IR 05000219/1981005
| ML20011A339 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 05/28/1981 |
| From: | Blumberg N, Caphton D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20011A329 | List: |
| References | |
| 50-219-81-05, 50-219-81-5, NUDOCS 8110090330 | |
| Download: ML20011A339 (19) | |
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e U.S. NUCLr_AR REGULATORY COMMISSION OFFICE 0F INSPECTION AND ENFORCEMENT Region I Report No.
50-219/81-05 Docket No.
50-219 License No.
DPR-16 Priority Category C
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Licensee:
Jersey Central power and Light Company Madison Avenue at Punch Bowl Road Morristown, New Jersey 07960
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Facility Name: Oyster Creek Nuclear Generating Station Inspection at:
Forked River, New Jersey Inspection conducted: March 9-13 1 81 Inspectors:
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'N. J. Blumberg, Ke~ actor Inspector date signed T
date signed Approved by:
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D. CapEton M hief, Management Programs Section, d4te sfgned Engineering Inspection Branch Inspection Summary:
Inspection on March 9-13, 1981 (Report No. 50-219/81-05)
Areas Inspected:
Routine, unannounced inspection by one region based inspector of licensee action on previous inspection findings; administrative controls for safety related calibrations and surveillance; program and implementation for Technical Specification surveillance calibrations; program and implementation for calibration of plant instruments which verify Technical Specification surveillance tests; program and implementation for calibration of test and measurement equipment; program and implementation for Technical Specification
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surveillance testing; and implementation of inservice test program for pumps and valves.
This inspection invol,ved 41 inspector-hours onsite by one region-based
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inspector.
Region I Form 12 (Rev. April 77)
8110090330 810529'
PDR ADOCK 05000219 G
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Inspection Summary (Continued):
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Results: Noncompliances - None in two areas and seven in five areas (Violation -
Test gages of insufficient accuracy for tests being performed, paragraph 6.c(1); Violation - Frocedures not used for calibration of safety related instruments and certain instruments not calibrated, paragraph 5.c(1); Violation -
Inservice Testing for certain valves not being accomplished, paragraph 8.c(1);
Violation - PORC meeting minutes not distributed in a timely manner, paragraph 2; Violation - Annual periodic procedure review not being performed, paragraph 2; Violation - Deficient procedure not revised, paragraph 7.c(1); and Violation -
Calibration data not recorded or reviewed, paragraph 5.c(2)).
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DETAILS 1.
Persons Contacted A. P. Crosby, Reviewing Engineer
- J. Early, Instrument Maintenance Supervisor R. Ferret, Inservice Test Engineer.
- K. Fickeissen, Manager, Plant Engineering
- V. Foglia, Preventive Maintenance Engineer T. Gaffney, Instrument Technical Analyst J._Hascek,-Instrument and Electrical (I&E) Group Supervisor E. Johnson, I&E Group Supervisor
- M. Laggart, Licensing Supervisor
- J. Maloney, Manager, Plant Maint'enance
- T. Quintenz, Operations Engineering Lead Engineer A. Rone, Plant Engineering Manager
- J. Sullivan, Jr., Manager, Operations
- R. Tilton, Jr., Quality Assurance Specialist III USNRC
- J. Thomas, Resident Reactor Inspector The inspector also interviewed other licensee personnel including members of the administrative clerical staff.
- denotes those present at the exit interview.
2.
Licensee Action on Previous Inspection Findinos (Closed) Unresolved Item (80-16-03):
Plant Operations Review Committee (PORC) meeting minutes had not been approved or distributed in the six month period between December 1979 and May 1980.
The licensee stated the backlog of PORC meeting minutes would be corrected by November 30, 1980.
During this inspection, the inspector determined that the last meeting minutes to be issued was PORC meeting 104-80, July 1,1980.
A licensee representative stated that.aost PORC meeting minutes since July 19ev had been handwritten or placed on tape but had not been typed, approved or c.istributed.
The inspector informed the licensee that the Technical Specification 6.5.2 8 recaires that copies of PORC minutes and determinations shall be providea to the Director Oyster Creek Operations; the Vice President, JCP&L; Director, Oyster Creek; ti,a Independent Safety Review Group (ISRG)
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Coordinator and the Chairman of the General Offices Review Board (GORB);
and that, to enstre proper safety reviews, this must be done in a timely manner as required by ANSI N18.7-1972.
The licensee's representative acknowledged the inspectors comments and concurred that the current status of PORC meeting minutes was unacceptable.
The licensee's repre-sentative fu>ther stated that the continuing pi..
im in this area was caused by the limited availability of typists.
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Based on the above, less-than'adequat'e resolution, this unresolved item
'is closed and is'now considered an item of noncompliance in that there was' failure to distribute PORC meeting minutes as-required by Technical
. Specification 6.5.2.8 (219/81-05-01).
(Closed) Unresolved Item (219/80-16-06):
Seven procedures had not received an annual review or had not had their performance documented as required by Procedure 107, Procedure Control.
The licensee stated that all periodic procedure reviews would be1 documented.
During this inspection, the inspector observed that three of the seven procedures had been revised within 'the last year which constitutes an acceptable review as specified
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in ANSI N18.7. However, there was no documentation of reviews of the following four procedures:
202.1, Pcwer. Operation;
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302.2,.CR0 Manual Control System;
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315.2, Turbine Lube Oil; and
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340.2,'24 V.D.C. Distribution-System.
Additionally, a licensee representative stated that periodic procedure reviews were not being conducted.
Based on the above, less than adequate resolution, this unresolved item is closed and is now considered an item of noncompliance based upon failure to conduct annual procedure reviews as required by Technical Specification 6.8.2 and Administrative Procedure 107 (219/81-05-02).
(0 pen) Unresolved Item (219/80-16-02): Technical Specificaticn change is required to reflect the upgraded fire protection systems. The Licensee was to submit a proposed change by June 30, 1980.
During this inspection, the inspector determined that a Technical Specification change proposal was submitted to NRR (Licensing) July 14, 1980. A licensee representative stated that based upon telephone conversations between licensee representa-tives and NRR, modifications were required to the Technical Specification proposal. The licensee has modified the proposal and on March 12, 1981 the PORC reviewed the change.
The proposal is to be resubmitted to the NRC for review. This item remains open pending final issuance of revised Technical Specifications concerning fire protection.
(Closed) Unresolved Item (219/80-16-05):
Log sheets used for logging Technical Specification parameters were not procedurally controlled; had not been reviewed by the PORC; had not been approved by the Station Manager; and did not specify maximum and/or minimum parameter limits.
The inspector reviewed Procedure 106, " Conduct of Operations," and deter-mined that the procedure has been revised to include the log for Technical Specification parameters and that the limits for each parameter have been included on the log sheet.
Based on this determination, this item is closed, i
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(Closed) Unresolved Item (219/79-16-02):
Stop watches used for timing
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surveillance tests not identified or controlled.
The inspector verified that stop watches for surveillance testing have been given a control number; are kept in a controlled location; are signed out prior to use; and the test for which it is used is listed on the signout sheet.
Based on this verification, this item is closed.
3.
Administrative Controls for Safety Related Calibrations and Surveillances Administrative controls governing the performance of safety related calibrations and surveillances were inspected to determine their conformance with the requirements of 10 CFR 50, Appendix B, " Quality Assurance Criteria for Nuclear P wer Plants and Fuel Processing Plants;" Technical Specification, Section 6, radministrative Control's;" ANSI N18.7-1972, " Administrative Controls for Nuclear Power Plants," and Regulatory Guide 1.33-1972,
" Quality Assurance Program Requirements (Operation)." The following procedures were reviewed:
103, Station Document Control, Revision 8, November 13, 1980;
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107, Procedure Control, Revision 13, October 31, 1980;
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112, Oyster Creek Calibration of Maintenance Test and Inspection
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Tools, Gauges and Instru;.qents, Revision 12, October 31, 1980; 112.1, Calibration of Technical Specification Supporting Installed
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Instrumentation, Revision 9, October 31, 1980; 112.2, Oyster Creek Calibration and Maintenance of Radiation Protection
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Instruments, Revision 0, November 9, 1979; 113, Calibration of Instslied Plant Instruments, Revision 5, October
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31, 1980; 114, Testing, Revision 2, October 31, 1980;
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116, Surveillance Test Program Schedule and Review of Test Results,
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Revision 8, October 31, 1980; and Inservice Inspection Program, Revision 3, September 8, 1980, Section
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2, " Pump and Valve Test Program."
No unacceptable conditions were observed.
4.
Surveillance Calibrations Required by Technical Specifications Program and Implementation a.
An inspection was conducted, on a sampling basis, of the Program and its implementation for surveillance calibrations of components as required by the Technical Specifications.
The program was inspected for conformance to the requirements of ANSI N18.7-1972, " Administrative
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Controls for Nuclear Power Plants," ar.d Regulatory Guide 1.33 -
1972, " Quality Assurance Program Requirements (Operation)." Addition-ally, the program was inspected for implementation in accordance with the above standards and the appropriate plant administrative requirements detailed Us paragraph 3.
The following areas were also verified:
A master schedule has been established for surveillance calibration
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testing; Responsibilities have been assigned for performance of surveillance
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calibrations and to assure that calibration schedule; are satisfied;
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Methods and responsibilities have been established for review and evaluation of calibration data, for reporting calibration deficiencies and failures, and for verification that LCO require-ments have been satisfied; Calibration frequency requirements in accordance with Technical
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Specifications were satisfied;
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Applicable system status during component calibrations was in
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conformance with Technical Specification limiting condition for operations, where applicable;
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Procedure format provided detailed stepwise instructions in the degree of detail necessary for performing the calibrations; Procedure review and approval were as required by Technical
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Specifications;
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The Technical content of procedures was sufficient to result in satisfactory component calibration and Technical Specification limits and setpoints were complied with; and Calibration data was adequately and accurately recorded and was
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within established tolerances.
b.
The following procedures were reviewed:
610.3.005, Core Spray System Instrument Channel Calibration and
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Test (High Drywell Pressure Core Cooling Section), Revision 9, July 1, 1980.
Data reviewed for tests performed January 22, 1981; December 30, 1980; and December 3, 1980.
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621.3.003, Main Steam Line Radiation Monitor Test and Calibration, Revision 3, March 17, 1980.
Data reviewed for test performed j
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621.4.007, Air Ejector Off Gas Radiation Monitor Front Panel
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Test, Revision 3, December 3, 1979.
Data reviewed for tests performed February 9,1981; February 2,1981; and January 26, 1981.
t 603.3.002, Reactor Recirculation Flow Calibration, Revision 1,
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February 29, 1980.
Data reviewed for test performed January 23, 1980.
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1001.9, IRM Calibration to Reactor Power, Revision 1, June 26, 1978. Data reviewed for tests performed November 22, 1980; September 23, 1980; and August 5, 1980.
No unacceptable conditions we're identified.
5.
Calibrations of Plant Installed Instrumentation Associated With Verification of Technical Specification Surveillance Tests Program and Implementation a.
A selective sampling inspection was conducted of the program and its implementation for calibration of plant installed instrumentation used to verify satisfactory performance of Technical Specification surveillance tests or Inservice Testing.
The program was inspected for conformance to the requirements of ANSI N18.7-1972, "Administra-tive Controls fo,r Nuclear Power Plants;" and Regulatory Guide 1.33-1972,
" Quality Assurance Program Requirements (Operation)." Additionally,
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the program was inspected for implementation in accordance with the a ove standards and the appropriate plant administrative requirements
detailed in paragraph 3.
The following areas were verified:
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A master schedule has been established for calibration testing;
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Responsibilities have been assigned for performance of calibrations and to assure that calibration schedules are satisfied; Calibration schedule and frequencies as established by the
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licensee are being followed;
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Procedures have been reviewed and~ approved in accordance with Technical Specifications, contain acceptance criteria consistent with Technical Specification requirements and contain detailed instructions commensurate with the complexity of the calibration;
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The technical content of procedures is adequate to perform satisfactory calibration; and
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Calibration data was adequately and accurately recorded and was within established tolerance _
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b.
The following instruments used for verification of surveillance parameters required by the Technical Specifications were selected at random and the specific requirements established for their calibration were verified:
ILOS, Liquid Poison Pump Pressure Indicator.
Data was reviewed
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for calibr ations performed June 6,1979; January 28, 1975; and March 21, 1973.
ILO4, Liquid Poison Pump Pressure Transmitters.
Data were
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reviewed for calibrations performed June 2, 1980; June 6, 1979; January 28, 1975; and March 21, 1973.
RD04, CRD Drive H ] vs. Reactor Pressure Indicator.
Data
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reviewed for calibrations performed January 4,1979; January 24, 1978; and October 28, 1976.
RD18, CRD Charging Water Pressure Transmitter.
Data reviewed
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for calibrations performed January 8,1981; January 4,1980; Oct3ber 1, 1979; and January 9, 1978.
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RD1., CRD Filter Flow Local Pressure Indicator.
Data reviewed
.for calibrations performed January 7,1980; January 9,1979; and January 25, 1978.
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IG06A, Engineering Condenser A Level Transmitter.
Data reviewed for calibration performed May 7, 1980.
IG078, Emergency Condenser A Level Indicator.
Data reviewed
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for calibrations performed May 5,1980; December 29, 1979; May 18, 1979; and May 1, 1978.
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FP-08, Fire Pump 1-2 Discharge Pressure Recorder.
Data reviewed for calibrations performed July 29, 1980; August 2, 1979; and February 25, 1978.
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EDST, Emergency Diesel Storage Tank Level Indicator.
Data reviewed for calibrations performed November 4, 1980; November 11, 1979; and December 28, 1978.
c.
Findings (1) ANSI N18.7-1972 requires that procedures be used for performar.ce of calibrations of safety related pisnt instruments.
The inspector observed that a Procedure (112.1) had been provided which established the safety related instrument calibration program.
However, this procedure did not define specific calibration requirements for each instrument.
Calibration requirements for each instrument were recorded on calibration record cards located in the instrument shop.
These cards provided for the input and output signals and their tolerances
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but had not been reviewed by PORC and approved by the Director Oyster Creek Operations.
The inspector informed the' licensee that data sh:6ts used for calibrations of safety related equipment
.must receive the same review and approval as procedures and must contain the essential aspects for calibrations,
specified in ANSI N18.7-1972.
Additionally, the Diesel Gerierator kilowatt and kilovar me,
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and the Fire Pump RPM Meters were'not being calibrated.
The instruments were being used to verify Lna acceptability of Technical Specification surveillance tests. The inspector informed the licensee that the accuracy of plant instruments used to verify surveillance requirements must be established.
. Failure to establish procedures for the performance of safety related calibration and' failure to calibrate certain safety related items is considered an item of noncompliance (219/81-05-03).
(2) Plant Procedure 112.1, " Calibration of Technical Soecification Supporting Installed Instrumentation" requi es thLt calibration data be recorded on the instrument history cards maintained in the instrument shop and that this data be reviewed by the Group I and C&E Supervisor for completeness and for instrtment problems and trends. The inspector determined that data had not been recorded for the following calibrations:
IL-05, Standby Liquid Control System discharge pressure
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calibration performed June 20, 1980; RC-04, Control Rod Drive Hydraulic Pump discharge pressure
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calibration, performed May 9, 1980 and January 8, 1981; and RD-15, Control Rod Drive Hydraulic Pump Flow calibration
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performed January 7, 1981.
Further, the inspector determined that the supervisory review required by 112.1 is apparently not being performed as evidenced by the fact that another record which documents completion of calibration reviews were signed verifying completion of reviews for the above instruments although the calibration data was not recorded.
Failure to record and review calibration data is contrary to Technical Specification 6.8.1 and Plant Procedure 112.1, para-graphs 5.4 and 5.7 and is considered an item of noncompliance (219/81-05-11).
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-(3) Standby Gas Treatment System filter differential pressure
' i n strun.ents 1-10, 1-11 and 1-12 have been changed from water manometer to 0-5" water magnahelic gages.
The licensee provided documentation verifying that these gages were calibrated prior to initial installation on November 17, 1980; and further stated that action had been taken to add these instruments to Procedure 112.1 for scheduling calibration of balance of plant instruments.
However, the licensee was unable to provide
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documentation to the inspector that such a change was in progress.
The licensee stated that Procedure 112.1 would be revised by June 1, 1981 to include the scheduling of calibrations for the SGTS filter differential pressure instruments 1-10, 1-11 and 1-12.
This item is unresolved pending licensee action and subsequent NRC:RI review (219/81-05-04).
6.
Calibration and Control of Test and Measurement Equipment Program and Implementation An inspection was conducted to determine the adequacy of calibration and control of test and measuring equipment. The program was inspected for conformance to ANSI N18.7-1972, " Administrative Controls for Nuclear Power Plants;" Regulatory Guide 1.33-1972, " Quality Assurance Program Requirements (Operation); and ANSI N45.2.4, " Installation, Inspection and Testing Requirements for Instrumentation and Electric Equipment During the Construction of Nuclear Power Generating Stations."
Additionally, the program was inspected for implementation in accordance with the above standards and the appropriate plant administrative require-ments detailed in paragraph 3.
The following arrar were verified:
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Establishment and maintenance of a master test equipment list;
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Establishment and adherence to calibration schedule; Test equipment was in calibration when in use;
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Calibration data, where available, was adequate and accurate and
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within specified tolerances;
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Maintenance of calibration records identifying standards used which have traceability to the National Bureau of Standards or other testing organization; Test equipment custody control records;
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Storage and labeling of test equipment;
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Test equipment used as primary standards for in plant calibrations of test equipment were in calibration;
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Traceability of out of calibration test equipment to usage of that
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equipment; Establishment of responsibilities for control of test equipment; and
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Control and calibration of eight pieces of test equipment used for
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calibration or test of components identified in paragraphs 4.b, 5.b, 7.b and 8.b.
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Findings
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(1) During review of calibration data for test gages, the inspector
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observed that there was considerable drift in test gages although they were being calibrat'ed every three months.
In some instances, gages were observed to be out of calibration. When questioned by the inspector as to what determinations were made as to the effect on tests previously performed using these gages, a licensee representative stated that this was not done since as a matter of policy all gages were given a one point calibration check prior to and after use and readjusted as necessary. The inspector noted that this check was not documented nor was it required by the existing administrative procedures.
The inspector informed the licensee that if one point calibration checks were used to verify the accuracy of instruments between calibrations, that this requirement must be included in station administrative instructions and that such checks must be documented.
Additionally, the inspector informed the licensee that adjustments to instruments based on one point calibration checks could not be made unless a complete recalibration u done.
Subsequent to the inspection, a licensee representative i.
7,ed the inspector that a revision to procedure 112 was in progress to reflect the latter and that this revision would be implemented by June 30, 1981.
This item is unresolved pending licensee action and subsequent NRC:RI review (219/81-05-13).
Further, the inspector observed that safety related tests were being accomplished using 0-1500 psig gages which had 5 psi minimum increments and were calibrated to an accuracy of 1 0.5 percent (17.5 psig) of full scale.
Examples of tests in which these gages were used as test equipment were as follows:
619.3.008, Turbine Load Rejection Scram Test (< 40% Load),
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performed February 13, 1981, with test setpoints established at 180 1 5/0 and 165 1 0/5 psig.
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602.3.004, Electromatic Relief Valve pressure Sensor Test and Calibration, performed December 1, 1980, vith test setpoints established at 1059 1 0/5 psig, 1084 ? 0/5 psig, 1076 1 0/5 psig, 1062 1 0/5 psig, and 1082 1 0/5 osi.
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609.3.003, Isolation Condenser Automatic Actuation Sensor
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Calibration and Test, performed February 5, 1981, with test setpoints established at < 1068 1 0/2 psig and < 1066 1 0/2 psig.
619.3.007, Low Pressure Main Steam Line Functional Calibration
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Test While Shutdown, performed July 4,1980, with test setpoint established at 850 1 5/0.
619.3.017, Reactor High Pressure Scram Test and Calibrattor,
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performed February 10, 1981, with test setpoints established at 1068 1 0/2 psig and 1066 1 0/2 psig.
The~ inspector informed the licensee that ANSI N45.2.4. requires that test equipment _be of acceptable accuracy for the tests being performed and that the test gage accuracy of 2 7.5 psig appeared to be unacceptable for setpoint tests with accuracies of from 2 to 5 psig. Additionally, some of the tests above, required readabilities to increments of 1 or 2 psig and that 5 psig increment gages could only expect to have a readability of 2.5 psig (i.e. one half of one gage increment).
The licensee's representative acknowledged the inspectors comments, however, did not concur.
The licensee representative stated that the gage drift had been recognized by their instrument personnel and that the calibration check and adjustment just prior to and after test gage use assured the required instrument accuracy. Additionally, the licensee stated that a 5 psig increment gage could be read to within 1 psig by a qualified instrument technician.
Failure to use test equipment with acceptable readibility and accuracy for the tests being performed is contrary to 10 CFR 50, Appendix B, Critrion XII: JCP&L Quality Assurance Plan,Section III, Operational Quality Assurance Program; and ANSI N45.2.4-1971; and is considered an item of noncompliance (219/81-05-05).
7.
Surveillance Testing Required By Technical Specifications Program and Implementation a.
An inspection was conducted, on a sampling basis, of the program and its implementation for the surveillance testing of equipment and systems as required by the Technical Specifications.
The program was inspected for conform:nce to the requirements of ANSI N18.7-1972,
" Administrative Cont ois for Nuclear Power Plants", and Regulatory Guide 1.33-1972, " Quality Assurance Program Requirements (Operation)".
Additionally, the program was inspected for implementation in accord-ante with the above standards and the appropriate plant administrative requirements detailed in paragraph 3.
The following areas were verified:
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A master schedule has been established for surveillance testing;
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Responsibilities have been assigned for performance of tests
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and to assure that test schedules are satisfied; Methods and responsibilities have been established for review
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and evaluation of test data, for reporting test deficiencies and' failures, and for verification that LCO. requirements have been satisfied; Tests required by Technical Specifications are available and
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' covered by properly approved procedures; Test format and technical content are adequate and provide
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satisfactory testing of related systems or components; Tests have been reviewed as required by Facility Administrative
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Requirements; and
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Tests were performed within the time frequencies specified by the Technical Specifications and appropriate action was taken for any item failing acceptance criteria.
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b.
The following surveillance tests and data for completed tests were reviewed:
612.4.002, Liquid Poison System Functional Test, Revision 4,
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January 9, 1980.
Data reviewed for tests performed June 28, 1980 and November 15, 1978.
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602.4.003, Electromatic Relief Valve Operability Test (ADS Function), Revision 1, June 30, 1978.
Data reviewed for tests performed January 24, 1981 and July 16, 1980.
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602.3.005, ADS Actuation Circuit Test and Calibration, Revision 1, April 25, 1979.
Data reviewed for tests performed April 14, 1980 and November 19, 1978.
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604.3.001, Reactor Building to Torus Power Vacuum Breaker Test, Revision 7, June 2, 1980.
Data reviewed for tests performed February 13, 1981 and November 19, 1980.
604.4.002, Reactor Building to Suppression Chamber Self Activating
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Vacuum Breaker Operability Test, Revision 0, March 15, 1977.
Data reviewed for tests performed February 11, 1981 and November 19, 1980.
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604.4.003, Reactor Building to Suppression Chamber Self Activating Vacuum Breaker Surveillance Test, Revision 0, March 15, 1977.
Data reviewed for test performed June 16, 198 _
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634.2.002, Main Station Weekly Battery Surveillance, Revision
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5, June 30,1980.
Data reviewed for Station Battery "A" tests performed February 6 and 13, 1981; and January 30, 1981.
634.2.003,' Main Station Battery Monthly Surveillance, Revision
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4, January 10, 1980.
Data reviewed for Station Battery "A" tests performed February 10, 1981; January 16, 1981; and December 19, 1980.
634.2.001, Main Station Battery Discharge (Load Test), Revision
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10, January 10, 1980.
Data reviewed for Station Battery "A" tests performed February 10, 1981 and August 26, 1980.
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645.1.007, Fire Protection System Flush, Revision 0, October 17, 1978.
Data reviewed for tests performed December 2, 1980.
645.1.008, Fire Hose Re-Rack and Blockage Check, Revision 1,
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January 29, 1980.
Data reviewed for test performed May 16, 1980.
Standby Liquid Control System Liquid Poison Analysis Data
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Sheet. Data verified per Figure 304-1 of Procedure 304, Standby Liquid Control System, Revision 7, January 14, 1980; and reviewed for tests performed March 3,1981; February 3,1981; January 31, 1981; and January 16, 1981.
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610.4.002, Core Spray Pump Operability Test, Revision 5, June 27, 1980.
Data reviewed for various tests performed from July 1980 to February 1981.
c.
Findings (1) Procedure 610.4.002, Core Sorsy Pump Operability Test, also tests the automatic start and stop features of the Core Spray Systems I and II fill pumps.
The procedures require that the fill pumps stop when their associated Core Spray pumps start and that the fill pumps restart when their associated Core Spray Pumps stop.
The inspector observed that since November, 1980, the test data sections that verify automatic start and stop features of the fill pumps were annotated as "Not Applicable".
The licensee's representative stated that an earlier modification had wired
"C" Core Spray Pump in System I and "D" Core Spray Pump in System II to the same power source.
However, this modification failed to change over the fill pump start and stop features associated w.th those pumps, hence starting or stopping either of the "C" and "D" Core Spray Pumps was starting and stopping the fill pumps in the opposite Core Spray Syste.
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A modification has been scheduled to correct this problem. As an interim measure, the licensee had jumpered the fill pumps so that they would remain in operation during Core Spray Pump operation.
The licensees representative stated that operation of fill pumps during Core Spray Pump operation appears not to adversely affect the capability of the Core Spray System.
The inspector concurred with the licensees evaluation that continuous
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operation of the fill pumps would not adversely affect performance of the Core Spray System.
The inspector informed the licensee that a procedure should accurately reflect status of all items being tested and that Administrative Procedure 107 requires that procedures which are not adequate be revised ~or temporarily changed.
Failure to follow procedures and failure to change a procedure to correct a procedural inadequacy is contrary to Technical Specification 6.8.1 and Plant Procedure 107; and is considered an item of noncompliance (219/81-05-12).
(?) Technical Specifications require periodic measurement of electro-lyte levels in the Main Station and Diesel Generator Batteries.
These measurements are accomplished per Procedures 634.2.002, 634.2.003, 636.2.005, and 636.2.006 and data is recorded in percent level. The percent level in relationship to actual battery level is defined in Procedures 632.2.002, 636.2.005 and 636.2.006 but not in Procedure 634.2.003.
The inspector informed the licensees representative that the percent level needs to be defined in each procedure in which electrolyte measurements are taken in order to determine accept-
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ability of data for procedures in use.
The licensee's repre-sentative acknowledged the inspectors comments and stated that Procedure 634.2.003 would be revised by June 1, 1981 to define percent electrolyte level.
This item is unresolved pending licensee action and subsequent NRC:RI review (219/81-05-06).
(3) The inspector observed that the following procedures required measurement of the specific gravity for the main station and diesel generator batteries:
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634.2.002, Main Station Weekly Battery Surveillance;
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634.2.003, Main Station Monthly Battery Surveillance;
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636.2.004, Diesel Generator Battery Discharge (Load Test);
636.2.005, Diesel Generator Weekly Battery Surveillance;
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636.2.006, Diesel Generator Monthly Battery Surveillanc.
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In each instance, a temperature correction factor is applied and the specific gravity recorded has been corrected to 77 F.
However, the correction factor is not included as part of the procedure.
The inspector informed the licensee that in order to ensure the correctness of the specific gravities and completeness of data review the method for obtaining the temperature correction and the actual correction must be included in the procedure.
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licensee's representative acknowledged the inspectors comments and stated the above procedures would be revised by June 1, 1981 to show the method of obtaining correction factors and their application to specific gravity measurements. This item is unresolved pending licensee action ar,o subsequen. NRC:RI review (219/81-05-07).
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(4) The inspector observed that battery specific gravities are not measured before or after the Main Statio.1 Batteries load tests performed per 634.2.001 nor after Diesel Generator Station Battery load test performed per 636.2.004 The inspector expressed a concern that battery load tests may have a deleter-ious effect on specific gravities.
The licensee's representative acknowledged the inspectors concern and stated an engineering evaluation would be accomplished and documented to determine if specific gravity measurements are required after completion of battery load tests; and that this evaluation would be completed by September 1, 1981.
This item is unresolved pending licensee action and subsequent NRC:RI review (219/81-05-08).
(5) Technical Specification 4.7 requires load test of the Main Station batteries every six months.
This test is accomplished per Procedure 634.2.001, Main Station Battery Discharge (Load Test).
The inspector observed that the procedure specifies that "C" battery can only be load tested during periods of cold shutdown or during refuelings.
The licensee stated that the "C" Battery is an added station battery which is physically and electrically separated from the
"A" and "B" Station Batteries.
The safety related electrical loads cannot be transferred to another station battery during surveillance load testing; while safety related loads on the
"B" Battery can be electrically transferred to the "A" Battery during surveillance load testing.
The licensee further stated the current Technical Specifications were written when the Station had only "A" and "B" batteries and did not recognize the "C" battery.
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t The licensee had previously submitted a Technical Specification change concerning testing of station batteries after the "C" Battery was installed.
This eb ge was not approved but verbal discussions have continued between the licensee-and NRC Licensing to arrive at an acceptable Technical Specification. The licensee stated that during these discussions NRC Licensing gave verbal concurrence for an 18 month interval for load testing of the
"C" stattun battery.
The licensee showed the inspector a revised proposed Technical Specification change submittal which would change the frequency of the load tests for station batteries to once per 18 months during shutdown.
Further, the licensee stated this change had been reviewed by the PORC and would be submitted to the NRC by April 30, 1981.
The inspector noted that this-revision was consistent with BWR Standard Technical Specifications which only require a load test every 18 months.
This item is unresolved pending final approval and issuance of this Technical Specification, and subsequent NRC:RI review (219/81-05-09).
-(6) During review of 602.4.003, Electromatic Relief Valve Operability Test, the inspector observes that the test had not been revised since June 30, 1978; and her.
, did not include verification of operability of the new acoi".: cal monitor which is used to verify the lifting of the relief valves nor was the acoustical monitor used in the procedure as verification of relief valve lift. A licensee representative stated that he had recently observed this deficiency in the procedure and provided to the inspector a proposed change dated February 6,1981 which incor-porated the acoustical monitor into the procedure.
Prior to completion of this inspection, the licensee provided documentation to the inspector that a revision to Procedure 602.4.003 which included use of the acoustical monitor had been issued.
The inspector had no further questions in this area.
(7) The inspector observed Procedure 610.4.002, Core Spray Pump Operability Test, was cerformed November 30, 1980 using Revision 4, although Revision 5 had been in effect and in use since July 28, 1980.
This error was discovered by the licensee during test data review and prompt corrective action was taken.
This appeared to be an isolated case and the inspector had no further questions concerning this item.
8.
Inservice Test Program Pumps and Valves Implementation a.
An inspection was conducted, on a sampling basis, for the implementation of the licensee's Inservice Test Program (IST), Pumps and Valves.
The IST program implementation was inspected for conformance to the Inservice Inspection Progr;m, Secton 2, " Pump and Valve Test Program" and the appropriate plant administrative requirements detailed in paragraph 3.
The following areas were verified:
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Tests are in conformance with inservice test program requirements.
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Test frequency is in conformance with Technical Specifications
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and the Inservice Test Program.
Te.<c results have been reviewed as required by facility adminis-
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trative requirements and appropriate action was taken for results failing acceptance criteria.
b.
The following inservice tests and data for. completed tests were reviewed:
642.4.001, Reactor Building Closed Cooling Water Operability
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Test, Revision 0, June 27, 1980.
Data reviewed for test performed December 25, 1980.
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607.4.001, Containment Spray and Emergency Service Water Pump
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Operability Test, Revision 11, August 25, 1981.
Data reviewed for tests performed January 29, 1981 and January 6, 1981.
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614.4.001, Fuel Pool Cooling Pump Operability Test, Revision 0, June 27, 1980.
Data reviewed for test performed December 17, 1980.
609.4.001, Isolation Condenser Vaive Operability Test, Revi,sion
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5, January 9, 1980.
Data reviewed for tests performed January 15, 1981; December 18, 1980; and November 25, 1980.
c.
Findings (1) The licensee's Inservice Test (IST) Program for pumps and valves was approved by the NRC for implementation December 8, 1979; and is established per the licensee's Inservice Inspection Program, Section 2, " Pump and Valve Testing." The inspector requested that the licensee provide documentation of tests for seven valves which were listed in the IST program for testing.
The licensee's representative was unable to provide test proced-ures for or documentation of satisfactory testing of the following valves:
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Core Spray System Valves, V20-60, V20-61, and V20-10; Containment Spray System Valve, V21-10;
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Closed Cooling Water Valve, V5-154; and
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Fire Protection Syste Valve, V9-13.
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In each case above, a licensee representative stated the valves had been tested in conjunction with existing tests but that the tests had not been revised to document satisfactory operation of the above valves. The inspector then requested from the licensee an overall status of IST test procedures for valves listed in the IST Program.
Prior to the end of the inspection, the licensee provided to the inspector a partial listing of IST program valves.
This listing indicated a significant deficiency in valve testing in that of 162 identified to the inspector, 81 were not tested or were not tested within the required frequency ard an additional 25 were being tested by existing test procedures which did not document satisfactory tests of these valves.
Additionally, NRC inspection 219/80-10, conducted March 17-20, 1980, identified an item of noncompliance that procedures had not been-established for Inservice Testing pumps and valves.
The licensee's reply to this inspection dated June 3,1980 stated that IST procedures would be fully implemented by July 31, 1980. As evidenced by the deficiencies in valve testing noted above, corrective action for the previous nonccmpliance was not adequately implemented.
Failure to provide test procedures to provide documentation or to perform Inservice Testing of Valves is contrary to the JCP&L Operational Quality Assurance Program and ANSI N18.7-1972; and failure to providc adequate corrective action for a previously identified item of noncompliance is contrary to 10 CFR 50, Appendix B, Criterion XVI and are collectively an item of noncompliance (219/81-05-10).
(2) The station Document Control Center was unable to locate completed Procedure 614.4.001, Fuel Pool Cooling Pump Operability Test, performed December 17, 1980.
The test apparently had either been misfiled or lost. A licensee representative was able to provide to the inspector a copy of the completed test data for review.
During this inspection, the inspector observed that the licensees method for maintaining document records appeared to be satisfactory and that this missing document appeared to be an isolated case.
Subsequent i.c the inspection, a licensee representative informed the inspector that this procedure could not be located; however, a copy of the data would be processed and placed on record to replace the lost original.
The inspector had no further questions concerning this item.
9.
Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable, deviations, or items of noncompliance.
Five unresolved items were identified during this inspection and are detailed in paragraphs 5.c(3), 6.a(1), 7.c(2), (3), (4) and (5).
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10. Management Meeting Licensee management was informed of the scope and purpose of the inspection at the entrance interview conducted on March 9, 1981.
The findings of the inspection were periodically discussed with licensee representatives during the course of the inspection. An exit interview was conducted on March 13, 1981, see paragraph 1 for attendees, at which time the findings of the inspection were presented.
Subsequent telephone discussions concerning the inspection findings were conducted between the inspector and Mr. J. Sullivan on March 26, 1981 and the inspector and Mr. V. Foglia on May 27, 1981.
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