications.
The inspector confirmed that in each case the documentation was complete, acceptance criteria had been met, and the proper surveillance procedure had been used.
The specific calibration records that were reviewed by the inspector are listed below.
Procedure Date of Calibration 575V-C11-002-IS, "High Scram Discharge Volume March 3, 1989 Level Instrument FT&C" 42SV-821-005-15. " Low Low Set LSFT" November 21, 1988 575V-B21-006-25. " Main Steam Line Pressure October 23, 1988 Instrument FT&C" 575V-C11-003-25, " Control Rod Scram Accumulator August 10, 1988 Pressure and Leak Detector FT&C" 575V-821-018-2S. " Safety Relief Valve Pressure March 2, 1988 Switch FT&C" The inspector examined the contents of selected functional test, calibra-tion, and time response test procedures that satisfy technical specification requirements. These examinations focused on the adequacy of precautions, limitations, prerequisites, specified special requirements, acceptance criteria, and provisions for data review.
The adequacy of instructions for removing and returning equipment to service and for signal insertion and readout were also considered in these examinations.
The inspector observed that nine of the twelve procedures selected for review had been processed through all phases of the licensee's PUP process.
The specific procedures that were examined by the inspector are listed below.
34SV-C71-001-IS, " Turbine Stop Valve Instrument FT," Rev. 1 57SV-D11-019-2S, "Drywell High Radiation Instrument FT," Rev. 0 575V-SUV-011-'1S, "ATTS Panel 1H11-P925 Channel FT&C," Rev. 6 57SV-CAL-005-OS, "GE NUMAC Logarithmic Radiation Monitor Calibration,"
Rev. 2 57SV-CAL-003-IS, "ATTS Transmitter Calibration," Rev. 6 57SV-CAL-003-2S, "ATTS Transmitter Calibration," Rev. 2 57SV-MNT-003-2S, " Relay Logic Time Response Test," Rev. 6 575V-D11-003-25, " Time Response Testing of Main Steam Line Radiation Monitor Isolation Logic Trains," Rev. 1 57SV-MNT-002-2S, " Time Response Testing of Pressure Sensors," Rev. 1 57SV-B21-018-15. " Safety Relief Valve Pressure Switch Channel FT&C,"
Rev. 2 57SV-P33-001-1S, "Comsip Delphi Model K-IV Hydrogen and Oxygen Analyzer FT&C," Rev. 1 57SV-C71-005-15, "RPS Power Monitors FT&C," Rev. 2
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.The licensee's program for the calibration of. components associated with
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- safety-related systems or functions. but not addressed in the' technical specifications, was reviewed in several, areas.
Licensee procedure 51GM-CAL-003-0S specifies that calibration intervals for such
' components : are to be determined based on.. experience. and manufacturer's -
recommendations, and may be exceeded by as much as.100 percent.. The I&C'
group has responsibility for maintaining the ~ data-base for these
!
l calibrations.
The inspector. reviewed this data base and verified that-
. calibrations are scheduled for selected components that are used to make
technical specification-related measurements.. The inspector also verified that. calibrations are scheduled for selected components that provide automatic control or activation ~ of a process.or components that are used-by operators during normal operations ~ or.. post-accident conditions.
As:
part of.this review, the inspector verified that approved procedures are
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available~ for the calibrations of these components. The MPL designations-of the components reviewed by the inspector are listed below along with the-associated ~ calibration intervals and - brief descriptions of the components' functions.
l MPL' Designation Calibration Interval Component Function 2P41-R611.
12 months River Level Indicator IC41-R003 12 months SLC Pump Discharge Pressure Indicator IE41-R613.
12 months HPCI System Flow Indicator
'2E21-R600A 12 months Core Spray Pump Discharge-Pressure Indicator 2R43-N001A 60 months EDG Day Tank Level Switch 1T48-R076 36 months Containment Nitrogen Flow Controller IT46-R601 18 months SGTS Discharge Flow Indicator 2E41-R764-1 18 months HPCI Woodward Controller Three procedures used to perform non-technical specification calibrations discussed in the previous paragraph were reviewed. These reviews focused on the adequacy of provisions for removing and returning equipment to service, provisions for data review, and acceptance criteria.
The specific procedures reviewed by the inspector are listed below.
57CP-CAL-092-25, Rev. 3. " Calibration Procedure for Sigma Model 1151, 9261, 9262, 9263, 9261X, 9262X, 9263X, 9264X, and 1151 VB Indicator RTD, Voltage, Current Input" 57CP-CAL-137-15, Rev. 6, " Calibration Procedure for Bourdon Tube Style and Bellows Type Pressure Indicator" 57CP-CAL-031-25, Rev. 3, " Calibration Procedure for GE Type 180 Indicator"
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Finally, :the inspector witnessed'several FT&Cs in progress.
On[ June-2
- 1489,. a ? main ' ' condenser vacuum FT&C in accordance with proce-( re 57SV-B21-005-15 was ' observed.
On June 22, : 1989, the repair. and ca.ibration of recorder IT47-R611 in.accordance. with proce-dure 57CP-CAL-162-1S and.MWO 1-89-2818 was observed.
The inspector confirmed that these activities were being_ performed in accordance with approved procedures.
No violations or deviations were identified. The inspector concluded that the licensee is. implementing = programs for' technical specification calibrations and-for. safety-related calibrations that'are not specifically addressed in the technical specifications.. 'It was also. concluded that technically adequate calibration procedures are available to support these programs.
6.
Surveillance Procedures and Records (61700) Units 1 and 2 Surveillance procedures were reviewed to determine whether the
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surveillance of-safety-related systems.and-components is being conducted
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' in 'accordance with approved procedures as required by the technical specifications, - the licensee's IST program, and the. approved fire
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protection / prevention' program. This inspection effort did not include the review of : calibration 3ctivities.
The calibration of safety-related components was previo>usly addressed in this report under NRC Inspection :
Procedure 56700.
The ~ inspector initially verified that selected surveillance requirements in the _ technical specifications, IST program,'and. fire protection / preven-tion program are covered by properly approved procedures. While reviewing these procedures, the inspector confirmed that they. contained appropriate prerequisites, preparation instructions, acceptance criteria, and instructions for' returning equipment to service following surveillance.
- The surveillance requirements selected for review and the associated
- surveillance procedures are listed.
Surveillance Requirement Associated procedure Unit 1 TS 4.3.C.2.a 42SV-C11-001-15, Rev. 0, " Control Rod Scram
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Testing" Unit 2 TS 4.1.6.1.b 345V-C11-001-25, Rev. 1, " Scram Discharge Volume Isolation Valve Functional Test" Unit 1 TS Table 4.1-1, 34SV-B21-001-15 Rev. 1 "MSIV Closure Itein 10 Instrument Functional Test" Unit 2 TS Table 4.3.1-1, 345V-C71-001-25, Rev. 3, " Turbine Stop Valve Item 9 Instrument Functional Test" Unit 1 TS 4.6.H.1.b 345V-821-004-15, Rev. 3, " Relief Valve Operability" l
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Surveillhce' Requirement Associated Procedure Unit 2 TS 4.4.1.1.1
'34SV-B31-001-25, Rev. 2, " Recirculation Pump.
R Discharge Valve Operability"'
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' Unit 1 TS 4.5.E.1.d 34SV-E51-002-15. Rev. 3, RCIC Pump
i Operability" Unit 2 TS 4.5.1.b 345V-E41-002-2S, Rev. 6. "HPCI' Pump j
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Operability"
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Unit.1 TS 4.7.A.6.a 34SV-T48-003-IS, Rev. 2,." Post LOCA.
. Containment. Atmosphere Dilution System i
Functional Test" I
' Unit 2 TS 4.6.1.4.b-34SV-E32-001-2,- Rev. O, "MSIV Leakage J
Control System Valve Operability" j
Unit 1 TS 4.9.D.1(a)
57SV-C71-005-15, Rev. 2, "RPS Power Monitor i
Functional Test and Calibration" j
Unit 2 TS 4.7.1.2.c 435V-ECV-001-2S, Rev. 0, " Monitoring River i
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Bottom Conditions. at the Intake Structure,
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and River Stage Relationship"
!
. Appendix B..Section 2.1.2 42SV-FPX-032-05,. Rev. O, " Automatic. Sl iding f
'of FHA Fire Door. Surveillance"
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Appendix.B. Section 2.1.1'.c 42SV-FPX-019-IS, Rev. 1, " Penetration Seal of FHA Surveillance"
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ASME Section XI, Part IWV 34SV-E11-002-IS, Rev. 5. "RHR Valve Operability" ASME Section-XI,'Part IWV 345V-E21-002-2S, Rev 3. " Core Spray Valve-Operability"
'ASME Section XI, Part IWV-34SV-E41-001-15, Rev. 4, "HPCI Valve
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Oper6hility"
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ASME'Section XI, Part IWP 34SV-C41-002-25, Rev. 3, " Standby Liquid Control Pump Operability Test" ASME Section XI, Part IWP 345V-P41-001-15, Rev. 1,~" Plant Service.
Water Pump Operability"
ASME Section XI, Part IWP 34SV-E11-004-25, Rev. 4, "RHR Service Water
]
Pump Operability" i
i Some of the procedures listed above were' selected for more detailed
reviews - The inspector examined the technical contents of the selected procedures to verify that performance of the procedures as written would ensure compliance with the associated regulatory requirements.
The inspector examined procedures 34SV-B21-001-15, 34SV-B21-004-1S,
. 34SV-B31-001-25, 345V-C11-001-2S. 34SV-E21-002-2S, 34SV-E51-002-15, and 34SV-P41-001-15.
No discrepancies were noted in these more detailed i
'
reviews.
No violations or deviations were identified.
Based on the review of l
procedures for selected surveillance requirements, the inspector concluded l
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that written, approved procedures are available to support the fulfillment
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-of surveillance requirements in the technical specifications, the IST program for pumps and valves, and the fire protection / prevention program.
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.7.
Review of Three Mile Island Items (92701) Unit 2 As' discussed in NRC Inspection Report Nos. 50-321/89-06 and 50-366/89-06, of the TMI items assigned to the resident inspectors for review, only item II.F.1.6 (item number from NUREG 0737) remained open.
This item concerned containment hydrogen monitoring and remained open for Unit 2 only. Related work was performed in Unit 2 under DCR 81-165. DCR 81-165 was closed by the. licensee on March 15, 1989.
The documentation package for this DCR was reviewed by the inspector on June 21, 1989.
Item II.F.1.6 is now closed for Unit.2.
8.
Exit Interview (30703)
The inspection scope and findings were summarized on June 23, 1989, with those persons indicated in paragraph I aoove.
Particular emphasis was placed on the two NCVs and two URIs discussed in paragraph 2.
The licensee was also advised that the TMI item II.F.1.6 discussed in paragraph 7 was considered to be closed for Unit 2.
The licensee did not identify as proprietary any of the material provided to or reviewed by the inspectors during this inspection.
Dissenting comments were not received from the licensee.
Item Number Status Description / Reference Paragraph
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321/89-10-01 Opened and NCV - Failure to Have a SLC Closed System Valve Locked as Required (paragraph 2)
366/89-10-02 Opened URI - Installation of Non-Qualified Circuit Breakers in Safety-Related Applications (paragraph 2)
366/89-10-03 Opened and NCV - Inadequate ADS LSFT Closed (paragraph 2)
366/89-10-04 Opened URI - Inadvertent Loss of Logic Power for Drywell Floor Drain and Equipment Drain Pumps (para-l
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graph 2)
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Acronyms and Abbreviations
' ADS. -
Automatic Depressurization System ASME -
American Society of Mechanical Engineers Boiler and Pressure Vessel Code ATTS -
Analogue Trip Transmitter System B0P Balance-of-Plant
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CFR - '
Code of Federal Regulations Deficiency Card DC
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Design Change Request DCR
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ECCS -
Emergency Core Cooling System Emergency Diesel Generator EDG
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C0F '
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Engineered Safety Feature ESF -
Fire Hazards Analysis FHA
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FT Functional Test
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FT&C -
Functional Test and Calibration GE General Electric
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High Pressure HP
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HPCI -
High Pressure Coolant Injection I&C Instrumentation and Controls
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Inservice Testing IST
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Limiting Condition for Operation LC0
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Licensee Event Report LER
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LOCA -
Loss of Coolant Accident LPCI -
Low Pressure Coolant Injection LSFT -
Logic System Functional Test MCC Motor Control Center
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Master Parts List
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MSIV -
Main Steam Isolation Valve MWO Maintenance Work Order
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NCV Non-Cited Violation
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Nuclear Regulatory Commission NRC
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NRR Office of Nuclear Reactor Regulation
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OSOS -
On-Shift Operations Supervisor PCIS -
Primary Containment Isolation System i
PSW Plant Service Water
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Procedures Upgrade Program PUP
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RBCCW-Reactor Building Closed Cooling Water System RCIC -
Reactor Core Isolation Cooling RHR
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Residual Heat Removal System Reactor Protection System RPS
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SGTS -
Standby Gas Treatment System Standby Liquid Control SLC
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Source Range Monitor SRM
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Safety / Relief Valve SRV
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TMI Three Mile Island
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Technical Specifications TS
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URI -
Unresolved Item I
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