IR 05000293/1986021
| ML20212P983 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 08/25/1986 |
| From: | Kister H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20212P953 | List: |
| References | |
| TASK-2.K.3.13, TASK-TM 50-293-86-21, NUDOCS 8609030356 | |
| Download: ML20212P983 (26) | |
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U. S. NUCLEAR REGULATORY COMMISSION
REGION I
Docket / Report No. 50-293/86-21 License: DPR-35 Licensee: Boston Edison Company 800 Boylston Street Boston, Massachusetts 02199 Facility: Pilgrim Nuclear Power Station Location: Plymouth, Massachusetts Dates: June 2, 1986 - July 7, 1986 Inspectors: M. McBride, Senior Resident Inspector J. Lyash, Resident Inspector G. Nejfelt, Resident Inspector Reviewed by: J. S osnider, Chief, Reactor Projects Section 18 Approved by: -
Harry B. $jJster, Ch'tef 2 h
/ Dat'e ~
Reactor Projects Branch No. 1
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Areas Inspected: A routine resident inspection was conducted of the control room, accessible parts of plant structures, plant operations, radiation pro- e tection, physical security, plant operation records, plant events, maintenance, y and surveillance. The inspection totaled 346 hours0.004 days <br />0.0961 hours <br />5.720899e-4 weeks <br />1.31653e-4 months <br /> by three resident inspectors. I Results: One violation was identified regarding use of an inadequate battery >
test procedure (paragraph 3.e). An additional deficiency with the battery test procedure is also discussed in paragraph 3.e. The latter deficiency made the results of a battery discharge test invalid, requiring a repeat test and the repeat test put unnecessary wear on a safety-related battery. Potential techni-cal inadequacies in several ECCS surveillance tests required by the technical specifications are noted in paragraph 3.e. The failure to promptly submit a response to a special NRC team safety system inspection (conducted in byember 1985) is noted in paragraph 2. Also in paragraph 2 are (1) a concern about inadequate training regarding compensatory operator actions for non-seismically qualified safety equipment, (2) an unresolved item concerning a potential vio-lation of 10 CFR 50, Appendix R training requirements for the station fire brigade, (3) a concern about the identification and inspection of fire barriers at the station, and (4) the failure to properly track and implement an NRC ,
commitment to TMI Action Plan item II.K.3.13. The missed commitment did not I affect the acceptability of licensee action on the item, but reflects a weak-ness in commitment control. Recurring failures of the recirculation motor generator set field breaker used for anticipated transient without scram protection is discussed in paragraph 4. Poor contamination control practices by maintenance supervisors are discussed in paragraph 6. This is a recurring
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8609030356 860825 PDR ADOCK 05000293 G PDR
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Inspection Summary (Continu(1) 2 proble Improperly scheduled surveillance tests were identified by the licensee during the inspection period (paragraph 7). This problem was not reviewed during this inspection. Finally, the failure of the primary contain-ment to meet established leak rate criteria based on local leak rate test results is presented in paragraph Licensee compensatory actions for several labor strikes during the inspection period were considered well planned and +horoug +,
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TABLE OF CONTENTS Page Summary of Facility Activities .............. 1 Followup on Previous Inspection Findings . . . . . . . . . . 1 Violations, Unresolved Items, Inspector Follow . . . . . . . 1 Items, TMI Action Plan Items . . . . . . . . . . . . . . . . 9 Routine Periodic Inspections . . . . . . . . . . . . . . . 10 Daily Inspections, System Alignment Inspection, . . . . . 10 Biweekly Inspections, Plant Maintenance and .......11 Surveillance Testing . . . . . . . . . . . . . . . . . . . 14 Review of Plant Events . . . . . . . . . . . . . . . . . . 16 Operators, Maintenance and Clerical . . . . . . . . . 16 Worker Strike Failure of 125 VDC Distribution Panel "C" . . . . . 16 Auto Transfer Switch "A" Recirculation Motor Generator Set Field . . . . . 16 Breaker Failure Residual Heat Removal Loop B Flow Anomolies . . . . . 17 Strike by Contract Security Force . . . . . . . . . . 18 Main Stack Sample Canister Particulate Filter . . . . 19 Missing
, Observation of Physical Security . . . . . . . . . . . . . 19 1
' Radiation Protection . . . . . . . . . . . . . . . . . . . 19 Master Surveillance Tracking Program Review . . . . . . . 21 Leak Rate Testing Progress / Observations . . . . . . . . . 22 Survey of Licensee Response to Selected Safety Issues . . 23 (TI 2515/77)
1 Review of Licensee Event Reports (LER's) . . . . . . . . . 23 11. Public Meetings and Meetings with Local Organizations . . 24 12. Management Meetings . . . . . . . . . . . . . . . . . . . 24 Attachment 1 - Persons Contacted Attachment 2 - Summary of LLRT Status
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DETAILS 1.0 Summary of Facility Activities The plant was shut down on April 12, 1986 for an unscheduled maintenance outage. Subsequently, the NRC issued Confirmatory Action Letter 86-10 which required that the licensee seek approval from the NRC Regional Administrator prior to reactor startup. The outage continued throughout the current inspection perio On June 13, 1986 the licensee identified that several surveillance tests including Local Leak Rate Testing (LLRT), degraded voltage trips and stand-by liquid control testing were past due. The licensee has begun perfor-mance of a LLRT test program to be followed by conduct of an Integrated Leak Rate Test (ILRT).
On June 12, 1986 the licensee-employed labor unions ended the strike which had begun on May 16, 1986. On June 30, 1986 the plant security force re-jected a contract offer from the onsite security contractor and went on stri ke. The secur';y workers contract dispute was resolved and the strike
! ended on July 1, l' 36. A strike by laborers against the chief contractor,
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Bechtel, continued curing the inspection perio .0 Followup on Previous _ Inspection Findings Violations (Closed) Violation (84-23-01), Failure to measure and record station 1 battery cell specific gravity. The inspector reviewed PNPS Procedure 1 8.9.8, Battery Rated Load Discharge Test, Revision 8. The procedure had (
recently undergone a major revision. The inspector verified that the pro- '
cedure requires specific gravity and voltage measurements of all cells prior to test start, subsequent to completion of the battery discharge but prior to recharge, and after completion of battery recharging. The inspector also witnessed performance of the procedure as described in section 3 of this report. This item is close Unresolved Items (Closed) Unresolved Item (81-19-04), reissue procedure 1.3.14 and revise training manual to include a list of effective pages. Concerns identified by the inspector included the uncontrolled status of the Training Manual and the inadvertent retirement of a quality related procedure. The licen-see has issued a controlled document entitled Index of QA Related Proce-dures to 10 CFR 50 Appendix B Criteria. This index identifies those pro-cedures which are necessary to implement the Quality Assurance Progra Nuclear Operations Department (N00) Procedure 1.3.4, Procedures, Revision 29, requires Quality Assurance Manager approval prior to retirement of any quality related procedure. The reference index in conjunction with this approval requirement provides assurance that inadvertent retirements
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will be precluded. The inspector reviewed N00 procedure 1.3.14, Training, Revision 13 and verified that the procedure had been reissued. The in-spector verified that the Pilgrim Nuclear Power Station Training Manual has been designated as a controlled document. A list of revised pages and discipline engineer concurrence with any changes has been include The inspector had no further questions. This item is close (Closed) Unresolved Item (83-22-02), verify completion of corrective actions in response to QA audit 83-11. The inspector reviewed the scope and results of Quality Assurance audit 83-11, Procurement Process, and discussed these findings with Quality Assurance personnel. The inspector examined the nineteen deficiency reports (DR) generated to track audit findings, verified that corrective actions had been taken in a timely manner and that Quality Assurance had reviewed these corrective actions for acceptability. The inspector also evaluated the actions taken in response to a sample of these DRs to verify that findings had been thoroughly addresse The inspector had no further questions. This item is close (Closed) Unresolved Item (83-22-03), verify that inspection procedure is issued, adequate and effectively implemented. The licensee has transferr-ed material from the Bechtel warehouse to the new materials storage faci-lity. Measures were developed to control the care, maintenance and test-ing of items stored in the new storage facility. No such measures had been established during Bechtel warehouse us The inspector questioned the evaluation process used to determine that material transferred from the Bechtel warehouse had been adequately maintained while in storage and was still fit for servic In response to this concern the licensee issued t Stores Department Procedure 7.02, Pre-Issue Inspection. This interim pro- T cedure was used to assess items prior to issuance for use. It included a (
visual inspection of material to assess physical damage and shelf lif Inspections conducted under this procedure were documented, with copies transmitted to QC for review. Maintenance of in storage material is cur-rently controlled and detailed by Stores Department Procedure 13.06, In-storage Maintenance Program. This item is close (Closed) Unresolved Item (83-22-04), verify in-storage maintenance proced-ures are issued and effectively implemented. The inspector discussed the in-storage maintenance program with stores personnel, toured the onsite storage facility, and reviewed stores department Procedure 13.06, In-storage Maintenance Program, Revision 3. Storage and maintenance require-ments are established by the nuclear engineering department and implemented by the stores department personnel using procedure 13.06. Upon receipt of equipment applicable requirements are assigned, documented and input to a computer management system. Maintenance action cards are issued to the stores department fcr implementation on a monthly basis. As maintenance is performed the computer tracking system is updated. The licensee appears to have implemented a program to ensure that equipment in storage receives the attention needed. This item is close :
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(Update) Unresolved Item (86-01-03), review licensee evaluation of the use of fuses and metal links. During the period the inspector reviewed the licensee response to IE Circular 77-09, Improper Fuse Coordination in BWR Standby Liquid Control System (SLCS) Circuits. This circular discusses fuse coordination between the SLCS main control power circuit fuses and the two explosive valve detonator circuit If a detonation circuit fault occurs, the detonator protective fuses must open so that the main control
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power circuit remains energized. The inspector reviewed controlled drawing MIF4-9, revision E6, Elementary Diagram Standby Liquid Control. This drawing indicates that 2 amp fuses are installed in series with the explo-sive valve detonators, and a separate fuse is installed in the main con-trol power circuit. A control power circuit fuse rating was not included on the drawin Drawing note 3 states that coordination between fuses is required, but does not detail required fuse design. The inspector, accom-panied by a licensee representative, verified that solid links had been installed in the field to replace the main control power fuse. The inspec-tor pointed out that while installation of a solid link resolved the fuse coordination problems this should be accurately reflected in controlled design documents to ensure that it remains as require The licensee stated that Engineering Service Request 86-019 had been initiated to identify incorrect fuse vs. solid link ccnfigurations, and drawing discrepancies. This ESR requests Boston Edison engineering to 1)
provide basis for fuse / solid link applications 2) provide operability justification for each case 3) review any inconsistencies and correct drawings where needed. This effort is ongoing. The inspector will review its conclusions during a future inspectio (Update) Unresolved Item (86-14-01), Operation with unqualified diesel $+
generator lockout rela In inspection report 86-14 it was stated that E on April 25, 1986 one of the three coils associated with the differential relay failed, initiating the generator breaker trip. Subsequent investiga-tion by the inspectors identified that the event was actually initiated when a licensed operator closed the compartment door causing a spurious trip of the differential relay. Damage sustained by the differential relay was a result of the failure of the lockout relay to actuate. The actual initiating event demonstrates, to some degree, the sensitivity of the GE 12CFD differential relay. The licensee indicated that a system operability evaluation will be complete by July 31, 198 The inspectors reviewed the training material provided to operations personnel regarding the actions required if a spurious relay operation occurs. Training material stated that a non-seismically qualified annunciator would alert the operator to a spurious trip during a seismic event. The inspector questioned the practice of instructing use of non-seismic instrumentation during a seismic event. The licensee stated that no other instructions advising similar dependence on unqualified instrumentation had been given.
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Training material specified that in the event of a generator breaker trip, reset of the lockout relay and the voltage reset would be require In fact voltage reset is required only if the diesel generator is running as a result of a manual start. Automatic starts are not affected by this feature. The licensee, in response to the inspectors questions, issued a revised training memorandum clarifying and correcting the discrepancie Inspector Follow Items (Update) Inspector Follow Item (85-31-02), Review licensee evaluation of potential problem with RHR pump impeller wear rings. The inspector questioned the licensee regarding this issue prior to the licensee's receipt of NRC Information Notice 86-39, " Failures of RHR Pump Motors and Pump Internals". The initial licensee response was that the material used in the Pilgrim pumps was not the same as that used at Peach Botto Further discussions with licensee personnel indicated that the RHR pumps and pump impeller wear rings installed at Pilgrim are very similar to those which failed at Peach Bottom. Licensee plans at that point were to dis-assemble and inspect one or more pumps during the upcoming refueling out-age. The outage was originally scheduled for October 1986, al' wing ap-proximately ten months of operation after the potential proble.- had been identified. The inspector expressed concern to licensee management re-garding the timeliness of the inspections, in light of the scope of the Peach Bottom problem, and the similarity of the Pilgrim pumps. Or. . June 20, 1986 during a meeting between Boston Edison and NRC: Region I senior management, the licensee was questioned regarding the proposed RHR pump inspection schedule. This management meeting is briefly described in section 12 of this repor In response, and after further evaluation, the +
licensee made the decision to disassemble and inspect all four RHR pumps kL prior to restart. The inspector will continue to follow the licensee action (Closed) Inspector Follow Item (81-24-09), Review implementation of the fire brigade training; review recommendations for improvement in the fire protection program. Performance Appraisal Inspection 81-20 identified three specific fire brigade training weaknesses under item 15 of that report. These weaknesses were 1) requirement that the Nuclear Operations Manager review the training of the fire brigade was not implemented, 2)
no provision to assure that problems identified during a drill were cor-rected, 3) training exercises were being counted as brigade drills in order to meet minimum drill requirements. These concerns were identified for followup in report 81-24, and updatea in report 83-0 The licensee issued Nuclear Organization Procedure (NOP) number NOP83FP1, PNPS Fire Protection Program, in February,1983. This document describes the overall fire protection program at Pilgrim, and supersedes any previously issued pla r
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Responsibilities of the Nuclear Operations Manager as well as all other individuals involved in program implementation have been redefined. PNPS Procedure 1.4.23, Revision 7, Fire Brigade Training, requires issuance of action requests to correct deficiencies identified during drills. In ad-dition copies of drill results are forwarded to the Resources Protection and Control Group Lead and the Chief Operating Enginee CFR 50 Appendix R requires that each fire brigade member participate in at least two drills per year. Discussions with the licensee indicate that no method of tracking has been established to ensure that this re-quirement is me Further it appears that not all members currently meet this requirement. Other requirements for unannounced and backshift drills also may not be met. Similar observations were noted in inspection report 83-04, section 7.3. This item is considered unresolved pending further evaluation. (86-21-01)
Based on the above, Inspector Follow Item 81-24-09 is close (0 pen) Inspector Follow Item (83-06-01), review licensee actions to assure other fire penetrations are not m: aed. On March 17, 1983 the licensee discovered several fire barrier p-netrations not identified by the 1977 fire protection review or the 1980 survey conducted by Insulation Consul-tant and Management Service (ICMS). The licensee's engineering department reviewed the scope and results of the ICMS survey and determined that the area in question had not been includeJ in the survey. A crosscheck of the ICMS penetration survey against the required fire boundaries and station surveillance procedures was conducted. This crosscheck identified several other areas not included in the ICMS survey, and several drawings not in- 5 cluded in the station procedures. Welkdowns of areas not previously sur- (
veyed were conducted by engineering, corresponding ICMS survey sheets were I
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corrected, and this information was supplied to the station via memorandum NED 84-46 Discussions with the licensee fire protection engineer indicate that a procedure rewrite effort conducted during 1984 should have made these ad-dition The licensee had not supplied the inspector with information on the scope of this effort prior to the end of this report period. The li-censee further stated that as the fire barrier surveillance procedures are implemented and additional penetrations identified, they are added to the procedure. The inspector questioned the confidence which can be placed in the present list of fire barrier penetrations, given that additional penetrations are identified during walkdown The inspector noted that as new penetrations are created, as a result of the on going appendix R modifications, the responsibility for update of all procedures / drawings lies with the fire protection engineer. This re-sponsibility, in conjunction with other assigned duties, appears excessive given the resources availabl This was demonstrated by the apparent back-log of new penetrations identified by the fire protection engineer which had not been incorporated. The licensee stated that plans for a contrac-tor fire protection program audit were in place. "atails of audit method and scope could not be given at that tim : :
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9 l This item remains open pending 1) further review of actions taken to as-sure all existing fire barrier penetrations are identified, 2) further review of the system to ensure that new penetrations are promptly incor-porated into applicable procedure (0 pen) Inspector Follow Item (86-14-02), cracked MSIV return springs. The licensee stated that the Atwood'-Morrel main steam isolation valve (MSIV)
springs would be visually inspected for cracks on the inner and outer springs, during this outag If cracks are found, all the springs will be tested with a 5% overload test; and the cracked springs would be re-placed. These actions are in keeping with the recommendations of General Electric to the licensee on June 19, 1986. It was later determined by the licensee that these MSIV springs were visually inspected on June 7, 1986, using maintenance report (MR) 86-1-34, and no cracked springs were foun Additional visual inspections by QA are planned in the futur This concern will be re-examined in a future inspection repor (0 pen) Inspector Follow Item (85-30-01), Review Implementation of Revisions tu . ocedures, and justification to exceed 7 day outage of compartment coe'ers. Inspection 85-30, Safety System Functional Inspection, was a i
special team inspection conducted by the Office of Inspection and Enforce-ment du.-ing November 1985 (report issued January,1986). Approximately 15 items sere identified for post inspection followup. The inspector examined the licensee's response to item 85-30-01 during this period and noted that while some areas of concern had been addressed, not all action had been completed. The inspector also noted that no licensee response to the in-spection had been written and submitted. The inspector questioned the <
licensee as to the status of the response and the reasons for its long i delay. The licensee stated that. sections of the response had been drafted !
but that no target date for its issuance had been se TMI Action Plan Item l (Closed) Unresolved Item (82-10-08), TMI Action Plan item II.F.1.3 Contain-l ment High Radiation Monitor, deviates from NUREG 0737. The inspector iden-tified several discrepancies between NUREG 0737 Item II.F.1.3 requirements and the design and installation developed by the licensee. Inspection report 82-13 cited an item of noncompliance for the licensee's failure to control the design process in this area. Corrective actions regarding the design control process in general, and item II.F.1.3 in particular, were reviewed in inspection report 84-11. NRC:NRR review and acceptance of the licensee's containment high radiation monitoring system is documented in a safety evaluation provided to Boston Edison on May 13, 1983. Followup inspection to verify proper implementation of the accepted design is docu-mented in inspection report 85-27. Based on the above described reviews /
inspections, this item is closed.
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(Closed) TAP Item II.K.3.13, RCIC Automatic Restart - Item II.K.3.13 of NUREG 0737 required 1) separation of HPCI and RCIC initiation levels and 2) modification of the RCIC initiation logic to allow automatic restar General Electric, on behalf of the BWR Owners Group, submitted generic evaluations and modifications addressing these items. The NRC staff's safety evaluation of the submittals was provided to Boston Edison by memo on March 16, 1983. Concerning the separation of HPCI and RCIC initiation levels the staff concurred in the GE position that no significant benefit would be gained by implementation of the modification. Concerning RCIC automatic restart on low level, the staff found acceptable the proposed modification relocating the existing high water level trip from the turbine trip throttle valve to the steam supply valve. This acceptance was based on licensee's meeting certain plant specific acceptance criteria. On April 29, 1983 the licensee transmitted BEC0 letter no.83-108 stating that all required criteria had been met, including successful completion of a system functional test. The inspector reviewed applicable system elementary diagrams and discussed system operation with licensed operations personnel to ensure that the described modifications had been implemente The inspector verified, through discussion with the NRR licensing project manager, that no open safety evaluation issues remain. This item is closed.
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in BECO letter 81-37 to the NRC the licensee stated that in order to pro-vide RCIC remote restart capability following other protective trips a motor operator would be installed on the turbine trip throttle valve.
r This modification was scheduled for implementation by July 1, 1982. Sub-sequently, in BECO letter 82-233 dated September 1, 1982, the licensee stated that the modification is desirable but that the July 1 date would not be met. No alternate date was specified. This modification was not a necessary to satisfy the requirements of TAP Item II.K.3.1 The inspec- '
tor pointed out that this motor operator had not been installed and ques- 0 tioned Boston Edison licensing personnel regarding its status. The li- !
censee stated that the item had apparently been overlooked. No decision had been made to cancel the improvement. The licensee indicated that the proposed modification would be reevaluated and a decision made as to its implementatio .0 Routine Periodic Inspections Daily Inspection During routine facility tours, the following were checked: manning, access control, adherence to procedures and limiting conditions for operations (LCO's), instrumentation and recorder traces, control room annunciators, safety equipment operability, control room logs and other licensee documentatio No unacceptable conditions were identifie __ .. _ _ _ _
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11 Systems Alignment Inspection The operability of selected piping system trains was routinely assess-ed. Major motor operated and manual valve positions for safety equip-ment were verified during routine checks of the control room. Valve power supply, breaker alignment, and safety equipment controller set points were also checke The inspector walked down accessible portiens of the High Pressure Coolant Injection (HPCI) system piping, instrumentation and electri-cal equipment. Major manual and automatic valve, instrumentation valve and electrical component positions were verified to be correct and consistent with applicable operations procedures. The inspector observed the general condition of equipment and areas. On July 3, 1986 during a tour of the HPCI pump / turbine area the inspector noted several inches of an oil / water solution accumulated around the tur-bine pedestal. The inspector informed the control room operators of the situation and a chemistry technician was dispatched to sample the liqui Results showed a solution of eight percent oil and 92 per-cent wate The fluid was removed and the area cleaned Operations personnel stated that investigation to determine the ca se of the leak would be conducte '
In addition to in plant walkdowns the inspector conducted j review of the surveillance tests used to satisfy HPCI technical spacifica-tion requirements. Detailed discussion of the findings identified by the review are included in paragraph 3.e below. The inspector also completed inspection of Temporary Instruction (TI) 2515/77, Survey of *
Licensee Response to Selected Safety Issues. This TI requires review h of issues affecting overall HPCI system reliability. Documentation h of this effort is included in paragraph 9 of this repor c. Biweekly Inspections During plant tours, the inspector observed shift turnovers and checked:
plant conditions, valve positioning and locking (where required), in-strumentation lineup, radiological controls, security, safety, and general adherence to regulatory requirements. Plant housekeeping and cleanliness were evaluate d. Plant Maintenance The inspector observed and reviewed maintenance and problem investi-gation activities to verify compliance with regulations, administra-tive and maintenance procedures, codes and standards, proper QA/QC involvement, safety tag use, equipment alignment, jumper use, person-nel qualifications, radiological controls for worker protection, fire protection, retest requirements, and reportability per Technical Specification :
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On May 25, 1986, during reassembly of Main Steam Isolation Valve (MSIV) A0-203-10 a worker mistakenly installed the valve stem to actuator coupling plate upside down. This component, designated as part 14M on drawing 2518-2-6 SH1, is part of the connection between the valve poppet stem and the actuator lower plate. It also inhibits the valve poppet stem from rotating and becoming disengaged from the actuator. The seven other MSIVs were inspected and it was found that the component had been incorrectly installed on two other valve This condition was corrected. Licensee followup identified that pro-cedure 3.M.4-8, Main Steam Isolation Valve Maintenance did not pro-vide sufficient reassembly guidance. The procedure was revised with appropriate references and cautions. The inspector reviewed revision 12 of procedure 3.M.4-8 and had no further question e. Surveillance Testing Surveillance Test Observation
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Observation of 125 VDC and 250 VDC Battery Rated Load Discharge Test On July 1,1986, the inspector witnessed portions of procedure 8.9.8, Revision 8, Bettery Rated Load Discharge Test, for the
"B" 125 VDC battery cerformed under maintenance request (MR)
86-46-176, and the 2,0 VDC battery performed under MR 86-46-17 Testing involved; 1) isolation of the battery bank to be tested, 2) connection of a test load bank and discharge of the battery, *
3) isolation of the battery charger from the distribution bus, kL and 4) connection of the charger directly to the battery for post test recharging. To allow isolation of the "B" 125 VDC battery and connection of the test load bank, cables between the load limiting switch and output breaker had been lifted. At the conclusion of the "B" 125 VDC battery discharge test the con-nections to the test load bank were removed. System alterations were then made to allow recharging of the battery directly from the charger, without passage through the bus. This was accom-plished by disconnecting cables between the "B" 12S VDC battery charger and the bus, and installation of a jumper between the charger and the load limiting switch terminals. Similarly during performance of testing on the 250 VDC battery, cables between load limiting switch and battery output breaker were lifted to allow test load bank installatio Procedure 8.9.8 did not give detailed instructions for lifting the cables and installing the temporary jumpers, or removing of the jumpers and relanding the cables. Attachment A, step 8 states only to connect proper load and test equipment to the batter Restoration steps were also not included. Discussion with the watch engineer indicated that he had not been consulted regarding these alteration '
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The inspector reviewed Engineering Response Memorandum NED 86-495,
referenced by procedure 8.9.8. This evaluation states that con-necting both busses to a single battery during the test is ac-ceptable because selected loads will be removed from the busses to limit surge loading on the single battery. However, on July 1, 1986 the 125VDC busses had been connected to a single battery during testing, but the requirement to strip certain excess loads as specified in NED 86-495 had not been included in the procedure and was not implemente Pilgrim technical specification 6.8.A specifies that written procedures be established and implemented, that meet or exceed the requirements of ANSI N18.7-1972, Sections 5.1 and 5.3. ANSI N18.7-1972, Administrative Controls for Nuclear Power Plants, Section 5.3.1 states that procedures shall be sufficiently de-tailed for a qualified individual to perform the required func-tion without direct supervision. Section 5.3.5 further states that special attention shall be given to restoration of normal conditions and that all jumpers will be controlled. The inspec-tor informed the licensee that failure to address the lifting of cables, placement of jumpers and prerequisites needed for inter-
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tie of the busses, in procedure 8.9.8, revision 8, constitutes a violation of technical specification 6.8.A. (86-21-02).
In response to the finding the licensee added instructions for the lifted cables and jumpers to the procedure, appropriate main-tenance requests and test control tags were placed in the fiel Procedure 8.9.8, Attachment A, Step 14 specifies 250 VDC battery minimum terminal voltage as 210 volts. Step 16 directs test i personnel to end testing when actual terminal voltage reaches '
the specified minimum 210 VDC. The inspector observed that testing of the 250 VDC battery was not terminated until voltage had dropped to 204 VDC. As a result of this error, the complete discharce test was reperforme This caused unnecessary wear on the battery and could have been avoided if the procedure required more frequent checks of battery voltage at the end of the test. At the Exit Meeting, the licensee acknowledged the problem and indicated that the procedure would be reviewe The inspector did note that during test performance maintenance supervisor oversight was apparent. Such participation should help to identify and correct potential problems in the futur Observation of HPCI suppression chamber level instrumentation testin _
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On July 2,1986, the performance of the high pressure coolant injection (HPCI) suppression chamber level instrument functional and calibration test, procedure 8.M.2-2.5.78, Revision 7, was witnessed. During the surveillance, level switch, LS-2351A, that is used to transfer the HPCI suction to the suppression pool using the suppression pool level, was found to be stickin A maintenance request (MR) was written to replace this switc The inspector had no further question Review of HPCI Surveillance Test Procedures The inspector performed a detailed review of HPCI surveillance test procedures to determine if technical specification requirements had been addressed. The inspector identified three areas of concer (1) PNPS Technical Specification Table 4.2.B, Minimum Test and Calibration Frequency for core standby cooling systems, requires performance of logic system functional tests once per six month The definitien of logic system functional tests, as given by the technical specifications, is:
"A logic system functional test means a test of all relays and contacts of a logic circuit from sensor to activated device to insure components are operable per design intent. Where prac-ticable, action will go to completion; i.e., pumps will be started and valves opened."
The inspector identified that the following system logic features <
are not fully tested in that testing is not taken to completion; #
i.e., valves are not stroke t ,
a) Automatic transfer of HPCI pump suction on condensate storage tank low leve b) HPCI system automatic isolation on steam line high differ-ential pressure, turbine exhaust cumpartment, torus cavity or HPCI/RHR valve station exhaust duct high temperatur c) HPCI system valve opening blocks during system isolatio This feature maintains valves closed with simultaneous initiation and isolation signals presen d) HPCI initiation signals for the steam supply line isolation valves, steam admission valve and injection valve are tested only once per operating cycl The operability of these functions is required by technical specification Proper functioning of these features ensures that on system initiation or isolation all components respond as designed. The inspector discussed these findings with the
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licensee. The licensee indicated that similar problems had recently been identified by a consultant review of surveillance testin The licensee had not informed the inspector of any actions to be taken as a result of these findings or those of the consultant prior to close of the inspection period. This item is considered unresolved pending further evaluation by the inspector. (86-21-03)
(2) The inspector noted that current testing philosophy draws virtually no distinction between logic system functional testing and simulated automatic actuation testin Testing conducted to satisfy the simulated automatic actuation requirement of technical specification 3.5.C is performed by a series of overlapping tests, the same tests used for logic system functional tests. The inspector questioned the licensee regard-ing the intent of the requi.ement. Specifically, was simultane-ous integrated system initiation, with the exception of injec-tion, needed to meet the intent of the simulated automatic actu-ation test? The licensee is reviewing the issue. The inspectors will be discussing this requirement with NRR. P.is item is unresolved, pending the completion of the licensee and NRR revie (86-21-04).
(3) The inspector reviewed PNPS Procedure 8.5.4.3, Revirion 11, HPCI Flow Rate lest at.150 psig. This test is performed as required by technical specification 4. The test as written starts, accelerates and achieves the needed flow / pressure values using t the test potentiometer to control HPCI turbine speed. However, normal and automatic system starts utilize the flow controller i
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to control system operation. By conducting the test via the test potentiometer the normal flow control system is not proven operabl If the intent of the test is to demonstrate system operability during startup, prior to continuing pressurization, then start and operation of HPCI using the potentiometer does not satisfy this objective. Deficiency Report 1322 identified a similar problem with RCIC testing in 1985. In response to this DR the RCIC test procedure was changed to transfer to the flow controller after start, but prior to taking flow / pressure dat No changes were made to the HPCI procedure at that tim The inspector discussed this matter with licensee technical representatives. This item is unresolved and will be reviewed in more detail during a future inspectio (86-21-05).
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4.0 Review of Plant Events Operators, Maintenance and Clerical Worker Strike On June 12, 1986 members of the three striking Boston Edison unions accepted a proposed contract offer and began returning to work after a four-week strike. Management personnel continued to fill vacant union positions until all personnel had returned and resumed their duties. The transition appeared to progress effectively, and no problems were note The inspectors routinely observed the performance of management per-sonnel during the strike. Observation of operations, maintenance and health physics activities indicated that work / operations were effec-tively coordinated, planned and conducted. Minor problems with MSIV reassembly work and health physics procedure adherence were observed and are described in paragraphs 3 and 6 of this report. The inspec-tors also reviewed training provided to licensed operators prior to their return to shift. This training was adequate with the exceptions noted under the unrer,1ved item (86-07-01) update in paragraph 2 of this report concernicg the unqualified diesel generator lockout rela Overall licensee response to the prolonged labor action was well organized and thorough. The careful approach taken in scheduling and conducting maintenance ard operations activities resulted in quality result Failure of 125 VDC Distribution Panel "C" Auto Transfer Switch i e
On June 5,1986 power to 125 VDC distribution panel "C" (06) was i momentarily lost when supply to panel 06 spuriously transferred from !
the normal to alternate source. Attempts to transfer the distribu-tion panel back to its normal supply were unsuccessful. Both normal and alternate feeds pass through auto transfer switch Y10. The func-tion of switch Y10 is to sense loss of normal supply voltage and transfer the feed to the alternate. Licensee investigation revealed that the Y10 undervoltage relay had failed, causing the spurious auto transfer. The failed relay was sent to the vendor for examination and repair. The licensee had not received results of the inspection or the rebuilt relay at the close of this report period. This item will be followed during future routine inspection "A" Recirculation Motor Generator Set Field Breaker Failure On June 29, 1986, while removing the "A" recirculation motor generator set from service, the generator field breaker did not trip as expecte One function of this breaker is to trip, halting recirculation flow, during an anticipated transient without scram event. Licensee inves-tigation revealed that the breaker was mechanically bound and could not be actuated by rotating the trip shaf The shunt trip device l
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had energized, rotating the trip shaft slightly, but the trip action was not completed. Because the breaker had not fully tripped, the breaker auxiliary switch "a" contact, in series with the shunt trip coil, could not interrupt the trip coil current. As a result the shunt trip coil burned up. At the close of this inspection period the licensee had not completed investigation of the incident and no cause had been identifie On February 21, 1980 the NRC issued an Order requiring installation of an ATWS recirculation pump trip function. In response, the li-censee installed a second shunt trip coil on the two field breakers, and the required instrumentation / logic to cause a breaker trip when certain parameters are sensed. Because of the breaker mechanical binding experienced, this ATWS trip function would not have actuated if called upo The licensee informed the inspector that three of the GE type AK-F-2-25 breakers are used as the A & B recirculation MG set field breakers and the main turbine generator field breaker at Pilgri The inspector noted that similar failures had occurred in the B MG set field breaker on April 2, 1983; the A MG set field breaker on February 9, 1985; and the main turbine generator field breaker on March 15, 198 .
The inspector expressed concern regarding the high failure rate of this breaker type. While the ATWS trip function is not considered safety related, as described in 10 CFR 50.62.d, its operability is required by technical specifications. The continuing tendency of m these components to bind calls into question their ability to perform h if require Licensee corrective action in response to the failure h will be evaluated. (86-21-06).
d. Residual Heat Removal Loop B Flow Anomolies i On June 28,1986, at 0858 hours0.00993 days <br />0.238 hours <br />0.00142 weeks <br />3.26469e-4 months <br />, control room flow indication was lost while operating D RHR pump in shutdown cooling mode. RHR pumps l
B and D share a common discharge to the reactor vessel. A flow ele-t ment and transmitter in the common discharge provide loop B flow
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signals to flow indicator FI-1040-1, and flow recorder FR-1040-7 in the control room. Both instruments dropped downscale when the problem occurred. The operator noted that pump motor amps appeared to remain
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normal. Preliminary I&C testing identified no problems with the in-struments. The D pump was subsequently returned to service with normal indication.
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At 2145 hours0.0248 days <br />0.596 hours <br />0.00355 weeks <br />8.161725e-4 months <br />, with the D pump in service, flow indication again i fell downscal The D pump was secured and the B pump started.
l With the B pump running flow appeared normal. The B was removed I
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from service and D pump restarted. Flow indication was erratic and it was noted that " Auto Blowdown Permissive RHR/ Core Spray Pump
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Running" alarm was not received as expected. This alarm should be energized when pump discharge pressure exceeds approximately 150 psig, as sensed by a pressure switch just downstream of the pump discharge. The D pump was removed from servic Following the incident on June 29, 1986, RHR pumps B and D were again started for conduct of procedure 8.5.2.1, LPCI Subsystem Operability Surveillance Test. The licensee stated that test results indicate acceptable pump performance. Extensive testing of all involved in-strumentation identified no problems. The licensee is currently evaluating available information on the events. The subject pumps are scheduled for disassembly and inspection during the current out-age as discussed in this report under open item 85-31-02. The in-spectors will review the results of the licensee's ongoing investi-gation (86-21-07).
e. Strike by Contract Security Force On June 30, 1986, at approximately 8:00 p.m., the licensee notified Region I that the United Plant Guard Workers Unicn of America Local 540, representing the contract security force (Globe Security System),
had voted to strike. The strike vote followed the union's failure to accept a new labor contract. The potential for a strike had been anticipated and a contingency force, consisting of licensee and Globe security supervisory personnel from other sites, was called in. The contingency force had previously been trained in accordance with the Pilgrim Security Training and Qualification Pla The licensee stated that fire brigades were manned with security personnel who had received t appropriate fire training and who were familiar with the plant layou ]
fl The regular security force members were escorted off site and all vital areas were inspecied by the licensee. No problems were en-countered. Orderly picket lines were established at all three plant gate Boston Edison union personnel, including reactor operators, subsequently honored the picket lines. ~ Boston Edison managemen personnel were used to fill in for missing union workers as was done during the recently concluded strike by licensee employees described in paragraph 4.a of this repor In a second ballot held July 1, at approximately 2:00 p.m., the union chose to accept the proposed labor contract, strike activities ceased, and the normal security forced resumed their duties. Both the licen-see and resident inspectors were in communication with regional security specialists during the labor actio Review of the licen-see's preparation for the strike and performance during the strike identified no problems.
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19 Main Stack Sample Canister Particulate Filter Missing On July 1,1986 during performance of routine weekly surveillance, the licensee discovered that no particulate filter had been loaded into the main stack sample canister. A continuous representative sample of main stack effluent is passed through the canister so these station releases can be quantified. The canister is removed weekly for analysis and then reloaded and replaced. Apparently after per-formance of the previous surveillance, the required particulate filter was not installed. As a result the weekly particulate sample required by Technical Specification Table 4.8.3, Radioactive Gaseous Waste Sampling and Analysis Program, could not be met. The plant is presently in cold shutdown with release rates a small fraction of allowable limits. The licensee is evaluating other available data to estimate the results of the missed weekly sample. The licensee has stated that the cause of the incident was an isolated personnel error
by an experienced technician and that no similar incidents have oc-curred. The inspector had no further question .0 Observations of Physical Security Checks were made to determine whether security conditions met regulatory requirements, the physical security plan, a'ad approved procedures. Those checks included security staffing, protected and vi al areas barriers,
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personnel identification, access control, badging, ind compensatory mea-sures when required. The inspectors observed security performance during the security union labor action as discussed in paragraph 4.e of this report. No problems were identifie +
6.0 Radiation Protection Radiological controls were observed on a routine basis during the report-ing period. Standard industry radiolog! cal work practices, conformance to radiological control procedures and 10 CFR Part 20 requirements were ob-serve Independent surveys of radiological boundaries and random surveys of nonradiological points throughout the facility were taken by the inspec-tor. The following problems were note During a tour on June 4, 1986 of the reactor building 51 foot elevation the inspector observed I&C personnel inside the reactor vessel instrument cages using a plant page. The instrument cages were posted as a con-taminated area. The page phone cord had been stretched across the con-taminated area boundary. After test completion, the page and several tools were removed from the area without a health physics survey or other precautions. This same problem was observed by inspectors on April 15 and April 30, 1986, and is described in inspection report 86-0 : :
On June 10, 1986, the inspector witnessed control rod drive pump disas-sembly activities. The area immediately surrounding the pump / motor was designated as a contaminated area, and full protective clothing was re-quired for entry. The inspector observed that poor contamination control practices were employed at the jobsite. Specifically, tools, various pump internal parts and the pump casing top half had been removed from the posted area and stored on top of plastic bags in a non-contaminated area nearby. In addition, tools and rigging were passed over the contaminated area boundary without being surveyed. The inspector contacted health physics and a survey of the materials and area was conducted. Contamina-tion levels measured were low. The area which had been established around the pump was not large enough to allow storage of tools and material Proper coordination between health physics and maintenance should have identified this during planning or corrected it prior to job star Both incidents described above occurred during the union employee strike and involved only maintenance supervisors. These incidents of poor prac-tices, though minor, demonstrate an underlying attitude of disrespect for adherence to health physics controls. The inspector discussed this con-cern with health fiysics department management on June 16. The Radiation Protection Managt n acknowledged the existence of the problem and said that steps to correct it would be taken. On June e7, the inspector met with the Plant Manager to discuss the issues. Proposed corrective actions in-cluded personnel couiseling for individuals involved, more aggressive follow up of identified problems and joint workshops with health physics and maintenance personne The inspectors will closely monitor health physics practices to verify j effectiveness of these measure On June 24, 1986, a Resident Inspector, using a NRC digimaster by !
Xetex, Inc., monitored a radiation field of 6 mr/hr outside and ad-jacent to the fenced radioactive material storage area around the nitrogen storage tank near a crate labeled as containing highly radio-active material. No "Do Not Loiter Sign" indicating the radiation field strength was posted, although these signs were posted along other fenced radioactive material storage areas with radiation fields as low as 2 MR/hr. It was found that the licensee does not have a procedure to control the posting nor the updating of these warning, i.e., "Do Not Loiter" sign On June 25, 1986, two TLDs that were each doubled wrapped in plastic bags and hung inside the protective area fence were found filled with water from condensation by a Resident Inspector. The licensee was asked, if the water would significantly affect the TLD readings; and the purpose for these TLDs being positioned. In response to these questions, a Health Physics (HP) Supervisor stated that the water would affect the TLD reading, but these TLDs were routinely inspected to ensure that the TLDs were kept dry. However, five days later, on July 1, 1986, these same two TLDs were checked and found to be in the same conditio Both the control of the instrumentation used and
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maintenance of the data obtained were not controlled procedurall At the exit meeting, the inspector questioned the implications of the problem for the environmental TLD program. The licensee stated that the TLDs questioned were used as part of a study to determine the validity of environmental TLD data cited in IE Inspection Report 50-293/86-09. The TLDs in question were not part of the environ-mental programs which do not have a similar TLD water problem. The long term solution for the TLDs questioned will be to place these TLDs into plastic containers used for environmental TLD The two examples discussed above represents a poor practice of con-trolling activities in a nuclear power plant without procedure On June 19, 1986, a conference call was held between the acting Pilgrim Plant Manager, Mr. J. Seery; two NRC Region I managers Mr. J. Strosnider and Db. W. Pasciak; and the Senior Resident Inspec-tor. The status of the environmental Thermoluminescent Dosimeter (TLD) program was discussed in light of the findings of NRC Inspec-tion 50-293/86-09. The licensee indicated that several actions either had been taken or were planned to address the NRC concern One of the actions was the deployment of duplicate TLD's. At the end of the inspection, the licensee's Radiological Section Manager indi-cated that environmental TLD's and occupational TLD's (fron Yankee Atomic) were deployed at selected environmental monitoring stations in addition to the routine Pilgrim TLD's. The Manager also indicated that the duplicate badges would remain in place until the deployment of the new Panasonic TLD syste The new system is expected to be operational by January 1, 198 +
7.0 Master Surveillan u Tracking Program Review L On June 13, 1986, the licensee notified the NRC via the ENS telephone line that two types of surveillance tests required by the technical specifica-tions had been improperly scheduled and may t;e overdue. These tests are required to be conducted once per cycle by the technical specification The tests included local leak rate tests (LLRT) of primary containment penetrations and isolation valves and a manual initiation test of the standby liquid control syste In response t'o the LLRT finding, the licensee started a full LLRT test program. This program will be completed prior startup from the current outage. An integrated containment leakage test is also planned for the current outage.
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Subsequently, on June 21, 1986, the licensee notified the NRC that a test
! involving the calibration of undervoltage relays was overdue. This test, 3.M.3-1, calibrates the relays that sense loss of power to the startup transformer and the two 4160 VAC safety busses and starts the emergency diesel generators. The calibrations were also required to be conducted once per operating cycle and had a similar scheduling problem.
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The technical specifications has two definitions for operating cycle:
Definition 0 (page 4): Interval between the end of one refueling
. cycle and the end of the subsequent refueling cycl Definition U-(page Sa): The operating cycle is considered to be 18 months ....
The differences in wording between these two definitions may have contri-buted to the~ scheduling problem. .In addition to the technical specifica-tion requirements,10 CFR 50 Appendix J also specifies a rainimum LLRT frequency, once per two years, that may not have been adequately factored into the licensee scheduling syste The licensee stated that the Master Surveillance Tracking Program (MSTP) 1 had been modified to change the due dates for the once per cycle surveil- '
lance tests during 1985. The MSTP is a computerized tracking system that schedules surveillance tests at Pilgrim. Originally, these tests were-scheduled on the MSTP to be done every 18 months. However, in 1985, the tests were rescheduled for the next refueling outage. As a result of the extended refueling outage in 1984 and the extended operating cycle'in 1985 and 1986, some of these. tests exceeded the 18 month surveillance interva The licensee had not completed their evaluation of this problem at the end
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of the inspection. This item is unresolved, pending further licensee and NRC review (86-21-08).
8.0 Local Leak Rate Testing Progress /0bservations
! The sum of the "as-found" Type "C" local leakage rate tests (LLRTs) mini- '
l mum leakage path values for the LLRTs conducted to date have exceeded the l
- 1.0% per day maximum allowable primary containment leakage of 210.21 slm .
as stated in the bases of the Technical Specification In obtaining these LLRT results, the licensee has quantified leak rates that were previously l listed as "in excess of 20 scfm" by using flow meters with greater ranges
! and utilizing pressure decay techniques. A summary of the the testing
!- completed and specific failures are provided in Attachment On June 24, 1986, during observation of a type "B" test on main steam line l drain bellows, the inspector noted that a test connection isolation valve l packing leak was corrected prior to taking the "as found" leakage. Depend-ing on the test valve type this packing may communicate with the test i volume. Further investigation indicated that the particular valve in l- question was a globe valve, so that inclusion of the packing leak in the
"as found" data was not necessary. The inspector stressed the need for
- careful evaluation prior to correction of such leakage. The licensee i stated that the lead engineer will be consulted prior to correction of
- such leakage. The inspector verified that LLRT personnel were subse-quently made aware of this requirement. Also, the inspector expressed
- concern that the LLRT procedure, 8.7.1.5, Revision 25, did not contain a l caution statement about this potential problem. At the exit meeting, the licensee stated that the procedure would be modified.
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During the inspection period, the inspectors witnessed a number of type
"B" and type "C" LLRTs. The testing and subsequent maintenance of the-feedwater check valves were closely observed. The inspectors will con-tinue to closely monitor containment leak rate testing activitie .0 Survey of Licensee Response to Selected Safety Issues (TI 2515/77)
The inspector completed inspection of Technical Instruction 2515/77, Survey of Licensee Response to Selected Safety Issues. The objective of the inspection was to determine the extent of action taken by licensees regarding industry identified problems. Included were responses to IE Bulletins, Circulars, and Information Notices and Institute of Nuclear Power Operations (INPO) Significant Operating Event Reports (SOER). The inspector discussed the intent of the inspection with licensee management, especially with respect to INPO SOERs, prior to start. Issues reviewed concerned the reliability of HPCI/RCIC systems, and the potential for befouling of safety related heat exchanger Specific recommendations for increased reliability of HPCI/RCIC systems have generally been implemented. More programmat'c recommendations such as assignment of a qualified engineer to monitor sy ;em performance, mainten-ance, and vendor / industry operating experience have not been adopted. The licensee has implemented an extensive program to prevent and monitor bio-fouling of both safety related and nonsafety relatec heat exchangers. This program is ongoing and appears to be very effectiv During the review the inspector noted that administration of the program for feedback of operating experience appears to be dependent on the efforts +
of a single individua Documentation of the evaluations performed, de-cisions made and actions taken was not available in .nany cases. The in- f o
spector could not draw a conclusion as to the effectiveness of the program ,
based on this small sampl This item will be reviewed further during followup on a previous item, (84-01-01).
10.0 Review of LER's LER's submitted to NRC:RI were reviewed to verify that the details were clearly reported, including accuracy of the description of cause and ade-quacy of corrective action. The inspector determined whether further information was required from the licensee, whether generic implications were indicated, and whether the event warranted onsite followup. The following LER was reviewed:
LER N Event Date Report Date Subject
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86-012 5/16/86 6/16/86 Minimum flow valve for HPCI once/ cycle auto actuation test found inadequate
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The event described in LER 86-012 was reviewed in inspection report 86-1 The inspector had no further questions concerning this LE .0 Public Meetings and Meetings with Local Organizations On June 10, 1986, the NRC Region I Administrator, Dr. T. Murley, and the Director of the Division of Reactor Projects, Mr. R. Starostecki partici-
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pated in a public meeting in Plymouth, Massachusetts. The meeting was held at the request of the Plymouth Board of Selectmen and focused on re-cent probleus at Pilgrim Station. Two Resident Inspectors, Dr. M. McBride and Mr. J. Lvash, attended the meeting. Licensee representatives also participated in the meeting. The meeting lasted about four hour On June 25, 1986, Mr. Starostecki and Dr. McBride participated in a public meeting in Juxbury, Massachusetts. This meeting was held at the request of the Duxbury Board of Selectmen and lasted about four hours. Mr. Lyash attended the meeting. Representatives of the Commonwealth of Massachusetts, FEMA, and Boston Edison also participated in the meetin On June 30, 1986 the Chief of the Emergency Preparedness and Radiological Protection Branca in NRC Region I, Dr. R. Bellamy, and Dr. M. McBride par-ticipated in a second public meeting in Plymouth. This meeting was held at the request of th3 Plymouth Board of Selectmen to discuss Emergency Planning issues. boresentatives from the Commonwealth of Massachusetts, FEMA, and Boston Ed son also participated in the meetin On June 24, 1986, Dr. M. McBride addressed a special meeting of the Board of Directors of the Plymouth Area Chamber of Commerce. The meeting was !
called to discuss recent problems at Pilgrim. Representatives of the #
licensee and Plymou;h County Nuclear Information, Incorporated, (a local (
public interest group) also participated in the meetin .0 Management Meetings On June 12, 1986, the Senior Resident Inspector attended a meeting between the licensee and NRC Region I management that was held at the Region I offices in King of Prussia, PA. The purpose of the meeting was to discuss the status of CAL 86-10, licensee corrective action programs, plant staff-ing, and QA program implementation and support. Licensee plans to upgrade management practices at the plant and upcoming RHR pump inspections were also discussed. The results of the meeting were documented in NRC meeting report 50-293/86-2 At periodic intervals during the course of the inspection period, meetings were held with senior facility management to discuss the inspection scope and preliminary findings of the resident inspectors. No written material was given to the licensee that was not previously available to the publi . . _ .
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s ATTACHMENT I TO INSPECTION REPORT 50-293/86-21 Persons Contacted L. Oxsen, Vice President, Nuclear Operations A. Pederson, Nuclear Operations Manager P. Mastrangelo, Chief Operating Engineer
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D. Swanson, Nuclear Engineering Department Manager K. Roberts, Director Outage Management '
N. Brosee, Maintenance Section Head T. Sowdon, Radiological Section Head J. Seery, Technical Section Head E. Ziemianski, Management Services Section Head S. Wollman, On-Site Safety and Performance Group Leader B. Eldridge, Acting Chief Radiological Engineer R. Sherry, Chief Maintenance Engineer J. McEachern, Resource Protection and Control Group Leader
. E. Graham, Compliance and Administrative Group Leader E. Gordon, Environmental and Radiological Health and Safety Group Leader (Acting)
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s IR PILGRIM 86-21 - 0052. /29/80
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ATTACHMENT 2 TO INSPECTI0h' REPORT 50-293/86-21 SurMRv 0F LlRT STATUS The Filgris Nuclear Fceer Station local leak rate test (LLRil results 3rci Apr:i !?, 1986, tc Jul. 7. St.,'
- are smarned below:
. For Type "B" LLRis, 62 of a; proximately 101 cozponents have teen teste: with ! failure . For Type 'C' LLRis, 43 of approximately 133 ccaponents have been tested eiith 11 f ailure SUMARY OF SPECIFIC LLRT FAILURES Systes Date LLRT Penetration 'As-Found' Leakage (slal Date Previously Exceeded Cocconent Tested Tyg Nueber Inboard Outboard Minisue-Path Tested 18 so interval MSIV 'A' 04-19-86 C 7A 70.31 - -
11-16-84 No 04-19-86 C 7A -
8.69 8.69 11-17-84 No J
MSiv 'E' 04-20-86 C 78 36.66 - -
11-le-84 No 04-20-86 C 76 -
4.40 4.40 11-16-84 No MSIV D' 04-20-86 C 7D 52.93 - -
11-16-84 No 04-20-86 C 7D -
13.43 13.43 11-16-64 No FW cl 'A' 06-26-86 C 9A 177.58 - -
01-02-84 Yes 07-01-86 C 9A -
648.50 177.88 09-05-64 Yes
- - 09-07-84 FW ct 'C' 06-27-86 C 9B 73.63 -
Yes 06-30-86 C 9B -
16.50 16.50 09-14-84 Yes MS drain 06-24-86 C 8 1.00 - -
12-03-84 No 06-24-86 C B
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15.75 1.00 12-03-86 No RWCU supply 07-07-86 C 14 4.00 - -
11-23-84 No 07-07 86 C 14 -
23.00 4.00 11-23-84 No D/W Access 06-10-86 B 2 0.00 - -
12-16-8 5 No 06-10-86 B 2 -
10.50 0.00 12-16-8 5 No Total Minisus Fath Leakace for Failed Valve Found ' : 225.90
!. Acceptance criterion at 45 psig for:
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Type B' LLRT except for access airlock, is 15.78 sin;
- Type 'B' LLRT,for access airlock is 7.89 sin;
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Type 'C' LLR1 except for MSIV, is 7.89 sle; and
. Type 'C' LLRT,for MSIV is 8.5C sit.
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