IR 05000272/1980006
| ML18082A860 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 05/21/1980 |
| From: | Hill W, Keimig R, Norrholm L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML18082A855 | List: |
| References | |
| 50-272-80-06, 50-272-80-6, 50-311-80-05, 50-311-80-5, NUDOCS 8008130233 | |
| Download: ML18082A860 (17) | |
Text
U. S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT 50-272/80-06 Report No /80-05 50-272 Docket No REGION I
Licensee:
Public Service Electric and Gas Company 80 Park Place Newark, New Jersey 07101 ief, Reactor Pro ects Branch Inspector Inspection Summary:
Inspections on March 30-April 30, 1980 (Combined Reports Nos. 50-272/80-06 and 50-311/80-05)
Unit 1 Areas Inspected: Routine inspections by the resident inspectors of plant operations including:
tours of the facility; log and record reviews; IE Circulars; management meetings; implementation of containment isolation and control rod guidance; and, followup on previous inspection item The inspections involved 67 inspector-hours by the NRC resident inspector Results: Five items of noncompliance were identified. (Infraction-Partial isolation of Auxiliary Feedwater without taking appropriate action as required by Technical Specifications - Details 4d, Infraction - Exceeding licensed power level during steady state operation - Details 4d, Infraction - Failure to record individual rod position every four hours as required by Technical Specifications Details 3d, Infraction - Failure to conduct monthly calibrations of feedwater flow instrumentation as required by 10CFR50, Appendix B, Criterion XII - Details 51 Deficiency - Failure to conduct quadrant power tilt ratio every twelve hours - Details 3d)
Unit 2 Are.as Inspected: Routine i nspecti ans of preoperati ona l and startup testing by the resident inspectors inc:luding:
tours of the facility; IE Circulars; preparedness for issuance of an operating license; participation in a Commission meeting; and, followup on previous inspection item The inspections involved 14 inspector-hours by the NRC resident inspector Results: No items of noncompliance were identifie /*
DETAILS 1. Persons Contacted J. Gallagher, Assistant Maintenance Engineer A. Meyer, Site QA Engineer L Meyer, Project QA Engineer H. Midura, Manager - Salem Generating Station P. Moe 11 er, NucJ ear,~L i cens:i.ng:,Eng.i neer W. Reuther, Site QAD.
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F. Schnarr, Station Operating Engineer R. Silverio, Assistant to the Manager J. Stillman, Station QA Engineer J. Zupko, Chief Engineer Attendees at April 18, 1980 Meeting NRC - Region I B. H. Grier, Director E. J. Brunner, Chief, RO&NS Branch J. W. Devlin, Chief Security and Investigation Section L. J. Norrholm, Senior RR!, Salem PSE&G F. W. Schneider, Vice President-Production F. P. Librizzi, General Manager-Electric Production H. J. Heller, Manager-Nuclear Operations H. J. Midura, Manager-Salem Generating Station J.M. Zupko, Jr., Chief Engineer-Salem P. A. Moeller, Nuclear Licensing Engineer
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T. J. Lesh, Security Supervisor-Salem
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- I The inspectors also interviewed and talked with other licensee personnel during the course of the inspections including management, clerical, maintenance, opera~ions, performance, quality assurance, and construction personne. Status of Previous Inspection Items (Closed) Follow Item (272/78-29-02) Completion of DCR-1-ED-0442, modification of containment air loc The inspector reviewed completed documentation for this design change, and the following Work Orders; 902981, 901251, 903779, 906010; and 93561 The work included stiffening of supports and renewal of parts in the airlocks~ The inspector had no further questions on this ite (Closed) Unresolved Item (272/79-14-01): Function of frequency selector switch on Sonic MARK I instrument. The inspector reviewed documentation indicating that an investigation was made to determine to what extent test results are influenced by different settings of the Sonic MARK I flow detector frequency selector switc The reviewed documents included:
--Southwest Research Institute (SWRI) letter dated October 12, 1979
--SWRI letter dated January 14, 1980 and attached photographs
--Calibtatiori sheet number SMW-07
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- -Based on the information reviewed, the inspector had no further question (Closed) Noncompliance (272/79-15-01) Inspection of containment fire hose stations. Based on review of documentation and tours of containment, the inspector verified that licensee action to ensure periodic inspection of containment fire protection equipment had been effective. The inspector had no further questi ans on this i tern*;
(Closed) Noncompliance (272/79-18-01) Inoperable spray-sprinkler system for # 13 Reactor Coolant Pum The inspector verified that corrective action in clearing the jumper, revision to the jumper control system, and incorporation of an unavailable equipment status listing had been accomp-lished. * The inspector had no further question (Closed) Unresolved Item (272/79-18-02) Deficiencies in jumper log control sheet Revision 4 to AP-13 has been issued, defining better controls over installed jumpers and conditions necessary to clear. The inspector also verified that the majority of jumpers in place have been converted to the new control syste Noting that some non-safety related jumpers have been in place since 1978 and should be replaced by an appropriate design change, the inspector had no further questions on this ite (Closed) Follow Item (272/79-18-06) NRB reviews of IE Bulletin and IE Circular corrective action The inspector noted that corrective actions precipitated by IE Bulletins or Circulars are manifested in* design changes or procedure modification In either case, the NRB requirement to review the minutes of SORC provides for review of these item In conducting this review, the inspector noted that two sets of minutes had not been provided for NRB members' review in a timely manne The minutes were promptly distributed when identifie The inspector had no further questions on this item at this tim (Closed) Unresolved Item (272/80-07-04) Restoration of redundant RWST level indication. The inspector verified that ~epair to the second level channel had been accomplished, restoring it to an operable status. The inspector had no further questions on this item..
(Closed) Unresolved Item (272/80-07-06) Alarm setpoint of channel R31A, letdown monito Design change lEC-0739 was completed which raised the R31A radiation monitor setpoints to lE+S (Warning) and 2.SE+S (High) CP Normal operation does not result in an alarm condition with these setpoint The inspector had no further questions on this ite Unit 1 3. Shift Logs and Operating Recotds a. The inspector reviewed the following plant procedures to detennine the licensee established requirements in this area in preparation for a review of selected 'logs and record AP-5~ Operating Practices~ Revision 9, April 23, 1979; AP-6, Operational.Incidents, Revision 6, February 22, 1979; AP-13, Control of Lifted Leads and Jumpers, Revision 4, February 11, 1980; Operations Directive Manual; and,
- AP-15, Tagging Rules, Revision 0, April 13, 197 The inspector had no questions in this are b. Shift logs and operating records were reviewed to verify that:
Contra 1 room 1 og sheet entries* are fi 11 ed out and i ni ti a 11 ed; Auxiliary log sheets are filled out and initialled; Log entries involving abnormal conditions provide sufficien detail to communicate equipment status, lockout status, cor-rection and restoration; Log book reviews are being conducted by the staff; Operating orders do not conflict with Technical Specification requirements; Incident reports detail no violation of Technical Specification LCO or reporting requirement; and, Logs and records were maintained in accordance with Technical Specifications and the procedures in 3.a abov The review included the following plant shift logs and operating records as indicated and discussiansr"wtth* ti:cenS:e~ :pet:SJonnel:
Log No. 1 - Control Room Daily Log, March 28 - April 29, 1980 Log No. 3 - Control Console Reading Sheet, March 28 - April 29, 1980 Log No. 6 - Primary Plant Log, March 28 - April 29, 1980 Log No. 7 - Secondary Plant Log, March 28 - April 29, 1980 Log No. 8 - Unavailable Equipment Status Log, March 28 - April 29, 1980
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Incident Reports - Numbers 80-73, 74, 78-86, 88-90, 92-96, 98, 101-105, 107-116, 118, 120-125 Tagging Requests - All active Night Orders, April 7-2lr 1980 Lifted Lead and Jumper Log - All active The following observations apply to logs and records reviewed during this reporting period:
On April 8, 1980 the inspector noted that the current valve lineup in the Deviation Book identified the suction and discharge valves for # 13 Charging Pump as cleared and tagged shu The valves were in fact open and the pump in service. The log was correcte A review of completed surveillance test documentation on April 24, 1980 identified that procedure SP(O) 4.. 5.4.1 which demonstrates Boron Injection Tank operability every 7 days, was not on fil The most recent surveillance filed was dated April 15, 198 Subsequent review of chemistry data verified that the surveillance had been conducted, however documentation had apparently not been completed.for the file. Appropriate documentation was prepare In reviewing outstanding tagging requests, the inspector noted that a red danger tag on valve 13LW16 had been temporarily released in accordance with procedure on April 7, 197 The tag was still re-moved and filed with the request on April 22, 198 The inspector expressed his concern over the long interval. permitted to elapse with a temporary release in effec No similar situation was identified and the tag replace During the course of this inspection, overhead alarm window D-37
{Upper Section Deviation) was inoperable.. Technical Specificatio.2.4.b requires that quadrant power tilt ratio be calculated every 12 hour~ with this alann inoperabl Log reviews indicated that the ratio was calculated at 1131 on April 12, 1980 and next at 0346 on April 13, 198 This interval exceeds the permitted 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, including the 25% extension provided in Technical Speci-fication surveillance requirements (272/80-06-05).
Additionally, during the course of this inspection, overhead alarm window D-40 (Rod Deviation and Sequence) was inoperable. Technical Specification 4.1.3.1.1 requires that, with this alarm inoperable, group positions of control rods be verified at least every 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Review of documentation, which consists of console reading sheets for rod position, indicated that during the following eight hour intervals, rod position was not verified, while the alarm was inoperable:
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February 2, 1980 1600 to 2400 February 3, 1980 0000 to 0800 February 16, 1980 0000 to 0800 February 16;. 1980 2000 to 0400 (2/17)
March 8, 1980-1600 to 2400 Apri 1 9, 1 980 0800 to 1600 Apri 1 1 0, 1980 1600 to 2400
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The above failures to conduct surveillance as required by Technical Specifications, constitute apparent noncompliance (272/80-06-03).
An identical item was identified in NRC Inspection Report 50-272/79-0 The inspector had no further questions relative to logs and records
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-- - Plant Tour During the course of the inspections, the inspector made observations*
and conducted multiple tours of plant areas, including the following:.
(1) Control Room (daily)
(2) Relay Room (3) Containment (4)
Auxi 1 i ary Building (5) Vital Switchgear Rooms (6) Turbine Building (7)
Yard Areas (8) Fuel Handling Building (9)
Radwaste Building (10) Penetration Areas (11) Control Point (12) Site Perimeter The following determinations were made:
Monitoring instrumentation: The inspector verified that selected instruments were functional and demonstrated parameters within Technical Specification limit Valve positions. The inspector verified that selected valves were iri the position or condition required by Technical Specifications for the applicable plant mod This verification included control board indication and field observation of valve position (Containment Spray and Auxiliary Feedwater System).
Radiation Control The inspector verified by observation that control point procedures and posting requirements were being followe The inspector identified no failures to properly post radi~tion a~d high radiation areas. During one tour of the radio-active waste areas, the inspector noted that the cage door to the enclosed waste*demineralizer area was not locked. The door is
- usually required to be locked since levels in the area frequently exceed 1 R/h At the time of this tour, the highest reading was less than 150 mrem/hr, permitt.ing control by REP, and not requiring a locki~g device. The licensee is treating the.event as a lapse in radiologicaLcontrols in order to prevent recurrence under less fortuitous condition Plant housekeeping conditions. Observations relative to plant housekeeping identified no*unsatisfactory condition Fluid Leak No fluid leaks were obs~rved which had not been identified by station personnel and for which corrective action had not been initiated, as necessar Piping vibratio No excessive piping vibrations were observed and no adverse conditions were note Selected pipe hangers and seismic restraints were observed and no adverse conditions were note Equ*i pment taggin The i rispector se 1 ected p 1 ant components for which valid tagging requests were in effect and verified that the tags were in place and the equipment in the condition specifie Control room annunciators. Selected lit annunciators were discussed with control room operators to verify that the reasons for them were understood and corrective action, if required, wa~ being take By frequent observation through the inspection, the inspector verified that control room manning requirements of 10 CFR 50.54 (.k) and the Technical Specifications were being me In addition, the inspector observed shift turnovers to verify that continuity of system status was maintaine The inspector periodically questioned shift personnel relative to their awareness of plant conditions and knowledge of emergency procedure Fire protection. The inspector verifi~d that selected fire extinguishers were accessible and inspected on schedule, that fire alarm stations were unobstructed, that cardox systems were operable, and that adequate control over ignitidn sources and fire hazards was maintaine **
Releas*e On a sampling basis, the inspector verified that appropriate documentation, sampling, authorization, and monitoring instrumentation, were provided for effluent release Technical Specifications. Through log review and direct observa-tions during tours, the inspector verified compliance with selected Technical Specification Limiting Conditions for Operatio The following parameters were sampled. frequently: accumulator levels and pressures, RWST level, BIT temperature, AFST level, rod insertion limits, containment pressure and temperature, valve power lockouts, axial flux difference. In addition, the inspector conducted periodic visual channel checks of protective instrumentatio Security. During the course of these inspections, observations relative to protected and vital area security were made, including access controls, boundary integrity, search, escort, and badgin No notable conditions were identifie Surveillance. The inspector observed portions of the following surveillance tests in progress; Service Water Pump Operability (Section XI testing).
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c. The following acceptance criteria were used for the above item Technical Specifications Operatings Directives Manual Inspector Judgement * The following specific comments apply to observations made during plant tours.*
On April 18, 1980, the inspector reviewed the control room strip chart record of average nuclear instrument power and the results of a calorimetric calculation. For a period of approximately two and one half hours, between 8:00 AM and 11:00 AM, the chart record indicates average NI power clearly in excess of 100%.
A calorimetric completed at approximately 10:04 indicated core power as 100.24%
(3346 MWT).
Shortly after this time, a power reduction to indicated 100% was accomplishe The reactor was operating in steady state conditions during this interval. A condition of Facility Operating License DPR-70 states, "Public Service Electric and Gas Company is authorized to operate the facility at a steady state reactor core power level not in excess of 3338 megawatts (one hundred percent of rated core power).
Continued steady state operation with the on-line power indication in excess of 100% constitutes apparent non-compli~n~e with a.condition of.the operating license {272/80-06-02).
An additional period of approximately one hour on April 21, 1980
was identified during which recorded average steady state NI power exceeded 100%.
--.On April 11, 1980, at approximately 1500, while conducting a tour*
of the Auxiliary Building, the inspector found manual isolation valve 14 AF 10 in the Auxiliary Feedwater System shut as indicated by stem position. The valve locking device (lock and chain) was in place and locked. This valve is in the flow path from the steam driven auxiliary feedwater pump (# 13) to # 14 steam generato The remaining manual isolation valves in the system were locked open such that flow from # 13 pump would have been available to *
the other three steam generators. Additionally, flow to all.steam generators would have been available from the two motor driven pumps (# 11 and # 12).
Review of control room status records failed to identify any re-cognition of this condition~ Log entries did not indicate im-position of an.Action 5-tatement, the Valve Deviation Book did not indicate the valve out of nonnal position (locked open), and the status board did not show an f.\\ction Statement of Technical Speci-fication 3.7.1.2 to be effectiv The most ~ecent operation inVol~ing the valve occur~ed during the evening of April 10, 1980 when an endurance run of# 13 Auxiliary Feedwater Pump was.conducted. Subsequent to this operation, a surveillance test to verify operability of the pump was conducted and a valve lineup completed which recorded valve 14 AF 10 as locked ope The inspector noted that Technical Specification 3.7.1.2 allows continued operation with one Auxiliary Feedwater Pump inoperable for a period up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. This item was discovered within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of initiation of the applicable Action Statement for the endurance tes Following identification of this discovery to the Senior Shift Supervisor, complete independent valve lineups of the Auxiliary Feedwater System were conducted by two individuals and no other problem identifie This finding was included as a subject of discussion at* a meeting held between NRC Region 1 and PSE&G management on April 18, 198 At this meeting, the licensee requested time to conduct a complete investigation of the circumstances and *to make appropriate recom-mendation The licensee committed to completion of the investi-gation and tran_smittal of_ the results to NRC by May 5, 198..
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The above failure to maintain equipment status as required by procedure constitutes apparent noncompliance with Technical Specification 6.8. 1 (272/80-06-01).
A similar j.tem was identi-fied in NRC Inspection Report 50-272/79-25 relative to Contain-ment Spray manual isolation valves llCS6 and 12CS The inspector had no further-questions relative to tours of the facilit **
5. * Calibration of Steam Generator Feedwater Flow Detectors/Indicators During the course of the inspection, the inspector performed independent calorimetric calculation One source of input data for this calculation is steam generator feed flo The inspector reviewed procedures and schedules established by the licensee for calibrating the steam generator feed flow.detectors and indicators used for performing calorimetric This review included the following:
Salem Generating Station Administrative Procedure 10 (Inspection Order System, Revision 3, dated October 2, 1979)
Performance Department Procedure PD-1.4.003 (General Instrument Calibration Procedure for Field Devices, Revision 3, dated October 3, 1975)
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Performance Department Procedure PD-1.4.005 (Instrument Data Calibration Sheet Instructions, Revision 0)
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Inspection Order Card (I.O. # 402309)
Instrument Calibration: Data Cards: Instrument No:
FL-8924, FL-8925, FL-8926 and.FL-8927 Type:
~ P Manufacturer:
Rosemount Nomenclature:
- 11, 12, 13, 14 Local Feedwater Flow DP Transmitter Instrument No:
FL-8924-I, FL-892~-I, FL-8926-I and FL-8927-I Type:
Digital Indicator Manufacturer: Fairchild Nomenclature:
- 11, 12, 13, 14 Local FW Flow
~p Indicator Inspection Order Request 402309 (dated December 6, 1977)
Inspection Order Request Form 402309 (dated December 11, 1979)
Instrument Manuals Rosemount Instrument Manual 4256/4257 Fairchild Instrument Book:
Model 53/80 Calorimetric Calculation Procedure (Reactor Engineering Manual, Part Revision 10, dated September 10, 1979)
Unit 2 A review of the Inspection Order Requests and the Inspe.cti on Order Cards identified the detectors and indicators as requiring monthly calibratio A review of the data cards indicated the calibration period had been exceeded twice. Specifically, the detectors and indicators had not been properly calibrated within one month prior to resuming power operation
in December 1979 and remained overdue until February 4, 1980. Subsequent calibration was not conducted until April 16, 198 The data recorded in columns labeled "AS FOUND" and "AS LEFT" indicated adjustments were necessary during both calibrations to bring the output within the manufacturer's specified accuracie *
- The procedure does not provide for maintaining the instruments within required accuracy but only requires periodic calibration. This condition pennits use of instrument having exceeded their required accuracy but still not due for calibration. There is no procedural guidance calling for corrective action relative to manufacturer's stability or drift specifications or observed instrument drift. Inspection of the Instrument Data Cards revealed that several infonnation items were not provided
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nor marked 11NA 11 as required by Perfonnance Department Procedure (P:O 1.4.005).
ln _aggregate, the above findings*constitute apparent noncompliance with 10 CFR 50, Appendix B, Criterion XII, Control of Measuring and Test Equipment (272/80-06-04).
- A review of the Calorimetric Calculation Procedure (Reactor Engineering Manual, part a, Revision 10 dated September 10, 1979) indicated that there were no provisions to ens.ure that input data is obtained from calibrated instruments. This item is open pending licensee review of procedures and appropriate modifications to ensure input data for safety related activities is valid (272/80-06-06):
Plant Tour a. The inspector conducted periodic tours of all accessible areas in the plant. During these tours, the following specific items were evaluated:
Hot Wor Adequacy of fire prevention/protection measures use Fire Equipmen Operability and evidence of periodic inspection of fire suppression equipmen Housekeepin Minimal accumulations of debris and maintenance of required cleanness levels in systems under or following testin Equipment Preservation. Maintenance of special preservative measures for installed equipment as applicable Component Taggin Implementation and observance of equipment tagging for safety or equipment protectio..
Maintenance. Corrective maintenance in accordance with esta-blished procedure Instrumentation. Adequate protection for installed instrumen-tatio Cable Pulling. Adequate measures taken to protect cable from damage while being pulle Corrununicatio Effectiveness of public address system in all areas of the sit Equipment Controls. Effectiveness of jurisdictional controls in precluding unauthorized work on systems in test or which have been teste Log Completeness of logs maintained and resolution of identi-fied problem Foreign Material Exclusion. Maintenance of controls to assure systems which have been cleaned and flushed are not reopened to admit foreign materia Security. Implementation of security provisions. Particular attention to maintenance of Unit 1 protected area boundar Testing. Spot-checks of testing in progress were mad The inspector had no questions relative to plant tours.
7~ Operational Readiness 10 CFR 50.57 states that the issuance of an operating license is, in part, contingent upon a finding that construction of the facility has been substantially completed, in conformity with the construction permit and the application, as amended, tbe provisions of the Act, and the rules and regulations of the Commissio In order to provide a basis for this finding, the inspector is conducting a continuing review of licensee readiness to operate the facility. This review includes, but is not limited to, the following areas:
Completion of the NRC inspection program to assess construction, testing, and operational preparednes Status of facility operating procedures and personnel trainin Status of all enforcement items and unresolved matter Status of the preoperational test progra Status of construction activitie Proposed facility Technical Specification **
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Review of licensee outstanding items, particularly those identified for completion or resolution after core loa Implementation of corrective measures to Unit 2 as a result of items identified in Unit 1 for Reportable Occurrences, inspection findings, and IE Bulletin and Circular Participation in a Commission meeting held April 16, 1980 relative to licensing this uni Facility Operating License DPR-75 was issued for Salem Unit 2 on April 18, 1980, authorizing core load and low power testing to 5% RTP, subject to the conditions therei.
IE Bulletin and Circular Followup The IE Circulars discussed below were reviewed to verify that they had been reviewed for applicability by cognizant management, and appropriate action initiated~
The review included discussions with licensee personnel and observation and review of items discussed in the details belo *
79-05 - Moi s.ture Leakage in Stranded Wire Conductor Licensee review of this item is documented in a memorandum dated April 23, 1980. *
The conclusion is made that, by design, differential pressures to cause moisture leakage is not expected at the Salem facilities *
79-08 - Attempted Extortion - Low Enriched Uraniu Appropriate action on this item was taken by cognizant licensee personne Regulator Diaphragms in Certain Self Contained Breathing Apparatu The licensee did not use either of the SCBA devices discussed in this Circular at the time of issuance. Applicability to the SCOTT Units now in place at Salem will* be reviewed. This is an open item (272/80-06-07).
79-10 - Pipefittings Manufactured from Unacceptable Materia The licensee has conducted a record search to verify that none of the suspect fittings have been used at either Salem facilit None were identifie Design/Construction Interface Proble At the time of receipt of this Circular, Salem Unit 2 construction and preoperational
- testing was virtually complete and the considerations discussed no longer applicabl Potential Diesel Generator Turbocharger Proble The problem is identified for EMO diesels and is not applicable to the Alco units installed at Salem Units 1 and Diesel Fire Pump Starting Contactor The problem identified with Cummins contactors is not applicable to the Waukesha diesels used at Sale..
79-15 - Bursting of High Pressure Hose and Malfunction of Relief Valve 110 11 Ring in Certain Self Contained Breating Apparatus. The SurvivAir Mk 1 SCBA is not used at Sale Proper Installation of Target-Rock Safety-Relief Valves. Target-Rock valves are not used at Sale Loose Locking Devices on Ingersoll-Rand Pump Impellers. In order to.preclude the potential problem in relation to I-R pumps installed at Salem, action has been initiated to procure 11retro-fit 11 kits from the manufacture Failure of GTE Sylvania Relay...* By memorandum dated December 7, 1979, the licensee has documented a review which indicates that tne subject relay is not used at Sale ~~21 - Prevention of Unplanned Releases of Radioactivit As documented tn Design Memorandum SGS/M-DM-0125 dated November 20, 1979, the ltcens.ee has initiated a design review to address the potential release paths outlined in the Circular. The memorandum also addresses the concerns expressed in *NUREG-0578 on this subjec Stroke Times for Power Operated Relief Valve The licensee includes the power operated relief valves as Category B valves in the ASME Section XI test program required by Technical Specification 4.0.5. Accordi_ngly, the valves are stroke-timed every three month. - Motor Contactors and Starters Failed to Operat The licensee verified that none of the subject Gould starters or contactors are used at either Salem facilit Proper Installation and Calibration of Core Spray Pipe Break Detection Equipment on BWR While the licensee was included as an addressee for this item, the concern does not appear to be applicable to PWR Shock Arrestqr Strut Assembly Interference. The licensee's review of this item indicates that no Bergen-Patterson Strut Assemblies described in the Circular are employed at either Salem Uni Service Advice for GE Induction Disc Relay Based on prior information from GE, the licensee had completed cleaning of IAC type relays installed at Sale Nuclear Power Plant Staff Work Hours. These guidelines are included as a condition of the Salem Unit*2 Operating License as they apply to licensed operators. The licensee is currently developing procedures to incorporate the guidelines for remaining staff personne *
The inspector had no further questions relative to Circulars reviewe. Administrative Controls on Containment Isolation A design deficiency had been detected at Salem Unit 1 which would have prev~nted containment ventilation isolation on a safety injection signal
- if ventilation isolation due to :-r.adiation monitors in alarm had been previously bypasse Identification of this problem and a similar situation at Millstone Unit 2 precipitated an NRC generic letter en-titled, "Containment Purging During Normal Plant Operation.
The letter requires, in part, that pennanent administrative controls be inaugurated to prevent, during purging, improper manual defeat of safety actuation signal Licensee corrective action has included the following; administrative tagging controls to preclude operation of containment purge valves during Mode 1-4 operation, procedural requirements (OI II-16.3.1 -
Containment Ventilation Operation, Revision 7, OSC through P-4) to prohibit operation of pressure vacuum relief valves unless all radiation alanns are cleared, and completion of a design change (OCR 1-EC-430)
which prevents reset of radiation monitor signals from also removing safety injection isolation signal Additional details relative to this 'item are contained in NRC Inspection Reports 50-272/78-26 and 50-272/79-15. Licensee Event Report 50-272/79-11 also refer The inspector had no further questions on this ite. Westinghouse 17 x 17 Control Rod Rodlets As a result of several failures detected at the spider finger/control rodlet connection, Salem Unit 1 replaced 25 Rod Cluster Control Assemblies (RCCA) during the summer 1979 refueling outage. Details are found in NRC Inspection Report 50-272/79-1 The failures were traced to receiving
- lots of spider finger production material and are not expected to pose a generic problem with Westinghouse 17 x 17 cores. Guidance has been provided to Westinghquse owners by the vendor relative to finger failure detection through the use and analysis of in-core flux map The Westinghouse letter was received by PSE&G on September 17, 1979, and has been incorporated into station procedures (Reactor Engineering Manual, Part 12, Appendix H, Fl~x Mapping Procedure, Revision 4). Five percent differences in measured to calculated reaction rate integrals are flagged for evaluatio All flux maps are evaluated and are performed at least every 31 full power days in accordance with Technical Specification requirement No dropped rods have been detected in flux maps compared during Cycle: The inspector had no further questions in this are *
11.. Unresolved Items Areas for which more information is required to determine acceptability are considered unresolved. Unresolved items are contained in Paragraphs 5 and 8. of this repor *
12 *.. Exit Interview At periodic intervals during the course of this inspection, meetings *
were held with senior facility management to discuss inspection scope and finding A special management meeting was held at the site on April 18, 1980 to discuss the item in paragraph 4d of this repor *