IR 05000293/1987057: Difference between revisions

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{{Adams
{{Adams
| number = ML20153A858
| number = ML20195B175
| issue date = 03/11/1988
| issue date = 03/11/1988
| title = Insp Rept 50-293/87-57 on 871207-880119.Violation Noted. Major Areas Inspected:Plant Operations,Radiation Protection, Physical Security,Plant Events,Maint & Surveillance.One Unresolved Item Identified
| title = Forwards Safety Insp Rept 50-293/87-57 on 871207-880119 & Notice of Violation.Equipment Failures Indicate Need for Stronger post-work Test Practices & Thorough Power Ascension Test Program
| author name = Blough A
| author name = Collins S
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
| addressee name =  
| addressee name = Bird R
| addressee affiliation =  
| addressee affiliation = BOSTON EDISON CO.
| docket = 05000293
| docket = 05000293
| license number =  
| license number =  
| contact person =  
| contact person =  
| case reference number = TASK-2.E.4.2, TASK-TM
| document report number = NUDOCS 8803210184
| document report number = 50-293-87-57, IEB-79-08, IEB-79-8, IEIN-87-030, IEIN-87-30, NUDOCS 8803210189
| package number = ML20153A853
| package number = ML20153A853
| document type = INSPECTION REPORT, NRC-GENERATED, INSPECTION REPORT, UTILITY, TEXT-INSPECTION & AUDIT & I&E CIRCULARS
| document type = CORRESPONDENCE-LETTERS, NRC TO UTILITY, OUTGOING CORRESPONDENCE
| page count = 30
| page count = 3
}}
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MAR 1l11988
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Docket No. 50-293 Boston. Edison Company ATTN: Mr. Ralph G. Bird Senior Vice President - Nuclear 800 Boylston Street Boston, Massachusetts 02199 Gentlemen:
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Sub.iect: Region I Inspection Report No. 50-293/87-57 This refers to the routine safety inspection (50-293/87-57) conducted by
V. S. NUCLEAR REGULATORY COMMISSION-
'  Messrs. C. Warren, J. Lye.sh and T. Kim of this office on December 7,1987 to January 19, 1988:at the Pilgrim Nuclear Power Station, Plymouth, Massachusett Areas examined during this inspection aredescribed in -the NRC Region I Inspection Report. which is enclosed with this lette Based on the resuits of this inspection, it appears that one of your activities related to high radiation area access control was not conducted in full com-pliance with NRC requirements, as set forth in the Notice of Violation enclosed herewith as Appendix A. - The problem was identified by your staff. However, a Notice of Violation is being issued because effective corrective actions apparently liave not been taken for previous problems with high radiation area-access control. In addition to following the irstructions of Appendix A in'
 
preparing the required response, please include chose actions you intend to take to preclude recurrence of this problem by insuring that your corrective actions are effective and' lastin Two significant integrated plant tests were successfully executed during-the inspection period. Preplanning and control of these activities was generally stron We also observed that increased management involvement in assuring effective problem followup has resulted in substantial improvement. Equipment failures identified as a result of an unanticipated safety system actuation however, indicate the.need for stronger post-work test practices and a thorough power ascension test progra The response directed by this letter and the accompanying Notice are not subject to the clearance procedures of the Office of Management and Budget as required by the Paperwork Reduction Act of 1980, PL 96-511.
==REGION I==
  . Docket / Report N /87-57 Licensee:
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_ Boston Edison Company 800 Boylston Street Boston, Massachusetts 02199 Facility:  Pilgrim Nuclear Power Station Location:  Plymouth, Massachusetts Dates:  December 7, 1987 - January 19, 1988 Inspectors: C. Warren, Senior Resident Inspector J. Lyash, Resident Inspector T. Kim, Resident Inspector Approved By:         1/ h cI8 A. Randy Bt6 ugh, Chie,f      Date Reactor Projects Section No. 3B Areas : Inspected:  Routine res. dent inspection of plant operations, radiation protection, physical security, plant events, maintenance, surveillance, outage activities, and reports to the NRC. The inspection consisted of 350 hours of direct inspection. Principal licensee management representatives contacted are listed in Attachment Observations made by the NRC Region I, Regional Administrator during a tour on December 8,1987 are documented in Attachment II of this report. A copy of Attachment II was provided to licensee management for followu 'Results:
Violation: Repeated occurrences of locked high radiation area doors being left open and unattended were identified by the licensee. Problems with high radia-tion-area access control have been previously identified and were the subject of . violations during inspections 50-293/87-03 and 50-293/87-11'.      Corrective actions taken in response to these findings have not prevented their recurrence. (Section 3.b, VIO 87-57-01)
Unresolved Item:  The licensee identified that two reactor vessel level gauges      .
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were incorrectly installed. A licensee investigation is currently ongoing to determine the cause and to assess the adequacy of post installation test.


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L OFFICIAL RECORD COPY IR PILGRIM 87-57 - 0001. ,  03/11/88 8803210184 880311     C A PDR ADOCK 05000293     .(/
  (Section 4.d, UNR 87-57-02)
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8803210189 880311 PDR ADOCK 05000293 Q   DCD
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Boston Edison Compan MAR 111988
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Your cooperation with us in this matter is appreciate
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Inspection Results-(Continued)
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Concerns:
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  - The ~ licensee. experienced safety rele+.ed equipment malfunctions upon receiving ~ a spurious reactor- scram ' signal 'on January 17, 1988. -(Section 4.d) Inadequa'te procedures and planning of - surveillance tests resulted in un-necessary engineered safety feature actuations. (Section 3.a) Poor preplanning and control of maintenance was noted during an-electrical relay replacement. A similar problem was the subject of a violation dur-ing inspection 50-293/87-50. (Section 4.c) Weak identification and tracking of lifted leads and jumpers led to a water spill in the.high pressure coolant injection system room 'during the integrated leak rate tes (Section 6.0) The prelube' pump for the "B" emergency diesel generator failed to. restart during a surveillance test. An identical failure occurred during a loss of offsite power event on November '12,1987.. Licensee followup appeared
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adequate but the failure root cause has not been identifie (Section 3.b) The inspectors evaluated the erosion of construction dirt into wetlands area. The inspector's independent survey of the area, and the licensee's analyses indicate that the level of activity does not represent a health or safety concern. However, the material should not be allowed to erod (Section 3.c)
Strengths: The licensee's preparation and execution of the reactor vessel hydrostatic test was well organized and controlle (Section 5.0) .The licensee's response to a January 17, 1988 reactor scram signal and subsequent equipment malfunctions was prompt, thorough and effectiv (Section 4.d) Using non-nuclear steam for testing of high pressure coola1t injection system and reactor core isolation cooling system enabled the licensee to discover problems which may not have been easily identifiable using nuclear steam due to radiological condition (Section 3.b)
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Sincerely,
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dgginal Stcncd I?2
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TABLE OF CONTENTS Page . Summary of Facility' Activities ........................ 1 Followup on Previous Inspection Findings .............. 1 Routine Periodic Inspections .......................... 4 I Surveillance. Testing ' Radiation Protection and Chemistry      _ Fire Protection
          { Review of Plant Events ................................ 15 . Spurious Isolations of RHR Shutdown Cooling System Reactor Water. Cleanup System-Spurious' Isolation Engineered Safety Feature Actuations Due to a Failed Logic Relay  .
  'd, Spurious Reactor Protection System Actuation Review of Reactor Vessel Hydrostatic Test Procedure
  : a nd - Te s t ' Re s u l t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 Integrated Leak-Rate Testing .......................... 21 Licensee Nuclear Organization Managenant Realignment ......................................... 23 Management Meetings ................................... 24 Attachment I - Persons Contacted Attachment AI - Regional Ad:ninistrator's Tour Observations l
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DETAILS 1.0 Summary of Facility Activities The plant was shutdown on April 12, 1986 for unscheduled maintenance. On July 25,1986, Boston Edison announced that the outage would be extended to include refueling and ' completion of certain modifications. The reactor core was .defueled on February 13, 1987. The licensee completed fuel re-load on October 14, 1987. Reinstallation of the reactor vessel internal components and the.vessei head was also subsequently complete During this report period, the licensee performed the reactor' vessel hydrostatic test and the primary containme- . integrated leak rate -test (ILRT) as described in Sections 5.0 and 6.0. On December 9, 1987, Pilgrim Station conducted a partial participation emergency preparedness exercis On December 14, 1987 the licensee announced as part of a planned manage-ment realignment, the appointment of eight managers to key management positions in the licensee nuclear organization at Pilgrim Statio The details of the management realignment are described in Section NRC inspection activities during the report period included: 1) observa-tion of the licensee's annual emergency preparedness exercise on December 9, 1987, 2) NRC Reactor Operator Licensing examinations were administered to eight candidates on the week of December 7, 1987, 3) ob-servation of the primary containment ILRT and review of the test results during the week of December 21, 198 The results of these inspections are documented in inspection reports 50-293/87-54, 50-293/87-56, and 50-293/87-58. In addition, representatives of the NRC's Office of Inves-tigation were onsite December 3, December 7, and December 8, 1987 to interview onsite security personne On December 8, 1987, the NRC Regional Administrator for Region I, Mr. William T. Russell, toured the plant with the resident inspector On January 7, 198,8, Dr. Thomas E. Murley, Director of the Office of Nuclear Reactor Regulation (NRR) and other NRC representatives toured the plant with the resident inspector .0 Followup on Previous Inspection Findinjs (Closed) Unresolved Item 82-E4-02 - Discrepancies in the Licensee's Response to IE Bulletin 79-08 Previous reviews of this item are documented in the inspection reports 50-293/82-30, 50-293/83-01, 50-293/83-14, and 50-293/84-2 IE Bulletin (IEB) 79-08 ano the TMI Action Plan Item II.E.4.2 required licensees to review the containment isolation initiation cesign and procedures to ensure proper initiation of containment isolation, upon receipt of an
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automatic containment isolation signa The licensee provided the results of their review in letters dated April 25, and August 21, 197 I
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Samuel J.' Collins, Deputy Director Division of Reactor Projects Enclosures: Appendix A, Notice of Violation NRC Region I Inspection Report No. 50-293/87-57-cc w/encis:
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R. Barrett, riuclear Operations Manager B.-McIntyre, Chairman, Department of Public Utilities Chairman, Plymouth Board of Selectmen Chairman, Duxbury Board of Selectmen Plymouth Civil Defense Director J. Keyes, Boston Edison Regulatory Affairs and Programs E. Robinson, Nuclear Information Manager R. Swanson, Nuclear Engineering Department Manager    >
 
, The Honorable Edward J. Markey      *
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The licensee stated that the RBCCW supply and return lines, instrument air line, RHR to spent fuel pool cooling tie line, and torus make ' up line would be manually isolated and that station procedures would specify the requirements for manual isolation if a containment isolation signal was received. Thit was documented as acceptable by NRC:NRR in letters to the licensee cated December 18, 1979 and April 3, 1980. However, an inspector identified that manual isolation of these lines with qualified valves is not possible. Any valve which is used for primary containment isolation must meet Seismic Class I (FSAR section 12.2) and applicable 10 CFR 50, Appendix J, containment leakage testing criteria. Further, if manual operation of a valve is required to effect containment isolation, the isolation point for the valve must also be accessible under those condi-tions which make its use necessar In response to the inspector's questions, the licensee re-evaluated their response to the IEB 79-08 and TMI Action Plan Item II.E.4.2, and concluded that isolation of these lines is assured by the .use of Seismic Class I check valve The licensee also agreed that isolation for the RBCCW supply .line, instrument air line, RHR to spent fuel pool cooling tie line, and torus makeup line cannot be performed by manual valve closur The RBCCW return line from the drywell can meet the isolation valve criteria with MOV-4002 which is seismic class I, local leak rate tested and can be closed by a control switch located in the main control room. The licensee subsequently submitted a supplemental response to IE Bulletin 79-08 and TMI Action Plan Item II.E.4.2 on October 24, 1984 correcting the previous respons The inspector reviewed the supp'emental response and verified that the contents were consistent with the conclusions drawn from the licensee's re-evaluation and the FSAR. Both RBCCW supply line and instru-ment air line are considered Class C lines in Section 7.3 of the FSAR since they penetrate containment but have no interaction with the primary
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containment free space or the reactor vessel. According to the original
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design criteria, a single check valve is provided to attain isolation for a Class C line. These check valves are seismic class I and local leak rate tested. The inspector reviewed the results of local leak rate test data for these check valves which were performed on June 12 and July 26, 1987 and found no discrepancie The torus makeup line is identified as Class B in Section 7.3 of the FSA The torus makeup line is non-essen-tial and ties the condensate transfer system into the RHR test line, which penetrate primary containment and ends below the torus water level. Ft e water-sealed Class B lines such as the torus makeup system, the original plant design bases allow one isolation valve in addition to the water seal to meet isolation requirements. Also, the Safety Evaluation by the NRR on Appendix J Review indicate that Type C testing is not required for valves in lines which terminate below the level of the suppression pool. As for the RHR to spent fuel pool line, the licensee revised the operating pro-cedures 2.2.85, Fuel Pool Cooling and Filtering System, prohibiting the use of the RHR to spent fuel pool lines except in cold shutdow The inspector had no further question This item is close .
The Honorable Edward P. Kirby The Honorable Peter V. Forman S. Pollard, Secretary of Energy Resources P. Agnes, Assistant Secretary of Public Safety, Commonwealth of Massachusetts R. Shimshak, MASSPIRG-Public Document Room (PDR)
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Local Public Document Room (LPDR)
Nuclear Safety Information Center (NSIC)
NRC Resident Inspector Commonwealth of Massachusetts (2)
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Region I Docket Room (with concurrences)
W. Russell, RA
  .M. Perkins, ORMA (w/o encis)
R. Blough, DRP L. Doerflein, ORP R. Bores, ORSS S. Collins, DRP C. Anderson, DRS D. Mcdonald, LPM, NRR
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T. Chandrasekaran, SPLB, NRR M.~ Callahan, OCA J. Nickerson 0FFICIAL RECORD COPY IR PILGRIM 87-57 - 0002. /11/88
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f (Closed) Inspector Follow Item (IFI 87-27-02) - Cracking of Surge Ring Brackets in Large GE Motors v On July ' 2, 1987, IE Information Notice 87-30, Cracking of Surge Ring Brackets in large GE motors, was issued. . The purpose of the notice was to alert recipients of a potential for failure of. surge ring brackets and cracking of felt blocks in l a rge , vertical electric motors' manufactured by General Electric Co. Felt blocks are used in large electric motors to
  . keep the windings separated where they loop back at the end of the . stato The blocks are attached to a surge ring that is held in place by L-shaped surge ring brackets welded to the surge ring and bolted to the motor cas-ir g. . Failure of these surge ring brackets and cracking of the felt blocks allows movement and wear of the end-turns, leading to a reduction in insulation resistance and possible - motor failur In addition, broken pieces of the surge ring bracket may enter the space between . the stator and the rotor, resulting in electrical or mechanical motor degradatio Following an investigation to determine the applicability of the subject notice to the Pilgrim Station, the licensee found that RHR, core spray, and recirculation pump motors were potentially . af fected. RHR and core spray pump' motors were overhauled on site by GE under contract with the licensee in 198 The surge ring brackets were not inspected during the overhau However, small cracks were found on the "A" and "C" RHR pump motor winding felt blocks. The amount of cracking found was dispositioned by GE to be acceptable and a normal phenomenon found in form-wound motor On July 27 through August 5, 1987, GE performed a surge ring bracket inspection of the RHR and recirculation pump motors using a boroscope with the motors in place. The inspection of the RHR motors (A thru D) revealed absence of cracks on the surge ring brackets. During the inspection of the "B" recirculation pump motor, it was noted that the recirc motor surge ring bracket construction is of the bolt and stud design, whereas the RHR and core spray motor brackets are of the L-shaped design. The L-shaped design configuration is known to have the potential of cracking, accoroing to the IE Notice 87-30 and the GE letter to the licensee dated July 14, 198 During the week of October 26,1987, "B" core spray pump motor was dis-assembled and the surge ring brackets inspected by G.E. Due to the geo-metry of the core spray pump motor internals, there is limited access for the bore scope, therefore, this inspection could not be accomplished with-out partial disassembly of the estor. It was verified that the design had 12 brackets per surge ring and two surge rings for the top end turn assem-bly and two surge rings for the bottom end turn assembl None of the brackets had indications of cracking. The licensee scheduled the inspec-tion of tne "A" core spray pump motor during the next outage because of scheduling conflicts. The licensee indicated that based on the inspection
 
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results of the RHR and "B" core spray pump motors, postponement of the "A" core spray pump motor inspection is justifie The licensee also added that the number of operating hours and starts are similar between the A and B core spray pump motors since both core spray systems' testing and surveillance requirements are simila The inspector had no further questions. This item is close (Closed) Unresolved Item 87-45-05 - Failure to Issue Licensee Event Reports In inspection report 50-293/87-45 the NRC identified three engineered safety feature actuations which appeared to be reportable under 10 CFR 50.73 but had not been reported by the license The licensee reviewed the three actuations, agreed that they should have been reported and agreed to issue License Event. Reports (LER) to document the occurrence In addition the licensee agreed to perform a review of previous actua-tions to determine if any additional reports were neede During this inspection period the licensee's compliance section conducted a review of all Failure and Malfunction Reports (F&MR) issued from April 1986 through the present. This review identified four F&MRs that fit the description of an ESF actuation under the current BECo interpretation of NUREG 1022. The licensee will submit LERs to document the following ESF actuations at a later dat /28/87 Initiation signal to both Emergency Diesel Generators (EDG)
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6/7/87 Actuation of Reactor Building Isolation and Standby Gas Treatment System start signal
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9/17/87 Auto start of "A" EDG
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10/6/87 Reactor Water Cleanup and Shutdown Cooling System Isolation These LERs will be reviewed upon issue as part of the normal resident inspection program. The inspector has reviewed the licensee's actions in addressing open item 87-45-05 and is satisfied that those actions were thorough and timely. This item is close .0 Routine Periodic Inspections The inspectors routinely toured the f acility during normal and backshif t hours to assess general plant and equipment conditions, housekeeping, and adherence to fire protection, security and radiological control measure Inspections were conducted between 10:00 p.m. and 6:00 a.m. on January 17, 18, and 19, 1988 for a total of four hours and during the weekends of December 12, 19, 27, 1987 and January 3, 9, 17, 1988 for a total of 17 hours. Ongoing work activities were monitored to verify that thty were
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being conducted in accordance with approved administrative and technical procedures, and that proper communications with the control room staff had been established. The inspector observed valve, instrument and electrical equipment lineups in the field to ensure that they were consistent with system operability requirements and operating procedure During tours of the control room the inspectors verified proper staffing, access control and operator attentivenes Adherence to procedures and limiting conditions for operations was evaluated. The inspectors examined equipment lineup and operability, instrument traces and status of control room annunciators. Various control room logs and other available licensee documentation were reviewe The ir.spector observed and reviewed outage, maintenance and problem inves-tigation activities to verify compifance with regulations, procedures, codes and standard Involvement of QA/QC, safety tag use, personnel qualifications, fire protection precautions, retest requirements, and reportability were assesse The inspector observed tests to verify performance in accordance with approved procedures and LCO's, collection of valid test results, removal and restoration of equipment, and deficiency review and resolutio Radiological controls were observed on a routine basis during the report-ing perio Standard industry radiological work practices, conformance to radiological control procedures and 10 CFR Part 20 requirements were observe Independent surveys of radiological boundaries and random surveys of nonradiological points throughout the facility were taken by the inspecto Checks were made to determine whether security conditions met regulatory requirements, the physical security plan, and approved procedures. Those checks included security staffing, protected and vital area barriers, personnel identification, access control, badging, and compensatory measures when require Surveillance Testing
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Diesel Generator Prelube Pump Failure On December 13, 1987, the prelube pump for the "B" emergency diesel generator (EDG) failed to restart on demand during a routine surveillance tes Upon disassembly it was identified that a small piece tf metal had become lodged between the pump rotor and idler gea The interference from the metal caused the pump motor breaker to trip on pump star An identical failure occurred during a loss of offsite power event on November 12, 198 In that case the failure caused a lengthy delay in returning an  idle diesel to servic While not required for diesei operation, the prelube system reduces EDG bearing wear during equipment star .
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In response to the failures, the licensee drained and. inspected the lube oil sump, and disassembled and inspected the lube oil filters, strainers and heate The lube oil heater _ was found to have failed in the energized mode resulting in significant carbon deposits in the heater and filte No appreciable deposits were found in the. lube oil sump. In addition, a piece of filter element ' packaging material was found in the lube oil filter housing. No foreign material which could have- contrib-uted to' the prelube pump failure, however, was found. The pump was replaced and the diesel was returned to service. No addi-tional failures occurred during the inspection period. The two pumps which failed had in-sequence serial number Licensee Quality Control personnel performed magnetic particle and dye-penetrant testing of the internals of a third in-sequence pump in the warehouse. No flaws were noted. The licensee is pursu-ing the root cause of the failures in cooperation with the pump vendor, Viking Pump. The licensee stated at the exit interview that the "A" EDG prelube pump and lube oil heater would be inspected during the next "A" diesel outage. The inspector will continue to monitor licensee followup to this proble Steam Testing of the High Pressure Coolant Injection and Reactor Core Isolation Cooling Systems The licensee completed full pressure steam testing of the High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) system turbines by utilizing te.nporary oil fired auxiliary boilers as a source of non-nuclear steam. The full pressure steam testing is part of a post-maintenance and system operability chec Both HPCI and RCIC systems were overhauled during the current outage. Utilizing temporary test procedures TP 87-198 and TP 87-199, the HPCI/RCIC testing included turbine j
overspeed trip, pump full flow capacity and operation from the alternate shutdown panels. Also during the test, the suction path was changed from the condensate storage tank to the torus and bac During the testing, several problems were identified by the licensee in both HPCI and RCIC syste:n In HPCI, problems with the governor control system were noted including a minor oil leak in the servo-moto Steam leaks at gauges and turbine drain line were also discovere In RCIC, the licensee dis-l  covered a previously installed blank flange in the turbine steam leak off line which caused steam leaks. A few problems were also noted on the RCIC governor control system. The licensee is in the process of dispositioning these items. The inspector noted that using non-nuclear steam for the testing enabled the
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identifiable using nuclear steam due to the radiological condi-l  tion The inspector will review the results of the tests and
!  dispositioning of the problems identified during the tests.
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Boston Edison Company 3 ER 111988
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Incorrect Installation of Fire Dampers On December 17, 1987, during performance of a. routine surveil-lance test the licensee' inadvertently actuated two fire damper One of the dampers failed to fully close - due to- interference with a hook used to secure it in the~ open position. When the fusible link was energized, the metal damper retaining strap should. have fallen away allowing full closure. The hook attach-ing the strap to _the fusible link was oriented -with the open side toward the damper. The damper caught on the hook and re-mained partially open. Upon discovery the licensee .immediately stationed fire watches at all areas containing suspect' damper Inspections were promptly conducted and it was identified that all of the installed hooks were oriented in this manne The hooks were repositioned so that the open side faces away from the damper. Three dampers were inaccessible and compensatory measures-remain in place pending inspectio The dampers were originally supplied to the licensee without the hook A revision to the plant design change (PDC) package added the hooks to facilitate surveillance testing. Installa-tion instructions contained in the PDC specified hook orienta-tion with the open side toward the damper. The vendor data sheet supplied by Air Balance Inc. also showed the hoox instal-led in this manne Licensee event report (LER) 87-020-00 was issued describing the problem and corrective actions taken. The LER states that pre-
  . liminary licensee assessment of the issue determined that it did not. meet the reporting threshold of 10 CFR Part 21. The inspec-tor discussed the Part 21 reportability with the licensee's Nuclear Engineering Department (NED). NED personnel stated that the failure mechanism wn created by the licensee when the hook was added. .In addition the presence of mitigating factors such as fire detection and suppression, and control of combustible materials support the conclusion that a substantial safety hazard did not exis The licensee also feels that LER 87-020-00 contains sufficient information to clearly define the problem. The inspector had no further questions in this are The inspector examined two dampers in the cable spreading room to verify that the hooks had been reoriente Both hooks had been modified, however, neither of the dampers had locking rings installed at the hook to retaining strap connection as required by the installation instructions in the PD The licensee reviewed the function of the locking rings and concluded that they were not required. A change to the PDC was initiated to delete the rin The inspector had no further question . ..
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d RI:DRP RI:DRP RI:DRP h,R1:DRP i DRP ((CWarren/mjd s  RBlough MJWiggins v SCollins Kane
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  'V(g\c . $3/3/88 0FFICIAL RECORD COPY IR PILGRIM 87-57 - 0003. ) y T. '% q i e g /ff Sp 03/05/88
b. Radiation Protection and Chemistry
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Locked High Radiation Area Access Control During the period covered by inspection report 87-57, four instances occurred in which the licensee failed-to' properly con-trol access to areas that had been designated as locked high radiation areas. In three of these cases, doors to. locked high radiation areas were found closed but not locked and in the fourth case a door into a locked high radiation area was found to not be on the list of doors that were being controlled under the locked high radiation area door procedur On December 15, 1987, a contract painter failed to check that the door to the locked high radiation area he was exiting was properly latched. The unlatched door was identified during the next routine check of high radiation area doors. Licensee par-sonnel immediately latched the door and initiated a radiological occurrence report (ROR) to document the occurrence and track all actions taken during the investigatio Surveys of the area showed no dose rates greater than 1000 millirems per hour (MR/hr). Interviews with the individual involved determined that the procedures and requirements were well understood and that the HP technician had informed them of their responsibil-
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ities prior to entry into the are On December 27, 1987, and again on January 8,1988, instances similar to the one described above took plac In both cases the licensee initiated RORs and took steps to determine: 1)who had been in the area, 2) were they aware of the procedure, and 3) had they been properly briefed. prior to entry into the areas involved. In both of these cases the root cause has been deter-mined as personnel error.
 
4  In one instance the licensee identified that one of the multiple  s doors into an area classified as a locked high radiation area was not on the list of doors to be checked on a routine basi The door was immediately checked and found to be locked. Records have been audited to determine if any unauthorized entry into the area had occurred and no instances were identified. The door has been placed on the list and is now routinely checke .
The inspector reviewed licensee actions as a result of these instances and is satisfied that in all cases, the immediate and followup actions were timely and complete. Surveys taken were comprehensive and conducted almost immediately af ter discovery of unlocked area Dose calculations were performed and dosimetry read in all case Involvement by senior HP and plant management was evident in all instances.
 
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Inadequate control of locked high radiation areas .has been an area of longstanding NRC concern. Notices of Violation have been issued in the past, during  inspectior,# 50-293/87-03, 50-293/87-11', . and 50-293/87-19 - which addressed these concern 'In regard to these violations the licensee instituted corrective actions which have been successful in addressing segments of the-
  -problem but have not been successful-in preventing recurrence of events involving high radiation area door contro The inspector has independently reviewed the licensee's program for control of high radiation areas and high radiation area key control and has found them adequate. Although the programs themselves are adequate and personnel have been trained on those programs, instances still occur where locked high radiation areas are not adequately controlle Based on review of these four instances coupled with the review of Unresolved Item 87-50-08, the inspector determined that the licensee actions in response to these previous findings have not prevented ' recurrence, Failure to comply with the requirements of Technical Specification 6.11 and Implementing Procedure G.1-012 is an apparent violation of NRC requirements as docu-mented in Appendix A of the cover letter to this report (87-57-01). Licensee response to Appendix A should include those measures taken to insure that corrective actions are effective and lastin Contaminated Clothing Offsite On December 17, 1987, at 7:26 hours a Bechtel pipefitter who-was exiting the reactor building, set off a whole body por-tal monitor alar The portal monitor indicated contamination of his chest area and left hand. The health p5ysics technician on duty at the access point removed the individual from the por-tal monitor and began performing a survey using a RM-14 with DT 260 prob The HP technician identified; 1) contamination on the individual's left hand, 1-2 thousand dpm per 100 square centimeters (K OPM), which was removed by washing, 2) contamina-tion on the shirt in both the chest (80K OPM) and lower stomach area (1K DPM). The shirt contamination was removed by tape (80K OPM) and washing with soap and water (1K DPM). The employee, now wearing an undershirt and trousers, was then sent to clear the portal monitor which again alarmed and indicated contamina-tion in the chest are The HP technician again surveyed the individual and identified contamination on the undershirt in the chest area (70K DPM). The individual was then sent into the portal monitor bare chested and was cleared. The individual was given his outer shirt, which was still wet from decontamination and cleared through portal monitor. At this point, the indi-vidual removed the wet shirt, put on his jacket, cleared the portal monitor again, and left for his hom i
 
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Upon returning to work December 18, 1987, the individual was given a whole body count to determine if any internal contamin-ation had occurred. The whole body count showed no internal'
contamination. After completion of the whole body count the individual was interviewed to determine how he had been contam- '
inated, where the occurrence took place and how long he was
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contaminated prior to detection, to calculate skin dose receive The interview revealed that the individual had been contaminated when he disconnected a partially pressurized service air hose and . depressurized i The interview also revealed that the individual used the portal monitor at the 91 ft, elevation of the reactor building, received an alarm, did not call for HP assistance but instead tried to decontaminate himself prior to proceeding to the reactor building acces Station procedures require that an individual who finds himself contaminated is to call health physics for assistanc The individual stated that he was aware of this requiremen During the interview the individual expressed concern about whether his heavy winter jack ' could have shielded the contamination on his shirt and undershirt from detection by the portal monitor To demon- ,
strate that this could not happen, a HP supervisor placed plastic bags, which contained the contamination removed from his shirt, inside the coat and attempted to exit through two por-tals. The portal monitors alarmed on each attemp The indi-vidual appeared satisfied with the demonstration put his jacket back on, with the plastic bags removed and attempted to leave the reactor buildin An alarm was actuated on the portal monitor and contamination was indicated on the left ar The on duty HP technician removed the individual from the portal monitor and identified 3K DPM contamination on the upper right sleeve (outside) of the jacket even though the jacket had not been worn into the reactor building. At _this juncture the indi-vidual expressed concern over whether the shirt that he had worn the previous day could still be contaminate The licensee had a HP technician accompany the individual to his hom The individual's shirt was found to be contaminated, was bagged and returned to the sit Surveys of the individual's home and vehicle identified no further contaminatio Efforts to determine how the contaminated shirt was worn through the portal monitors without setting of an alarm yielded positive results. The individual stated that he had purposely kept him-self away from the portal monitor in an attempt to keep his wet shirt away from his ski The licensee taped the plastic bags, with the contamination in them, back onto the snirt and an HP supervisor attempted to pass through the portal monitors by
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f mimicking the' body posture used by the individual _when he cicared the monitor. The HP supervisor was able to pass through six different monitors without setting off an alarm. The HP super-visor then used the portal monitors in the correct manner and all six monitors alarmed proving that the equipment was func-tiona The licensee has evaluated the occurrence to identify the root causes ~ and -immediately implemented corrective action. This occurrence was _ caused by one sequence of events that involved two distinct personnel = errors. The primary cause involved the failure of the HP technician to perform an adequate survey of
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the contaminated individual's clothing when the portal monitor alarm was received. The second problem involved the f ailure to properly use the installed portal monitors at the reactor build-ing acces In addition to personnel interviews.to identify the sequence of events the licensee also reviewed procedural adequacy, personnel-training and portal monitor calibration and performance. These F-  reviews verified that training was adequate and portal monitor performance'was as designed. Procedures for control of contam-inated individuals at the reactor building access did not spec-ifically require that all articles of clothing require a 100%
frisk prior to this occurrenc Instructions have been posted at the reactor building access which now clarify the procedure to be followed when an individual is found to be contaminate The portal monitors in use at Pilgrim do not presently have a switch at chest level. which must be actuated to start the moni-toring proces Lack of this feature allowed the individual
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wearing a contaminated shirt to lean away from the machine suf-ficiently to clear the monitor without any alarm. The licensee has determined thats the manufacture of the portal monitor now produces a chest high switch for the installed model and will install them in the futur Calculations have been performed by the licensee to determine the radiation dose received by the individual and the amount of
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radioactive material that was released from the site on the con-taminated shirt. The results of these calculations show that the individual received a localized radiation dose to the skin of 260 Mrem, which is below the federal limits for skin exposure, and that the amount of radioactive material on the individuals clothing was 0.2 microcuries which meets the federal criteria as an exempt quantity of Co-6 The inspector is satisfied with the licensee's analysis and corrective actions and has no further questions.
 
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Allegation of I'mproper Disposal of Radioactively Contaminated Shrubs . ( RI-87-A-0107)
On August 31 and September 11, 1987, the NRC resident office at t
Pilgrim received. allegations that radioactively contaminated shrubs had been removed from the site and improperly dispose .The alleged improper disposal occurred on ' July 23, August 26 a August 28,1987. During this time period the licensee removed a large number of shrubs from various areas of the site, including those - planted near the old. administration building and the switchyar The shrubs were removed to facilitate site con-struction activities _and to alleviate certain security' concern Upon receipt of .he first allegation on August 31, 1987 the NRC requested that the licensee perform -an evaluation and provide
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the results .for revie In addition an independent NRC review was initiate Resident and specialist inspectors reviewed the licensee's con-clusions. The licensee evaluated material release records and interviewed personnel regarding removal of shrubs -during the
!.. week of July 20,1987. Several truckloads of shrubs that 'were transported offsite during the midnight shif t on July 24 were
!  examined in de tail. Because trace amounts of Cobalt-60 had pre-i-  viously been found in soil onsite, some .of the shrubs had the soil removed from the roots prior to release. Each shrub was hand surveyed and found to meet established offsite release c ri te ri a . They .were transported first to the licensee's shore-front area and later to a dump site on licensee propert The licensee concluded that the shrubs had been adequately surveyed and that no radioactive material had been improperly release The resident inspectors reviewed the licensee's program for con-trol of release of material from the sit This area was also evaluated by NRC specialist inspectors during inspection 50-293/
87-19. Both inspections concluded that appropriate surveys and release liniits have been established and implemented. Resident and specialist inspectors examined licensee release records for the dates in question to verify that vehicles leaving the pro-tected area had been properly surveyed. No discrepancies were identified. An NRC resident inspector accompanied by a licensee representative collected four samples of the shrubs which had been deposited in the dump site discussed above. Each of the four samples consisted of root, branch and foliage clippings from a number of different shrubs. The samples were indepen-dently analyzed by the NRC. Three of the samples indicated no contamination. One sample indicated only trace levels of Cobalt-60. Measurements showed that the amount of C0-60 present in
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this sample was about 2% of the average radioactivity typically found in soil due to naturally occurring isotope c
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The . licensee's program for release of material from the site appears adequat Appropriate survey . techniques and' release limits-have been established. Review of -records confirmed that the program is being implemented. Samples of the shrubs col-
  .lected by the NRC showed' zero or negligible contamination and pose ' no health and safety concer Based on the above this allegat'an is considered close NRC Region I staff provided status briefings concerning this allegation to. Senator Kennedy's staff. and to the Massachusetts Department of Public Healt Allegation of Airborne Radioactivity in the Trash Compaction Facility (RI-87-A-0120)
On October 5, 1987, the resident office received an anonymous allegation that personnel working at the sort table in the trash compaction ; facility .(TCF) were being routinely exposed to air-borne radioactive contamination. The alleger stated that the two filter systems designed- to treat exhaust air from the sort -
table prior to discharge into the room were not functioning, and that the filter differential pressure alarm circuits had been disable On October 7 and 8, 1987, NRC specialist inspectors toured the TCF and examined the design and condition of the equipment. The sort table is used to separate contaminated materials for com-paction and disposal. Potentially contamitated air is exhausted from the table, passed through two filters operating in parallel,.
and released into the room. Airborne radiation levels in the room are measured by means of a separate air monitor which is operated whenever the sorting table is use The alarm is
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typically set at 3 X 10 -10 (3E-10) microcuries per cubic cen-timeter (cc). In addition the filters are surveyed daily and changed if contact dose rates exceed 2mR per hour. The inspec-tors also examined the trash compaction unit in the area and found that similar controls had been applie Based on the above, no immediate health and safety concerns were indicate On January 15, 1988, the resident inspectors toured the TCF, examined equipment operation and interviewed licensee and con-tractor personnel involved in ongoing work activities. A radia-tion work permit specifying protective clothing, health physics coverage, and use of a continuous air monitor is in place to  -
control work at the sort tabl Personnel involved stated that trash bags were surveyed prior to sorting and rejected if radia-
,  tion levels exceeded Smr/hr, if they contained liquid, or if any l  powdery material was presen The health physics technician on auty stated that filter radiation levels are monitored daily.
 
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Workers and health physics personnel also' stated that filter:
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filter plugging, however no one had.been clearly assigned this responsibility and no' dp -limit was. established. The inspector-observed the operation of the table and noted that .the '"filter restricted" alarm ' actuated - for one of the two filter The alarm actuated ' for the filter displaying - the -lower differential pressure.- When questioned workers stated ' that much of the -
monitoring and alarm circuitry for the table was not functional, and that - the filter alarm was. not reliabl The . table was originally part of a larger processing system and much .of the disconnected circuitry was intended to perform functions which are no longer needed. The f aspector verified that current filter dp readings are consistent with the manufactures name plate dat ;It appears that the general-process applied, incibding insoec-tion ana. survey of trash bags' prior to sorting, daily filter surveys . and continuous air monitoring would preclude airborne radioactivity problems. Based on the above this : all.egation is closed. However, th'e inspector noted that no work instructions existed-describing the controls applied and equipment monitoring requirements. When discussed with licensee radiation protection management they promptly committed to review the situation and-issue appropriate guidanc This was confirmed during the inspector's exit intervie Erosion of Construction Dirt'into Wetland On January 15, 1988, at 5:45 the licensee made an ENS notificatfor in accordance with 10 CFR 50.72 (b)(2)(vi) which requires the licensee to inform the NRC of an event or situation related to health and safety of public for.which a news release
  .was made or notification of another government agency has been mad During routine environmental monitorin'g, the . licensee observed erosion from a pile of construction dirt into an adja-cent licensee controlled wetlan The Plymouth Conservation Commission and the Massachusetts Department of Public Health were notified and the press release was made by the license Also on January 16, 1988 two representatives from the Plymouth Conservation Commission toured the are In the last several years during onsite excavation for plant modifications, dirt, asphalt and concrete containing low le"els of contamination were stored in a fenced in storage area outside the protected area on the licensee's propert The licensee estimated that the storage area contains 110,000 cubic feet of materia Before removal from the protected area, samples of
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material were obtained and isotopic analyses was performed by the licensee. The activity found was - reasonably uniform at1 levels of 10(1E-6) and 10(1E-7) microcuries of Cobalt-60 and Cesium-137.per gram. Sampling and- storage of this material was previously reviewed during inspection 50-293/87-1 On January 21, 1988 the inspector toured the area, accompanied b a licensee health physics technician, and performed a survey of
  'the storage area and found no detectable radiation above .back-ground levels. During the tour the inspector noted that bales of hay had been put around the perimeter of the fence which borders wetlands area to prevent further erosion . of materia The fenced in storage area was secured with a locked gat The inspector's survey of the area and review of licensee's analyses indicate that the level of activity does not represent a health or safety concern. However, the inspector raised a concern to the licensee management that the material should not be -allowed-to erod The inspectors will continue to monitor the licensee actions in formulating long term solution to properly dispose of the materia Fire Protection On January 17,1988, at 4:55 a.m. the control room received a report from a security guard of smoke coming from a contractor lavatory trailer, which is located adjacent to. the Bechtel warehouse inside'
the protected area fenc The onshift fire brigade chief was dis-patchtd to the scene and confirmed smoke and smoldering in the are The fire brigade was immediately dispatched and fire was extinguished using a portable dry chemical extinguisher and a hose from a nearby hydrant hous Electrical maintenance was called to shut off the power to the trailer. By 5:30 ~ a.m. , .the fire brigade members had cleared the scene and a continuous fire watch was posted ii, the are ' The cause of the fire was believed to be overheating of an overhead heating unit for the trailer. No personnel injury . occurre The inspector toured the scene with a licensee fire protection engineer on January 18, 1988. Minor damage to a small area of the ceiling in the trailer was observed. The Plymouth Fire Department was notified by the licensee in the morning of January 18, 198 .0 Review of Plant Events The inspectors followed up on events occurring during the period to deter-mine if licensee response was thorough and effective. Independent reviews of the events were conducted to verify the accuracy and completeness of licensee informatio __    _    .. -  - - - - - -
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a. Spurious Isolations of RHR Shutdown Cooling System On December 7,1987, at 2:28 p.m. , an inadvertent isolation of both inboard and outboard containment isolation valves on the RHR shutdown cooling suction line occurre Preparation for the reactor vessel hydrostatic test was in progres As part of the hydrostatic test procedure, a technician was installing an electrical jumper in the primary containment isolation system logic panel C-941 to bypass the reactor coolant system (RCS) high pressure interlock on the inboard isolation valve. When the termination screws were loosened to in-stall the jumper, the leads lost contact and caused a false high pressure isolation signal. RHR was in its shutdown cooling mode when the isolation signal was generated, and the shutdown cooling suction valves (MOV 1001-47, 1000-50) automatically closed as designe Coincident with the closure of the valves, the "A" and "C" RHR pumps tripped automatically to protect the pumps from loss of adequate suctio The licensee determined the actuation was due to a person-nel erro The licensee revised Procedure 2.1.8.1, Class I System Hydrostatic Test, to caution the I&C technician of potential isola-tion of RHR shutdown cooling system while installing the jumpe On December 8, 1987, at 9:45 p.m. , the inboard isolation valve (MOV 1001-50) on the RHR shutdown cooling suction line automatically close The automatic isolation occurred when the plant reached 100 psig during pressurization for performance of the class I hydro-static tes The outboard isolation valve (MOV 1001-47) was already closed to form a pressure boundary for the tes The licensee's investigation determined that the cause of the isolation was that Procedure 2.1.8.1 did not identify all the jumpers necess4ry to b3 pass the RCS high pressure interlock on the inboard isolation valv As immediate corrective action, the licensee halted the pressuriza-tion of RCS and reviewed the logic print The licensee revised Procedure 2.1.8.1 to reflect the need to install an additional jumper in panel C-94 In reviewing this event along with other similar events documented in previous inspection reports, the inspector noted that inadequate planning and inadequate procedures appear to be a common root cause for several ESF actuations which occurred on September 17, september 22, October 15 and October 24, 198 The inspector expressed this concern at the exit meeting with licensee managemen The licensee informed the inspector that the Technical Group is in the process of developing generic guidance for isolating or jumpering an electrical component which may cause inadvertent safety system actuations. The inspector will continue to monitor the effectiveness of licensee's corrective action to prevent further ESF actuations due to inadequate planning and inadequate procedure l
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b. Reactor Water Cleanup System Spurious Isolation On December 17, 1987, at 11:05 a.m., the inboard primary containment isolation valve on the reactor water cleanup (RWCV) system suction line automatically isolate I&C technicians conducting a routine surveillance of the RWCL' high area temperature isolation logic inad-vertently grounded a lead which had been lif ted during the tes Grounding the lead resulted in a blown logic power fuse and isolation of the valve (MOV 1201-2). Following investigation by the control room supervisor, the fuse was replaced and the isolation was rese The licensee's investigation concluded that the root cause is a per-sonnel error. The licensee informed the inspector that the proced-ure, 8.M.2-1.2.2, Reactor Water Cleanup Area High Temperature, will be revised to provide cautions to the control room operators and the I&C technicians. Also, an effort is ongoing to review recent ESF actuations caused by personnel error to formulate appropriate corrective action c. Engineered Safety Feature Actuations Due to a Failed Logic Relay On January 6, 1988, at 2: 50 p.m. , the coil of primary containment isolation system (PCIS) electrical relay 16A-K57 failed, creating a f aul t and resulting in blown logic power fuse The aeenergization of this portion of the PCIS logic caused a partial primary contain-ment isolation along with a reactor building isolation and start of the "8" Standby Gas Treatment system (SBGT).      The licensee notified the NRC at 5:12 p.m. via ENS. The failed relay was a GE type CR120A rela The licensee has experienced several failures of this type of relay in the last few years. The licensee's evaluation of this nigh failure rate and corrective actions to address it are described in the inspection report 50-293/87-5 On January 7, 1988, the inspector reviewed maintenance request (MR)
88-9 which had been initiated to investigate the cause of the above mentioned ESF actuations and to replace the blown fuse and the faulty rela The inspector noted that the relay replacement was performed using only procedure 3.M.1-11, Routine Maintenance. This procedure contains general guidance and its stated use is for performing main-      '
tenance activities which are not complicated or critical enough to require detailed written procedure In this case, no step-by-step instruction was initiated to control the sequence of work, to control and tag lifted leads and jumpers, and to ensure verification and independent verification of system restoration. A similar problem involving lack of a sufficiently detailed controllir g procedure and the appropriate reviews during an electric relay replacement on November 24, 1997 was the subject of a violation as documented in the inspection report 50-293/87-50, The licensee informed the inspector that the corrective actions to address the violation are being formulated and will be submitted to the NR _ _ _ _ _ _ _ _ __ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __
 
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d. Spurious Reactor Protection System Actuation On January 17, 1988, at 1:13 a.m.,  a spurious reactor scram signal was generated during the performance of a reactor level instrument calibration. The full scram signai on low water level was received due to a disturbance in the reactor water instrument line when an I&C technician was valving a level instrument (LI-263-59A) back sin ser-vice. The Rosemount level transmitters (LT-263-57 A&B) which initi-ated the. scram signal are on. the same ~ instrument rack. The licen-see's preliminary investigation indicated that the root cause of the event is attributed to a combination of personnel error and inade-quate procedur The investigation also identified that' the level instruments (LI-263-59 A&B) were incorrectly installed in that the sensing lines were reverse The new Barton level instruments (LI-263-59 A&B) were recently installed during this outage and would only be used for local indication during a shutdown from outside the control room. The licensee is currently reviewing the plant design change (PDC 85-07) records and post-installation test data to deter-mine the caus Surveillance test records are also being reviewed by the license This item -is unresolvea pending the completion of the licensee investigation (87-57-02).
 
Upon receiving the spurious scram the control rnom staff noted that scram discharge i' :trument volume (SDIV) vent valve CV302-238 primary containment vent and purge valves A050448 and A05035B and one of two redundant secondary containment isolation dampers in each line did not close. In addition the "B" standby gas treatment system (SGTS)
did not star Based on the initiating event, these components should have actuated. The licensee notified the NRC of the failures via ENS at 5:00 a.m. on , January 17, 1988.
 
l  The control room staff conducted an immediate critique with available I&C personnel, and documented observations for management followu Later on January 17, the licensee inspected the physical condition l  of the SDIV vent and drain valves and noted paint on the stem of CV302-238. The paint was removed and the va'.ve successfully stroke time The licensee htid a second critique with management repre-sentatives on the morning of January 18, 1988 to assess the situa-tion. Subsequently, a walkdown of involved isolation logic components was performed to verify relay contact configuration cnd to identify any jumpers or lif ted lead This walkdown was performed to the extent possible without disturbing components. No discrepancies were noted. Early on January 19, the licensee performed a test in which a reactor scram was intentionally initiated. The same equipment failed to actuate as during the January 17 scra Based on this licensee management stopped all work on the affected components. A task force composed of members from the technical staf f, systems group, I&C and operations was designateo to investigate tne inciden This team reviewed available information and developed an action pla _ _ _  _ _ _ _ _ _ _ _    __
 
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Walkdowns of the air system piping and components supplying motive air to SDIV vent valve CV302-238 were performed to verify ' hat the as built configuration is in accordance with design doc  's and that components are in good physical conditio No disteepancies were identified. Valves CV302-238 and CV302-228 are supplied air by the same solenoid operated valves. The licensee cecnergized these solenoid valves and observed that CV302-22B closeo while CV302-233 did no This indicates a mechanical problem with the valve or operato The licensee was identifying replacement parts and pre-paring to disassemble the valve by the close of the inspection perio The inspectors will continue to monitor licensee followup to this failur Licensee review of logic drawings confirmed that the remaining equip-ment which had not properly actuated shared common isolation logic component A series of survefilance tests was performed to allow
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monitoring of key relay ac%ations. A single contact on a General Electric (GE) HFA relay was determined to be misfunctioning. The contact is required to close when an isolation s1gnal is received, actuating the affected equipmen However, contact resistance remained high with the contact close The relay was replaced and the system successfully teste The licensee contacted GE to coor-dinate disassembly and inspection of tt.e relay. Dissassembly had not begun by t he. close of the inspection perio The inspector will continue to monitor licen ee investigation of this failur The inspector exprev.ed concern that ^hree separate equipment mal-functions had occurred during the inadvertent actuation. This may reflect weakness in the surveillance and post-work tect progra However, the licensee's response to the a c'.ua ti on and subsequent malfunctians was prompt, thorough and effec v Control room aper-ators quickly recognized each of the failures. They held a critique on the same shift with insclved personne Critique observations -
were clearly documented and provided to nanagement. An additional critique with managercent present established pri( rities. Action was taken to freeze equipment until a r. investigation pin could be developed and implemente Followup was well cocrdinated and in-volved representatives of several portions of the organization. In this case licensee commitmert to determining and correcting the problem root cause was eviden '
Review of _ Reactor Vessel Hydrastatic Test Procedure c,d Test Results Duriew - 1spection period the licensee comp.eted the reactor vessel
  ; : est. Several reactor vessel instrument nozzles were repaired outage, prompting performance of a hydrostatic test rather n leakage tes The reactor vessel reached minimum test
  . a all inspections were completed on December 9, 198 Only i -- age associated with mechanical connections, such as flanges and ni ting was identifie The reactor vessel was depressurized on Deco -
12, 1987 af ter completioc of excess flow check valve testing.
 
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The_ inspector reviewed the licensee's hydrostatic test procedure to verify that appropriate prerequisites, precautions and instructions had 5een
  : included. A sample of valve lineups was reviewed to datermine the ade-quacy of established test boundaries. Completed valve lineups were~also examined. -Control of temporary electrical and mechanical jumpers was evaluated to ensure proper documentation and restoration. The-inspector observed installed pressure instrumentation and verified appropriate range and calibration status. The adeouacy of staffing - to support test per-formance was periodically verified. The inspector reviewed test results and discussed them with engineering, operations, and quality control oersonnel to ensure that test changes were properly processed, adequate inspections were conducted, and that inspection results were promptly dispositione \
The licensee's preparation for and execution of the test was generally well organized and controlled. Procedures for test performance and con-duct of visual inspections were clear and comprehensiv A detailed
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Quality Control (QC) work instruction was developed specifying components-and piping requiring inspectio Inspection assignments were broken down by location, elevation and component. This QC instruction also included a series of piping diagrams depicting the test boundaries which were utilized to assist in inspection performance and documentation. The licensee's Technical Engineering Section, Quality Control staff and Nuclear Engineering Department each reviewed test boundary adequacy. In-spection results were well documented, and maintenance requests were promptly initiated to correct identified leakag The licensee experienced two shutdown cooling isolations during implemen-tation of the test procedure. These isolations are discussed in detail in section 4.a of this report. During the test the licensee identified leak-age past the seal ring at the stuffing box to pump casing joint on both racirculation pumps. Leakage flow was estimated to be one to two gallons per minute for each pump. The leakage wet the pump casings and portions of the suction piping, and acceptable inspections could not be completed
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in these areas. The licensee stated that similar leakage on at least one of the pumps was noted during the last outage. That leak sealed as system
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temperature increased during startu The licensee believes that the leakage observcd during the recent test will also stop as temperature is increasM, and no pump repairs are planned. The licanses stated at the ins.nector's exit interview that the pump casings and suction piping will be reinspected during start; .
The inspector noted that the test procedure did not contain valve lineups for manual instrument isolation valves within the test boundary. Many instruments and a significant portion of instrument piping has been replaced this outage. 'lisuai inspections were performed of clas; I piping
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downstream of these v31ves. The inspector questioned the basis for licen-see confidence in instrument line isolation valve positions during the test. The licensee pointed out that hydrostatic :esting of these lines was not required dur'ng this outag In addition excess flow check valve
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testing was conducted immediately cf ter completion of the hydrostatic test with the system still pressurize Successful completion of the check valve testing requires proper alignment of the manual isolation valves, and provides assurance that the piping was pressurized during the visual inspections. Tre licensee however, agreed that the intent of the test had been to pressurize and inspect this piping and that the current procedure does not adequately assure the correct valve alignment. Licensee manage-ment stated that the procedure would be revised to address this weaknes .0 Integrated Leak Rate Testing On December 21, 1987, the licensee began performance of the primary con-      $
tainment integrated leak rate test (ILRT). The containment was pressur-ized with air to the full test pressure of 45 pounds per square inch and maintained at this pressure for 24 hours. The 24 hour test period started at 10:15 p.m. on December 21, 1987. A regional specialist inspector was onsite during the ILRT to review the adequacy of the test procedure and to observe the conduct of the tes The preliminary licensee test results indicated a succe .sful test, with measured leakage slightly greater than 20 percent of the allowable leakag A primary contributor to the ob-cerved leakage was identified as a drywell pressure transmitter piping cap which had not been fully tightene Upon completion of the specialist inspector's review of the ILRT results, inspection report 50-293/87-58 will be issued documenting the inspectors finding While preparing for the primary containment integrated leak rate test (ILRT) the licensee observed that several torus temperature and moisture elements were not functioning properl Troubleshooting identified cir-cuit faults at a torus electrical penetration assembl The licensee removed the penetration assembly protective cover inside the torus and found that it was filled with water. The penetration is installed ver-tically through the top of the toru On both the inboard and out: 3rd sides of the penetration a metal frame is attached on which 28 terminal boards are mounted. Cables passing through the pe ietration, and supplying instrumentation in the torus also landed on trese terminal board A protective cover is bolted over the f rame n.u terminal boards on both sides of the penetratio Design drawings specify that cover joints are to be sealed with s'licone tape. The licensee stated that the protective cover had not been oroperly sealed, allcwing water intrusion and buildu The water caused significant corrosion of the cable connectors, terminal boards and metal framework. This corrosion and water buildup resulted in the observed electrical circuit fault Licensee inspection of the other torus electrical penetration identified similar condition Temporary repairs of the temperature and moisture elements were made to allow ILRT performanc Cables for communications, lighting, and torus to drywell vacuum breaker indication also run through the penetratio The penetra-tion is not considered by the licensee to require environmental qualifica-tion but is designated as a "Q" component. The licensee is evaluating the root cause of the water intrusion and is developing a temporary procedure
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to control repair and testing of the penetration. The inspectors will
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continue to monitor licensee followup and corrective action . _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ - _ _ _ _ - - - _ _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ .
 
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The licensee informed the inspector that penetration repairs would not be completed until after ILRT performance. The inspector questioned the
  .effect of the planned repairs on the penetration leak tightness, and the ability to perform adequate leakage test after the planned rework. The licensee stated thr.t the work would not affect penetration leakage but that-adequate testing could _ be performed after work completion. Based on available drawings however, the licensee could not demonstrate adequate testability. In response to NRC concern the licensee obtained the needed drawings from the vendor and verified that the penetration was completely testable. The inspector had no further question During the ILRT, the licensee identified a water leak in the high pressure coolant injection (HPCI) turbine roo It was determined that the in-creasing pressure in the torus air space caused the suppression pool water to back up through the HPCI turbine exhaust line and throgh the drain
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piping, overflowing the HPCI gland seal condenser onto the HPCI room floo The turbine exhaust line disenarges to the torus through a check valve .and a locked open stop-check valv To prevent any condensation from collecting in the turbine exhaust line downstream of the check valve, a drain piping drains any condensation to the HPCI gland seal condenser through a drain pot. Two solenoid operated drain valves on the drain pot close automatically on a HPCI (Group IV) isolation signa This is to provide the isolation from the torus to the gland seal condenser. The licensee's investigation determined that leads had been lifted in the HPCI isolation interlock logic circuit since October 30, 1987 in support of the HPCI steam testing . utilizing temporary oil-fired auxiliary boilers. With the HPCI isolation signal bypassed, the drain valves remained open as the
  - drain pot was filleJ with the suppression pool water. The licensee sub-sequently relanded the leads in the HPCI isolation interlock logic circuit and the drain valves close After reviewing the ILRT procedure, HPCI test procedure and interviewing licensee personnel, the inspector concluded that licensee review of the active maintenance requests prior to the ILRT was not thorough in that the lifted leads controlled by the MR 87-663 were not identified. The LtR tags were attached on the HPCI isolation logic circuit inside a logic panel and thus the tags were not identified during a system walkdown prior to the ILRT. The dra:n valve positions were verified by the light indications on the control room panel 903 as prescribed in the ILRT procedur The inspector also determined that the maintenance  <. quest above may not be an adequate method of identifying and tracking jumpers and lifted  .
leads, especially for a long term application and for components which could affect other ongoing maintenance or surveillanc Station proce-dures do not require temporsy modification controls ;ur jumpers and lif ted leads which are controllea by active maintenance request The inspector discussed these findings at the exit interview with licensee
  , management. The licensee informed the inspector that a lifted leads and jumper log will be kept in the control roon to aid the operators in con-trolling lif ted leads and jumper . - -
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l 7.0 Licensee Nuclear Organization Management Realignment On December 14, and on December 31, 1987, the Boston. Edison Co. announced,
  .as . part of a planned realignment occurring over the next several weeks, the appointment of the following managers to -key management positions in-the licensee nuclear organization at Pilgrim Statio Mr. Kenneth L. Highfill was named to assume the new position of Station Director. In this capacity, Mr. Highfill will oversee day to day operation of the Pilgrim Station including plant operations, planning and outage, nuclear training, plant- support' functions, and administrative services. Mr. Highfill will report directly to Mrb  '
Ralph G. Bird, Senior Vice President-Nuclea Mr. Robert J. Barrett was named the new Plant Manager. Mr. Barrett  ,
will report to Mr. dighfill, the Station Directo Mr. Roy Anderson, currently Deputy Outage Manager, was named to assume the new position of Planning and Outage Manage Mr. Anderson will report to Mr. Highfill, the .itation Directo Mr. Ed Kraf t was named to assume the new position of Plant Support Manage In this capacity, Mr. Kraft will oversee radiological, sacurity, industrial safety and fire protection, and other station support function Mr. Kraft will report to Mr. Highfill, the Station Directo Mr. Donald Gillespie, currently Director of Planning and Restart, was appointed to the position of Quality -Assurance Department Manage Mr. Gillespie will assume the position after completing his Senior Reactor Operator' training. The Quality Assurance Department Manager reports to Mr. J. E. Howard, Vice President-Engineerin M Frank Famulari, currently Operations Quality Control Group Leader, was named to assume the newly created posi tion of Deputy Quality Assurance Department Manane Mr. Famulari will report to Mr. Gillespie, and be acting Department Manager until Mr. Gillespie assumes the position after completing the Senior Reactor Operator trainin Mr. John F. Alexander was named to assume the position of Operations Section Manager. Mr. Alexander will report to Mr. Barrett, the Plant Manage Mr. Donald J. Long was named Security Section Manager. Mr. Long will report to Mr. Kraft, the Plant Support Manage . ~ . , - . - . - . - - . - - , . - - . . - . - . - . .  . . - . . - - , - - , - - - - , . . ,
 
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.      l 24  l 8.0 Management Meetings l
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 inspector On January 26, 1988, the inspectors conducted a final inspection exit interview to formally present inspection finding FE '
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Attachment I to Inspection Report 50-293/87-57 Persons Contacted  ,
  * R. Bird, Senior Vice President - Nuclear
  * K. Highfill, Station Director K. Roberts, Plant Manager R. Barrett, Deputy Plant Manager R. Anderson, Planning and Outage Manager E. Kraft, Plant Support Manager F. Famulari, Deputy Quality Assurance Manager D. Swanson, Nuclear Engineering Department Manager J. Alexander, Operations Manager N. Brosee, Maintenance Manager J. Jens, Radiological Protection Manager J. Seery, Technical Manager R. Grazio, Fie!d Engineering Manager P. Mastrangelo, Chief Operating Engineer
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R. Sherry, Chief Maintenance Engineer N. Gannon, Chief Radiological Engineer D. Long, Security Manager F. Woznick, Fire Protection Manager
*Sonior licensee representatives present at the exit meetin __  .-
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      - ATTACHMENT.1II January 6, 1988
  - MEMORANDUM'FOR:  Ken = Roberts
    . Plant Manager FROM:  . Clay Warren-      -
    - Senior Resident Inspector - Pilgrim SUBJECT:  FACILITY TOUR FINDINGS, DECEMBER 8,198 .
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  - The items' on - the~ attachment were noted during the - facility tour on December 8,.198 Please-contact the Resident inspector Office when your staff is ready to discuss the evaluation of-the-items and the status'of any actions
  - taken. Please note. the items and the facility response will be. addressed in a
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routine inspection. repor Thank you-for your time and attention to these matter
 
Sincerely, i
Clay C. Warren Senior Resident Inspector    ,
Attachment:          ;
As stated          ,
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R. Blough
;. W. Kane W. Russell
   - J._Wiggins
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ATTACHMENT
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Numerous motors appear to have failed grease seals caused by overgreasing without first removing grease drains. This condition causes a buildup of grease and dirt in the cooling airflow path and in extreme cases grease in the motor winding (SBGT fans and SLC pumps)
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Nuts and bolts were noted 10ying inside an electrical cabinet in the RCIC roo Multiple cases of open junction boxes, terminal boxes and conduit pulled away from terminal boxes were note Motor heaters for the "B" RHR pump appear to have overheated causing the insulation on the heaters to mel HPCI room cooler drip pan is full of paint scrappings which could lead to drain cloggin Standby Liquid Control system relief valves have boric acid c rystal buildup which could alter setpoint Painting effort should be more closely controlled to prevent painting inappropriate surfaces, i.e., linkages, valve packing glands, trip throttle valves, limit switches, et Numerous instances of scaffolding materials, i.e., nails and wood chips, laying on floors. This material could migrate to drain systems and cause pump or valva damage. S:af folding was also noted attached to permanent equipment such as piping and condui Valve 1001-36A niotor operator conduit had melted plastic cove :
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Latest revision as of 01:18, 17 December 2020

Forwards Safety Insp Rept 50-293/87-57 on 871207-880119 & Notice of Violation.Equipment Failures Indicate Need for Stronger post-work Test Practices & Thorough Power Ascension Test Program
ML20195B175
Person / Time
Site: Pilgrim
Issue date: 03/11/1988
From: Collins S
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To: Bird R
BOSTON EDISON CO.
Shared Package
ML20153A853 List:
References
NUDOCS 8803210184
Download: ML20195B175 (3)
v • d • e


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MAR 1l11988

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Docket No. 50-293 Boston. Edison Company ATTN: Mr. Ralph G. Bird Senior Vice President - Nuclear 800 Boylston Street Boston, Massachusetts 02199 Gentlemen:

Sub.iect: Region I Inspection Report No. 50-293/87-57 This refers to the routine safety inspection (50-293/87-57) conducted by

' Messrs. C. Warren, J. Lye.sh and T. Kim of this office on December 7,1987 to January 19, 1988:at the Pilgrim Nuclear Power Station, Plymouth, Massachusett Areas examined during this inspection aredescribed in -the NRC Region I Inspection Report. which is enclosed with this lette Based on the resuits of this inspection, it appears that one of your activities related to high radiation area access control was not conducted in full com-pliance with NRC requirements, as set forth in the Notice of Violation enclosed herewith as Appendix A. - The problem was identified by your staff. However, a Notice of Violation is being issued because effective corrective actions apparently liave not been taken for previous problems with high radiation area-access control. In addition to following the irstructions of Appendix A in'

preparing the required response, please include chose actions you intend to take to preclude recurrence of this problem by insuring that your corrective actions are effective and' lastin Two significant integrated plant tests were successfully executed during-the inspection period. Preplanning and control of these activities was generally stron We also observed that increased management involvement in assuring effective problem followup has resulted in substantial improvement. Equipment failures identified as a result of an unanticipated safety system actuation however, indicate the.need for stronger post-work test practices and a thorough power ascension test progra The response directed by this letter and the accompanying Notice are not subject to the clearance procedures of the Office of Management and Budget as required by the Paperwork Reduction Act of 1980, PL 96-511.

L OFFICIAL RECORD COPY IR PILGRIM 87-57 - 0001. , 03/11/88 8803210184 880311 C A PDR ADOCK 05000293 .(/

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Boston Edison Compan MAR 111988

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Your cooperation with us in this matter is appreciate

Sincerely,

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dgginal Stcncd I?2

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Samuel J.' Collins, Deputy Director Division of Reactor Projects Enclosures: Appendix A, Notice of Violation NRC Region I Inspection Report No. 50-293/87-57-cc w/encis:

R. Barrett, riuclear Operations Manager B.-McIntyre, Chairman, Department of Public Utilities Chairman, Plymouth Board of Selectmen Chairman, Duxbury Board of Selectmen Plymouth Civil Defense Director J. Keyes, Boston Edison Regulatory Affairs and Programs E. Robinson, Nuclear Information Manager R. Swanson, Nuclear Engineering Department Manager >

, The Honorable Edward J. Markey *

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The Honorable Edward P. Kirby The Honorable Peter V. Forman S. Pollard, Secretary of Energy Resources P. Agnes, Assistant Secretary of Public Safety, Commonwealth of Massachusetts R. Shimshak, MASSPIRG-Public Document Room (PDR)

Local Public Document Room (LPDR)

Nuclear Safety Information Center (NSIC)

NRC Resident Inspector Commonwealth of Massachusetts (2)

bec w/encls:

Region I Docket Room (with concurrences)

W. Russell, RA

.M. Perkins, ORMA (w/o encis)

R. Blough, DRP L. Doerflein, ORP R. Bores, ORSS S. Collins, DRP C. Anderson, DRS D. Mcdonald, LPM, NRR

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T. Chandrasekaran, SPLB, NRR M.~ Callahan, OCA J. Nickerson 0FFICIAL RECORD COPY IR PILGRIM 87-57 - 0002. /11/88

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Boston Edison Company 3 ER 111988

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'V(g\c . $3/3/88 0FFICIAL RECORD COPY IR PILGRIM 87-57 - 0003. ) y T. '% q i e g /ff Sp 03/05/88

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