IR 05000293/1987018

Insp Rept 50-293/87-18 on 870407-0518.No Violations Noted. Major Areas Inspected:Plant Operations,Radiation Protection, Physical Security,Plant Events,Maint,Surveillance,Outage Activities & Repts to NRC

ML20215A289
Person / Time
Site:	Pilgrim
Issue date:	06/03/1987
From:	Wiggins J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20215A276	List: IR 05000293/1987018 ML20215A273
References
50-293-87-18, GL-84-23, IEIN-83-63, IEIN-85-082, IEIN-85-82, NUDOCS 8706160598
Download: ML20215A289 (25)

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U. S. NUCLEAR REGULATORY COMMISSION

REGION I

Docket / Report N /87-18 Licensee: Boston Edison Company 800 Boylston Street Boston, Massachusetts 02199 Facility: Pilgrim Nuclear Power Station Location: Plymouth, Massachusetts Dates: April 7, 1987 - May 18, 1987 Inspectors: M. McBride, Senior Resident Inspector J. Lyash, Resident Inspection T. Kim, Resident Inspector L. Plisco, Senior Resident Inspector (Susquehanna)

T. Foley, Senior Resident Inspector (Calvert Cliffs)

H. Kaplan, Lead Reactor Engineer Approved By: be '

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J.A'iggins, C(/eg, Reactor Projects

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%ction 18, DRP Inspection Summary:

Areas Inspected: Routine resident inspection of plant operations, radiation protection, physical security, plant events, maintenance, surveillance, outage activities, and reports to the NR Results: No violations were identif ted. Inspector review and evaluation of the licensee's response to radiological protection program issues identified i in inspection report 50-293/86-41 indicates that certain corrective actions may need further review (Section 5). The following observations were also made by the inspectors during the period: A number of procedure changes were implemented without required review and approval (Section 3.b).

' Improperly performed heat exchanger maintenance which contributed to tube sheet and divider plate erosion (Section 3.b). Recurring problems with use of improper procedure revisions (Section 3.c).

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8706160598 070610 PDR ADOCK 05000293 G PDR

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- :) Inspection Summary (Continued)

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4. Successful planning and execution of an endurance run on the B emergency diesel generator (Section' 3.c).

5. Poor training and communication.which contributed to failure of the diesel generator lockout relay (Section 4.a).

6. Improper restoration of instrumentation after modification testing which

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resulted in an inadvertent ECCS initiation signal (Section 4.b).

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.E TABLE OF CONTENTS t

. Page . Summary.of Facility Activities ........................ 1 1 . Followup on Previous Inspection Findings . . . . . . . . . . . . . . 2 Violations, Unresolved Items, Inspector Follow Items Routine Periodic Inspections .......................... 5 Pla.it Tour Observations ~ Plant Maintenance and Outage Activities

^ Surv'eillance Testing _ Radiation Protection.and Chemistry Review of Plant-Events ................................ 16 Failed Diesel Generator Differential and Lockout-Relays Inadvertent Engineered Safety Features Actuation Control Rod Drive Hydraulic Control Unit Seismic Qualification Selective Followup.of Licensee Corrective Action

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Regarding Management Meeting 86-41 ................ ... 19 ' Review of . Licensee Event Reports (LERs) . . . . . . . . . . . . . . . 20 Management Meetings ................................... 21 l Attachment I - Persons Contacted  :

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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 modification Fuel was removed from the reactor in February 1987 to support these modification The reactor remained defueled throughout the period. Major ongoing work included diesel generator, motor control center, valve motor operator and main turbine overhauls. Repair of a reactor building closed cooling water heat exchanger and hydrostatic testing of the salt service water piping were also performed during the perio On April 8, 1987, the NRC: Region I issued the Systematic Assessment of Licensee Performance (SALP) Board Report (No. 50-293/86-98) containing the SALP Board evaluation of the performance of activities at the Pilgrim Nuclear Power Station for the period of November 1, 1985 through January 31, 1987. The SALP Board report identified significant recurring program weaknesses in some functional areas. Improvements, such as in the area of emergency preparedness, were also noted. However, the SALP Report indicated the rate of such change was slow during most of the assessment perio On April 15, 1987, the licensee announced that Mr. A. L. Oxsen, Vice President of Nuclear Operations will leave Boston Edison on June 1,198 The licensee does not plan to name the successor in the immediate futur Currently, the Plant Manager and the group leaders of site security, fire protection and emergency preparedness are reporting directly to Mr. R. G. Bird, Senior Vice President - Nuclea NRC inspection activities during the ' period included: 1) review of licen-see's radiation and contamination control programs during April 7-10, 1987 by a NRC specialist inspector, 2) review of licensee's QA/QC effectiveness during April 27 - May 1,1987 by two NRC region-based inspectors, 3) re-view of licensee's radiation and contamination control programs during April 27 - May 1, 1987 by two NRC specialist inspectors, 4) review of licensee's . fire protection program during May 4 - May 8,1987 by an NRC specialist inspector, and 5) Appendix R fire protection team inspection during May 11 - May 15. 1987. In addition, the senior resident inspectors

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from the Susquehanna facility and the Calvert Cliffs facility assisted the resident inspectors during the week of April 13, 1987. The senior resi-

, dent inspector from the Calvert Cliffs facility was also onsite during the week of April 20, 1987. NRC Region I Administrator William Russell toured the Station on May 7, 1987.

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2.0 Followup on Previous Inspection Findings

Violations

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(Closed) Violation (85-03-01), failure to implement prompt corrective actions in response to QA surveillance findings. The inspector identified numerous QA surveillance findings to which station and corporate manage-ment had not responded. In reply to this notice of violation, the licen-see eliminated QA. surveillance finding sheets as a method of documenting deficiencies. All existing surveillance findings and all new QA findings were documented as QA deficiency reports (DR). The Boston Edison Quality Assurance Manual (BEQAM) and other appropriate procedures were revised to reflect this practic The objective of this revision was to increase senior management involvement in resolution of QA Findings. NRC inspec-tion report 50-293/86-14 identified that licensee management response to QA DRs was also unacceptably slow, and that this appeared to be a recur-ring problem. Based on the termination of the use of QA surveillance findings and the existence of violation 86-14-07 this item is close Unresolved Items (Closed) Unresolved Item (85-03-02), review safety evaluation of recircu-lation loop A flow instabilit Following a 1985 recirculation piping replacement outage Pilgrim experienced unexpected fluctuations in indi-cated recirculation loop flow. These flow fluctuations resulted in a one to two percent power increase. General Electric (GE) performed a safety analysis specifically for the Pilgrim plant. This analysis concludes that the phenomenon is not a safety concern for this plant. -The recirculation flow fluctuations at Pilgrim and other jet pump BWRs and the referenced GE safety analysis are discussed in IE Information Notice 86-11 The inspector had no further questions. This item is close (Closed) Unresolved Item (85-31-01), review the root cause for drifting main steam line high flow switch setpoints. This item was last updated in inspection report 50-293/85-36. During 1985, the licensee experienced significant drifting of the Barton Model 2E8A differential pressure switches used to sense main steam line high flow conditions. The cause of the drifting was found to be improper alignment of the switch linkages during previous modification The licensee subsequently realigned the linkages and initiated more frequent surveillanc During the current outage, these mechanical switches are being replaced with Rosemount elec-tronic pressure transmitters and trip units. This replacement eliminates the mechanical linkage type switch and lengthens the calibration interval from quarterly to each refueling outage. This item is close . - . - . _ - _ - _ - _ . - _ - . - . - - - _ , .-_ __ .. - _ _ _ -

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(Update) Unresolved Item (86-34-01), evaluate the root cause and licensee corrective actions on salt service -water pipe corrosio This item is discussed in detail in paragraph 3.b of this repor (Update) Unresolved Item (86-40-03), evaluate the root cause and signifi-cance of a HFA relay failure. On December 16, 1986, a HFA relay associ-ated with the D refueling floor ventilation radiation monitor failed to properly operate when deenergize During the current inspection period, the inspector reviewed General Electric Service Advice Letter (SAL) 188.1. }

This SAL describes the potential for the failure of certain HFA relays to

. provide correct contact operation when deenergized. SAL 188.1 recommends performance of a test to determine if the problem exists. The licensee stated that a procedure to perform the test had been written and would be implemented for all affected relays during this outage. This item is up-dated to provide for followup of the. prescribed testin Inspector Follow Items (Closed) Inspector Follow Item (84-26-01), licensee to develop instruc-tions containing guidance on breaker use and bus transfer. This item was last - updated in inspection report 50-293/85-14. On September 28, 1984, while shutdown, an improper attempt to crosstie safety-related and non-safety-related 480 VAC electrical busses resulted in the generation of an unplanned reactor scram signal. In response to this incident, the licen-see committed to develop a procedure controlling the transfer and crosstie of 480 VAC load centers. Procedure 3.M.3-35, Dead Bus and Live Bus Trans-fer of 480V Load Centers, was issued and appears to provide the detailed instructions necessary to perform these activitie It also contains appropriate operational restrictions on the intertie of safety and non-safety busses. The procedure was recently utilized to perform a series of bus transfers needed to support ongoing maintenance. The inspector re-viewed these activities to ensure that the procedure had been properly performed. The inspector had no further questions. This item is close (Closed) Inspector Follow Item (85-11-01), review modifications to reactor vessel level instruments described in LER 84-19 and the response to Generic Letter 84-23. Generic Letter 84-23 required implementation of improvements that will reduce level indication errors caused by high dry-well temperatures. BWR Owner's Group studies concluded that to minimize errors, the vertical drop of both the reference and variable legs within the drywell should be minimize The maximum recommended vertical drop inside - the drywell was six fee Plant Design Change (PDC) 85-07 was issued to implement these change PDC 85-07 requires removal of the Yarway heated reference columns and installation of new cold reference l

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legs outside the drywell. This will result in a maximum reference leg vertical drop inside the drywell of two feet. The wide range variable leg vertical drop inside the drywell will also be reduced. Implementation of PDC 85-07 is in progress and will be completed prior to startup from the current outage. This design change should correct the previously observed level instrument error. This item is close (Update) Inspector Follow Item (86-06-13), review implementation of PM program for 480 VAC molded case breakers. In response to NRC inspection 50-293/86-06, the licensee committed to expand and strengthen the preven-tive maintenance (FM) program for 480 VAC molded case circuit breaker The station Maintenance Section issued Engineering Service Request (ESR)86-097 requesting establishment of breaker test acceptance criteria and replacement breaker procurement specifications. The licensee Nuclear Engineering Department (NED), in response to ESR 86-097, supplied appro-priate procurement specifications and breaker test acceptance criteria based on the ongoing coordination study and manufacturer's performance information. The recommended test criteria were subsequently incorporated into procedure 8.Q.3-3, 480 VAC Motor Control Center Testing and Mainten-anc This procedure includes steps for removal and inspection of the breaker cubicle, electrical testing of the breaker, contactors and over-load relays, and cubicle reinstallation and post work testing. Procedure 8.Q.3-3 is currently being utilized to perform testing of safety related breaker The licensee stated that all safety related motor control center breakers would be tested during this outage. The scope and fre-quency of future preventive maintenance and testing of these breakers has not yet been established. The licensee stated that a periodic maintenance program was under developmen This item remains open pending formaliza-tion of this progra (Closed) Inspector Follow Item (86-14-06), review licensee evaluation of loose wiring. This item was last updated in inspection report 50-293/

86-34. The licensee completed a review of station records spanning ap-proximately seven years, and identified seventeen cases of loose or failed terminations. These failures were classified as loose connections, loose fuse clips, improper crimps and loose compression type connections. The latter two failure types are the result of poor practices utilized during initial construction. Similar cases of poor initial construction workman-ship have recently been identified by QC. The licensee appears to be taking appropriate action to address these instance During the licen-see's followup to the spurious primary containment isolations experienced in April, 1986, a detailed walkdown of PCIS and RPS wiring was performed to identify any additional problems. The licensee also issued temporary procedure TP 87-65, Terminal Board Checks of Selected Panels in the Con-trol Room and Cable Spreading Room. This procedure contains instructions for conduct of inspection and torqueing of all control room and cable spreading room connection Inspection findings and quality control

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S n observations of this work will be ' documented on data sheets for revie . Thsse inspections have begun and will be completed during this outage. To

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address concerns regarding loose fuse clips, the licensee is reviewing all

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-surveillance test procedures to identify fuses which are routinely pulle An inspection' of these fuses will then be conducted to identify any de-gradation. The steps taken by the' licensee in conjunction with continuing

'4 , QC efforts and required surveillance testing appear to be adequate assur-

, ance that any remaining problems will be identified and corrected. This

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3 item is c'lose .0 Rout'ine'?eriodic Inspections The inspectors routinely toured the facility during normal and backshift

, hours ta assess general plant and equipment conditions, housekeeping, and acherence to fire protection, security and radiological control. measure .. 'Tnspections were conducted between midnight and six a.m. on May' 5, 1987

.for one hour and weekends on May 3,1987 for two hours and May I,1987 for one hour. Ongoing work activities were monitored ~ te verify that they

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were being conducted iri accordance with approved administrative and tech-nical- 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 consis-tent with system operability requirements and operating procedure '

i During tours of.the control room, the inspectors verified proper stliffing,

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access control and operator attentivenes Adherence to procedures and limit 1no conditions for operations were evaluated. The inspectors. examined equipment lineup and operability, instrument traces and status of control

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room arinunciators. Various control room logs and other available. l'icensee

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documentation.were reviewe '

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q The inspector observed and reviewed outage, maintenance and problem inves-

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tigation~ activities to veri fy compliance with regulations, procedures, t codes and stanaard Involvement of QA/QC, sa fety tag use, personnel

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, qualifications, fire protection precautions, retest requirementsf, and

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, 'The inspector observed tests to verify test performance in accorda'nce with approved procedures and LCO's, collection of valid test results, control-

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led removal and restoration of equipment, and deficiency review and 1' $ resolutio Radiclogical controls were observed on a routine basis during the report-ing' period. Standard industry radiological work practices, conformance to radiological control procedures and 10 CFR Part 20 requirements were ob-served. lndependent surveys of radiological boundaries and random surveys of nonridiological points throughout the' facility we're taken by the-inspecto ,

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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 identif1 cation, access control, badging, and compensatory measures when require Plant Tour Observations NRC Management Tour On May 7, 1987, the NRC Region I Administrator William Russell and ot.hers toured the reactor building,-turbine building, screenhouse and the control room. During the tour, it was noted that scaffolding on 23' elevation in the reactor building contained wood with no NCX merkings nor fire retardant paint on them. The inspector reviewed the fire protection inspector's shift log for May 7,1987 and found several entries for unpainted wood identified in the reactor build-ing. The inspector discussed this observation with the station Fire Protection Group Leader and determined that the control of wood in the process buildings is primarily through procedural direction and the fire protection inspector's routine tours. The licensee. stated that the scaffolding material in question had been cut from fire retardent wood, but because of the small dimensions involved, the NCX markings were not captured. The additional untreated wood was the result of equipment packaging and transport material generated by ongoing maintenance and was subsequently removed from the area. The inspectors will continue to evaluate this area during future routine tour During the tour through the High Pressure Core Injection (HPCI) pump sump room, the inspector found a few equipment tags on the floor, apparently detached from their originally installed position. The inspector discussed this observation with the Operations Section Manager and determined that the licensee had planned to inspect and restore all equipment tags prior to the restar The inspector also observed during the tour that water was on the floor near the south stairway on elevation 91' in the reactor build-in The inspector informed the health physics technician in the area. The health physics technician surveyed the water on the floor and no detectable contamination level was measured. The inspector also noted that the water was coming from the condensation of over-head pipings. The health physics technician contacted maintenance and dried the water in the area.

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Housekeeping The inspector noted the housekeeping and cleanliness of the plant to be generally good. There is a major ongoing effort to paint struc-tures and . components and it is making a significant improvement in -

the appearance of the plant. The licensee is making progress in this are Additional Observations

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During the inspection period, plant operators discussed the following concerns with the inspector:

(1) The Watch Engineers were not always told what maintenance activities were ongoing in the plant and sometimes had dif-ficulty determining which -maintenance supervisors were on shift and available to handle routine problem Licensee management issued a memo during the previous month which required that representatives from all work groups partici-pate in shift turnovers in the control room. The purpose of this policy was to ensure that the operations staff was adequately briefed on station activities. Although attend-ance at the meetings was initially good, participation by station work groups had fallen off in recent week (2) Tagging requests for equipment isolations were not always l dated and occasionally did not provide enough information

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to identify the appropriate system or component. As a re-l sult, the operations staff either did not know when these l tagouts needed to be processed or could not readily deter-mine isolation location. This hampered work planning for the operators. The inspector reviewed examples of tagout requests which did not specify due dates or did not contain l enough descriptive information to identify the system and agreed with the operator's concer Subsequently, the inspector discussed the problems with the Operations Section Manager and the Station Manager. As correc-tive action, the memo on shift turnover attendance was reissued and the licensee indicated that the tagging problems would be addressed. The inspector had no further questions at this time and will review the adequacy of shift turnovers and tagging requests during future routine inspection During a tour of the main stack elevation 65'-6" area on April 21, 1987, the inspector noted that the self frisking device (Model RM-14) located near the step off pad did not have

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a probe. The area was posted radiation / contaminated area and 1 - - .

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required- frisking of hands and feet prior to leaving the are Subsequently, the inspector reviewed the routine radiation sur-vey for the area dated April 3, April 7, and April 14,-1987, and noted that the contamination level was below 1000 dpm/cm2 . The inspector discussed this observation with the licensee Chief .

Radiological Engineer and determined that the licensee was in the process ~ of replacing the probe. During a routine RM-14 electronic and source check surveillance on April 21, 1987, the licensee discovered the faulty probe and subsequently replaced i The inspector reviewed the RM-14 electronic and source check surveillance log sheets dated April 21, April 24 and April 28, 1987. No discrepancies were note On May 12, 1987 during a tour of the emergency diesel generator building, the inspector noted two fi f ty-five gallon drums of diesel lube oil stored upright outdoors, exposed to the weathe The instructions on the drum specifically recommended storage indoors to preclude oil contaminatio No standing water was noted on or near the drums at the time of the observation. The licensee verified the integrity of the container seals and sub-sequently removed the oil to a more suitable storage are Maintenance Diesel Generator Control Panel Rewiring and Inspection Activities Nonconformance Reports (NCR)87-121 and 87-126 were_ issued on March 18, 1987, following a licensee QC inspection finding regarding loose connections and unlugged wires in the 'B' diesel generator con-trol panel The licensee engineering department's disposition of NCRs87-121 and 87-126 required all unlugged wires in the diesel generator control panels to be lugged and terminated. Maintenance

. Requests (MR) 87-46-149, 87-46-150, and 87-46-151 were generated to perform lugging and proper termination of the wires in each diesel generator control panel (C102, C104A and C1048).

A calibration test of the diesel generator instrumentation was per-formed following the relugging and termination of the wires in the control panel. During this test, an error was discovered when low jacket water temperature alarm failed to illuminate as the signal was injecte Upon investigation, the licensee found discrepancies on certain wire terminations against the drawings. The error was at-tributed to random personnel error during relugging and retermination of wires in the control panel The inspector reviewed the completed procedure packages for relugging and wire termination in the diesel generator control panels to det,m mine that lifting and landing of the leads in the control panels were controlled properl The following completed procedures were reviewed:

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3.M.3-17-4 - Crimping of terminal lugs and splices, Revision 0,

} dated February 19, 1987, performed on April 13, 198 TP 87-83 - Control Panel Wiring Inspection, Revision 1, dated April 15, 1987, performed on April 18, 198 I The inspector noted that an independent verification and a QC verifi-

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' cation was done for lifting and landing each wire in the control panels. Additional evaluation of the licensee's actions in this area i are contained in NRC specialist inspection 50-293/87-2 :

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test of the diesel generator instrumentation, 8.E.38 - Diesel Instru-mentation Calibration, Revision 10, dated March 9, 1987 and performed on April 22, 1987. The inspector noted that 25 changes were made to

,. the working copy of the procedure during the performance of the test

{ without proper authorization as required by the station procedure 1.3.4. The inspector discussed this concern with the Chief Mainten-ance Engineer and I&C supervisors. The licensee informed tha inspec-tor that the licensee had planned to perform the procedure with the proper revision prior to the fuel load. The licensee also counseled the individuals who performed the test on the station requirements for proper procedure change. The inspector had no further question Salt Service Water pipe Corrosion On October 14, 1986, the licensee identified a leak from a one-eighth inch diameter hole in the Station Salt Service Water (SSW) system piping at the outlet of the 'A' Reactor Building Closed Cooling Water (RBCCW) heat exchanger. Quality Control (QC) inspections found pipe wall thinning in the area surrounding the hole, and evaluated the damage as localized corrosio The anti-corrosion rubber coating applied to the inside diameter of the carbon steel piping had been torn away resulting 'an accelerated corrosion attack in the area of the leak. Further QC inspections identified other areas in the pip-ing system internal surfaces and piping flanges were significant corrosion damage had occurre The SSW piping corrosion was identified as a startup issue in NRC Management Meeting Report 50-293/86-41 dated December 31, 198 The licensee submitted a response on March 1, 1987 (BEC0 letter 87-039)

to discuss their corrective actions related to the material condition of mechanical components in the SSW system and integrity of equipment in the screenhous The licensee has established an inspection program to evaluate the scope of the problem and identified the root cause of the corrosion as the original material selection and corrosive salt water environ-men Eddy current testing of the salt water heat exchangers was

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performed during the 1984 outage and is being reperformed during this refueling outage. Repairs and tube plugging were required to restore the heat exchange The licensee's response also stated that UT examination of 100 per-cent of the safety-related piping above ground had been completed with no problems found. Inspector review of the current status of .

the testing determined that the UT examinations were not complete, and that several problem areas were identified. Discussions with NED and the Systems Group Leader verified that approximately 60 percent of the above ground piping had been tested, with two areas of concern identified in the 'B' loop which were repaired by replacement of the applicable spool pieces. One spool piece was also replaced on the

'A' loop. A subsequent licensee response to report 50-293/86-41 con-tained an update status report on corrosion in the system and cor-rected the initial respons In order to determine the integrity of the underground SSW piping, the licensee performed the system leakage test at 110 psig on April 16, and April 19, 1987. This test determined that there was an undetermined leakage source of approximately seven gallons per minute within the test boundar Subsequently, a visual remote camera inspection of SSW 'B' loop underground piping was performed on April 23 and April 24, 1987. Preliminary review of high resolution video tapes indicated some lining delamination and galvanic attack of two areas on the screenhouse end of the underground piping. The licensee also performed a hydrostatic pressure test of the 'A' loop piping with no unidentified leakage observed. A visual remote camera inspection of the 'A' loop piping is also planne The data from all of the piping examinations will be utilized later to determine repair plans and inspection criteria for further routine inspection Other corrective actions planned by the licensee in-clude an improved painting system for exterior surfaces and'an evalu-ation to identify improved materials for SSW syste The inspector will follow up on the licensee's continued piping inspections and implementation of the necessary permanent repairs under existing open item 50-293/86-34-0 Salt Service Water /RBCCW Heat Exchanger Repairs An Eddy current test examination performed in March,1987, on the 'B'

loop Reactor Building Closed Cooling Water (RBCCW) Heat Exchanger (E-2098) revealed 56 leaking tubes. In addition, visual inspections during disassembly determined that two of the tubeside partition plates and the adjacent grooved area of the tube sheet were damage ,

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l The wastage found on the tubeside partition plates and tubesheet was caused by inadequate installation of the channel during previous maintenance activities. The channel was not fit-up_ properly, due to problems with the stiffener clearances, which caused inadequate com-pression of the tube sheet gasket. Due to the excessive clearances, the partition plates were eroded approximately one-half inch from the bottom, allowing some bypass flo Repairs to the heat exchanger were completed by plugging the 56 tubes, rebuilding the tube sheet with Belzona, and performing a weld buildup repair to the tubeside partition plates. The stiffeners were modified to allow the proper clearances, and blue checks were per-formed to verify adequate fitup. The inspector reviewed the asso-ciated Maintenance Requests, Safety Evaluations, Work Repair Plans, and engineering calculations, and observed portions of the repair The inspector reviewed Safety Evaluation (SE) No. 2090, dated March 17,~1987, which evaluated the proposed repairs to the heat exchangers. The SE concluded that the tube plugging and Belzona re-pairs were acceptable to restore the heat exchanger's ability to per-form a heat removal function. With the additional 56 tubes plugged, the total number of tubes plugged reached 80. Calculation M-290, dated March 18, 1987, determined that 362 out of a total of 2160 tubes could be plugged without affecting the design heat removal capability, therefore the tube plugging was found acceptabl Although the safety evaluation states that the partition plates would be repaired with Belzona, only the tube sheet was actually repaire The partition plates were repaired by weld buildup to their original configuration. The Belzona epoxy repair was determined to be accept-able because the Belzona would be applied in areas which were not pressure retaining boundaries or heat transfer surfaces of the sys-tem. The safety evaluation also concluded that the heat exchanger should be inspected for structural integrity and leaking tubes during the next refueling outag Following completion of repairs and testing to the 'B' RBCCW loop, the licensee plans to perform inspections on the 'A' heat exchange The results of the 'A' heat exchanger inspection will be reviewed in subsequent routine inspection activitie RHR Pump Wear Ring Examination The inspector reviewed the General Electric (GE) report F&PMT-87-178-00 dated January 23, 1987, covering wear ring samples removed from two Pilgrim RHR pump impellers. The pumps were manufactured by Bingham-Willamett The wear rings were fabricated for SA-182-F6

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type 410, a hardenable martensitic stainless stee The wear rings had been removed for examination as a result of failures experienced by Peach Bottom and Browns Ferry plants. The GE report concluded that microscopic cracking found in one ring was attributed to inter-granular stress corrosion cracking (IGSCC). Unusually high hardness levels for the stainless steel wear rings resulted in. an increased sensitivity to IGSCC. This condition is similar to that found in RHR pump wear rings at other site Based on review of the photomicrographs in the report and the hard-ness values the inspector agreed with GEs conclusion. The inspector confirmed in conversation with the licensee's cognizant engineer, that the four RHR pumps were installed with new 410 SS wear rings, heat treated to a lower hardness (RC 28-32). The inspector also ob-tained a copy of the GE heat treatment procedure EQDE-48-0786 and found it technically acceptable. The process however could have been better controlled if the procedure had specified a tempering range instead of stating tnat the material was required to be " tempered at a temperature to meet the hardness requirements". By specifying a specific tempering range the likelihood of developing an undesirable hardness is significantly reduced and avoids the necessity to retem-per or reheat-treat (austenitize and temper) as determined by the hardness level. A final hardness check, however does provide assur-ance that the material exhibits the proper hardnes Surveillance Testing Diesel Generator Post-Maintenance Testing The inspector observed portions of post maintenance testing activ-ities associated with Maintenance Request (MR) 87-61-15, Preventive Maintenance on B diesel generator, performed on May 1, 1987 to deter-mine that the testing was conducted in accordance with approved pro-cedures, regulatory guides, Technical Specifications, and applicable industry codes and standard The vendor recommended post maintenance testing included motor cur-rent checks, compressor capacity checks, individual starting air motor performance checks, air motor start solenoid checks, starting air bank capacity test, cylinder head torque verifications and radia-tion fan vibration analysi The post maintenance testing on the 'B' diesel generator was per-formed in accordance with procedure No. 3.M.4-36, Attachment R, Post Work Testing / Diesel Generator Maintenance. The inspector noted that the as-run copy of the test was revision 5, dated November 28, 198 The procedure revision obtained by the inspector on April 29, 1987 from the licensee's document control center was revision 7, dated April 22, 198 The inspector discussed this concern with the test engineer and compared the applicable portions of the revision 5 against revision No apparent discrepancies were noted. However,

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the inspector expressed a concern to the licensee management that the similar. problems involving failure to use a current revision of a test procedure were identified in the inspection reports 50-293/

85-0 The licensee management agreed with the inspector - that more effective corrective actions are necessary to prevent recurrenc _The inspector will review licensee action taken in this are .

On May 2 and May 3,1987, the licensee performed 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> endurance testing of the 'B' diesel generator as suggested in the NRC Regula-tory Guide 1.108, Periodic Testing of Diesel Generator Units. The licensee is not committed to the Reg Guide 1.108, however, licensee management decided that an endurance testing of the diesel generator would ba prudent following the extensive overhaul of the diesel generato The test demonstrated full-load-carrying capability of the diesel. generator for a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> while maintaining re-quired voltage and frequency. The test was performed in accordance with procedure No. 8.9.1 - Manually Start and Load Diesel Generator Once Per Mont The inspector will review the test data with the licensee system's engineer in a subsequent inspectio d. Radiation Protection and Chemistry

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On April 16, 1987, a small radioactive particle was detected on a worker's ankle by a whole body monitor near the drywell inside the reactor building. The worker had previously installed new piping in a posted contamination area above the drywell guard house on the 23- ft. elevation of the reactor buildin The particle was subsequently removed and analyzed by gamma spectro-scopy. The gamma detector was calibrated using a spec source of known activity. The particle was found to have 0.544 micro Ci of C0-60 with an associated skin dose rate of 2.58 rems /hr. The licensee subsequently assigned an. extremity dose of 5.16 rems to the worker from the particle. The worker's total quarterly skin of the extremity dose was 5.2 rems, which was less than the NRC dose limit of 18.75 rems / quarter. A whole body count indicated that the worker had no detectable internal contaminatio No additional radioactive particles were identified during sub-sequent tape surveys of the. work area, the reactor building and -

the worker's protective clothing. The work area had relatively low contamination levels,1,000 to 4,000dpm per 100 cm2. The licensee stated that tape surveys were routinely conducted in the plant to check for particle Particles have only been detected in the past when sticky tape was used to remove surface material during the surve Standard smear surveys have not detected the particles, e

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The inspector noted the following weaknesses in the completed Radiological Occurrence Report (ROR) no. 87-4-16/0528, on this inciden (1); Documentation of survey results_in the reactor building was incomplete. The ROR package indicated that tape surveys <

-had been conducted in the reactor building 23 ft. elevation and the work area. However, the documentation did not indicate what portions of these areas had been surveye bsequently, the licensee clarified the survey document (2) Approximate (rather than exact) times were used to evaluate the worker's exposure in the reactor. building. The expos-ure period was decreased from the initial estimate in the ROR evaluation based on the worker's qualitative estimate of the times he entered the reactor butiding and work are However, the ROR indicated a 30-minute uncertainty in the reactor building entry and the shortest (rather than long-est) estimated stay time in the building was used during the evaluation. The worker was not required to sign in on a radiation work permit to enter the work area, but secur-ity records could have been checked to verify his entry time into the reactor building. Subsequently, the licensee confirmed the worker's estimates with security record (3) ~The work sequence was incorrectly stated in the ROR. Spec-ifically, the ROR stated that the worker changed into pro-tective clothing after he entered the reactor buildin Following discussions with the inspector, the licensee determined that the worker had changed outside the reactor buildin In summary, the licensee's evaluation of the particle incident-was poorly documented, considering the historical presence of radioactive particles in the station. At the exit meeting, the inspector emphasized the importance of thoroughly evaluating and documenting these incidents, so that the source of the particles can be identified and acceptable dose estimates mad '

A similar incident involving radioactive specs occurred 'on March 16 and was reviewed during inspections 87-16 and 87-1 The licensee indicated that the following actions had been taken since that time or were planned.

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Approximately 10% of routine contamination surveys in the -

plant will use tape instead of paper smears. Approximately

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50% of job-specific contamination surveys will also use tape smears. These guidelines have been incorporated into Procedures 6.3-110 and SI-RP.300 The actions to be taken for a detected spec have been added to Procedure 6.3-110. These actions included contacting health physics management, immediately stopping work in the area, and obtaining the required authorizations to resume wor The licensee plans to formalize a training program for health physics technicians in June 1987 which will include:

1) traiaing in skin dose assessment, 2) use of adhesives to survey for particles, and 3) use of survey instruments to survey for particle A requirement to routinely survey a sample of laundered protective clothing for particles at the station prior to use has been added to station procedure This particle event and the effectiveness of these measures and of the evaluations of future incidents will be monitored during future NRC inspections (50-293/87-18-02).

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- During onsite excavations for Appendix R modifications several years ago dirt,- asphalt and concrete' containing low levels of contamination were identified. At that time, a fenced in stor-age area outside the protected area on BECo property was create Slightly contaminated dirt, asphalt and concrete resulting from onsite excavations during the past several years has been placed in this area in a large pile. Recent excavations (October 1986)

have produced additional contaminated dirt which has been re-moved from site and placed in the area. The licensee currently estimates that the storage area contains 70,000 cubic feet of material. Samples of the materials were obtained and isotopic analysis performed by the licensee before removal from the sit The level of activity found was reasonably uniform at levels of-10 -6 and 10 -7 micro curies per gram of Cobalt-60 and Cesium-137. The inspector reviewed a sample of soil analyses and ver-ified that results were consistent with the stated levels. These concentrations are well below the allowable release limits es-tablished by 10 CFR Part 20 and do not represent a health or safety concern. In additicn the inspector and a licensee health

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physics technician performed a survey of the storage area using a micro -R meter and found no detectable radiation. During the tour the inspector noted that the gate had been removed to facilitate construction work and that the area had been left unattended. The licensee stated that the gate was replaced and secured nightly and that the entrance to the area would be blocked when left unattended in the future. The inspector had no further question .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 Failed Diesel Generator Differential and Lockout Relays On April 28, 1987, prior to a post maintenance test on 'B' diesel generator, a Westinghouse SA-1 differential relay for the feeder -

breaker 52-609 from the 'B' diesel generator to the 4160V safety bus (A6) failed, energizing a diesel generator lockout relay (GE-HEA).

The relay failure occurred just after the diesel generator output breaker 52-609 had been racked up, prior to a post overhaul diesel generator load testing. Activation of the relays would have tripped the diesel generator and opened the diesel generator output breaker, if the diesel generator had been runnin The function of the differential relay is to provide generator pro-tection against phase-to phase or phase-to ground faults by energiz-ing the lockout relay if such a condition is sensed. When energized, the lockout relay trips open the generator breakers and mechanically locks in the tripped state until manually reset. As the lockout re-lay reaches full locked out position, its relay coil is deenergize The licensee determined that the differential relay failure was caused by a failed silicon-controlled rectifier (SCR) in the rela The failure of the SCR in the differential relay tripped the lockout relay and prevented the lockout relay from being reset. This condi-tion was further aggravated when an operator held the lockout relay reset switch in the reset position, resulting in the burnout of the lockout relay coil. The faulted differential relay was examined by the licensee and it was determined that the relay experienced a ran-dom failure of a SCR. The licensee is planning to ship the relay to Westinghouse for destructive evaluation to confirm this diagnosi NRC IE Information Notice 83-63 addresses a similar failure mode of

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Westinghouse Electric Corporation Type SA-1 differential relays where certain SCRs in the relay can cause a random trip output. The in-spector provided a copy of the Information Notice to the license The inspector and a licensee system's engineer also examined the re-lay and determined that the relay did not contain a S.T. SEMICON SC According to the Information Notice, only some SCRs manufactured by S. T. SEMICON for Westinghouse Electric Corporation Relays since January 10, 1980 have been found to cause the random trip outpu The licensee replaced the faulted relay with a new differential relay of the same type (Westinghouse SA-1). The standard relay calibration has been performed on the new relay and has shown the relay to oper-ate as designed. The failed lockout relay was also replaced. The root cause of the lockout relay failure was attributed to inadequate operator training, resulting in prolonged reset attempts. A night order entry for the operators was made to advise against repeated or prolonged lockout reset attempts. Also, a yellow warning label was put on the lockout relay panel. The licensee management informed the inspector that the Senior Electrical Engineer is planning to review the operating procedures governing lockout relay resetting for en-hancement The equipment history indicates that the licensee experienced a similar problem on April 26, 1986 where the 'A' diesel generator differential and lockout relays have failed. In this case, a faulted coil associated with the differential relay caused an erroneous signal to be generate The lockout relay energized but did not operate to the full locked out position and the coil remained ener-gized. The continuously energized coil failed, resulting in a small fire. The failed differential relay was a General Electric Model 12 CFD relay, which IE Information Notice 85-82 identified as being not seismically qualified. The licensee subsequently replaced the GE 12 CFD relays with Westinghouse SA-1 relays, during this outag b. Inadvertent Engineered Safety Feature Actuation On April 28, 1987, an inadvertent Emergency Core Cooling System (ECCS) initiation signal was generated which resulted in automatic start of the 'A' emergency diesel generato The inadvertent ECCS initiation signal was geaerated when electricians were terminating leads on the analog trip system drywell pressure panels C932 00-3 and C932 0D- The 'B' emergency diesel generator, the standby gas treatment system, and all ECCS systems except 'B' core spray system were tagged out of service for maintenance and thus did not start on the initiation signa __

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O The 'B' core spray pump was operable but the operations staff had i deenergized the DC power to the pump breaker and the discharge valve power supply breakers prior to the pre-operational testing of' the Analog Trip System to prevent inadvertent actuation of the syste Upon investigation, the licensee found that the drywell pressure transmitters (PT-1001-89 A and C) and their associated tubing were isolated and pressurized at about 3 psig. On April 20, 1987, the licensee had performed preoperational pressure testing of the Rose-mount transmitters (PT-1001-89 A and C) using Temporary Procedure (TP-86-52, Revision 0) dated November 6, 1986. The test involved the isolation of the transmitter and pressurizing the transmitter and its associated tubing with nitrogen to 3 psig. Then, a leak test on all joints within the test boundaries was nerforme The procedure required depressurization of the tubing upan completion of the leak detection tes The licensee determined that the individuals wh)+erformed the pres-sure test had left the tubing pressurized at the test pressure of 3 psi As a corrective action, the licensee W iefu' all personnel performing the test on the intent of the procedure 't.c f.he importance of the venting process. On April 30, 1987, the 1.censet reperformed the pressure testing of the transmitters to determine the ac'equacy of the procedur The - inspector witnessed the test and agreed with the test director that the steps in the procedure adequately addressed the venting pro-cess at the end of each pressure test. The inspector suggested to the licensee that an independent verification would reduce the pos-sibility of recurrenc During the outage, the licensee is replacing Yarway temperature com-pensated reactor vessel level instrumentation and Barton pressure switches with the Rosemount model 11530 pressure transmitters along with the analog trip system (ATS). The analog trip system, utilizing the Rosemount pressure transmitters, should increase the reliability and accuracy of the safety related trip systems. The inspector had no further questions at this tim c. Control Rod Drive Hydraulic Control Unit Seismic Qualification On October 22, 1986 the licensee received written notification from General Electric (GE) of improper hydraulic control unit (HCU) in-sta11ation at a BWR 6 plant. The notification stated that the hold down bolts for many HCU frames were either missing or did not appear to be tightened sufficientl GE investigation identified that torque values specified for the bolts differed from the values used

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during seismic qualification testing. The GE letter concluded that missing or loose hold down bolts at the BWR 6 did not constitute a safety problem 'during a seismic event, nor would they at any other BWR. Subsequent to this notification, the licensee initiated a plant specific evaluation of the HCU hold down bolt Loose and missing bolts were found to exist at Pilgrim. Preliminary licensee evalua-tion indicates that seismic qualification of the HCUs may not be sup-ported in this configuration. The HCUs were declared inoperable and 1 the NRC notified via ENS on May 8, 1987. The licensee is conducting a 10 CFR Part 21 evaluation. The completion of this evaluation and repairs proposed will be reviewed during a future inspection (87-18-01).

5.0 Selective Followup of Licensee Corrective Action Regarding Issues identified in Management Meeting 86-41 In Systematic Assessment of Licensee Performance (SALP) report (50-293/

86-99) and in NRC-Boston Edison Company Management Conference Report 86-41, NRC _ identified concerns to the licensee regarding weaknesses in followup to radiological problems. The licensee responded to Management Meetin'g 86-41 specifically identifying corrective actions as follows:

Action Upgrade the Radiological Occurrence Report (ROR) procedure to improve categorization of RORs by severity leve Stated Status: Completed June 198 . Assign a full time ROR Coordinator position (technicians rotate through this position).

Stated Status: Completed January 198 . ROR's to be transmitted and assigned where possible to Section Manager level or abov Stated Status: Completed February 198 . ROR tracking system to be computerize Stated Status: Completed March 198 A review was conducted of the aforementioned corrective actions through inspection of documents and discussions with Radiological Control super-visors, technicians, contractors and the ROR coordinato Discussions with numerous licensed and non-licensed operations department personnel were also conducted in order to corroborate or lend insight to the causes of the identified problem The following information was obtained:

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' Categorizing RORs by Severity Level. This has been incorporated into the - program and RORs . are . now categorized into three prioritie Priority one RORs get acted ' on immediately and generally get a-thorough review. However, category.two and three RORs do not receive such timely action. Currently, there are 87 RORs that remain open from 1986 and approximately 227 remain open from~198 <

2. . Assignment of a full time ROR Coordinator. This has been completed as state .

- RORs to be~ transferred and assigned to section managers where possibl This is .done with all priority one ROR However, . re-sponses to the RORs are occasionally inconsistent, especially when the corrective action involved disciplinary action. It appeared that corrective action 'taken for personnel performance. related HP viola-

tions differs depending upon whether the individual involved is a contractor or a permanent BECo employe . ROR tracking system to be computerize RORs are input into a com-

.pute The computer has the capability of trendin However, com-ptterized trends of the RORs is not generally used or evaluate The. licensee's activities in the Radiological Controls area will be fol-lowed in future inspection .0 Review'of LER's LER's submitted' to NRC:RI were reviewed to verify that the details were I clearly reported, including accuracy of the description of cause and ade-quacy of corrective actio The inspector determined whether further information was required from the licensee, whether generic impitcations were indicated, and whether the event warranted onsite followup. The fol-lowing LER's were reviewed:

LER N Event Date Report Date Subject 86-028-01 12/22/86 05/08/87 Failure to Recognize Effects of Electrical Isolation Resulting in an Unplanned Engineered Safeguard Feature Actua-tion  !

86-027-01 11/19/86 05/07/87 Loss of Offsite Power due to Severe Winter Storm 87-005 3/31/87 4/29/87 Loss of Preferred Off-Site Power during a Storm i

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LER 87-005 documents loss of preferred off-site power event and the re-lated vital bus (86) failure to transfer. The inspector noted that numer-ous occurrences of a loss of preferred offsite power have been reported in the last ten year During the exit interview with the licensee, the inspector discussed whether a reliability study of preferred off-site power is require The inspector will follow up on the licensee's permanent corrective action to address the root cause of vital bus (B6) failure in a future inspec-tio Problems with preplanning for electrical equipment isolations may have led to the failure of bus B6 to reenergize following a loss of off-site powe .0 Management Meetings 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 On April 27, 1987, management representatives of the NRC: Region I appeared before a special Massachusetts Legislative Committee on the Pilgrim Nuclear Power Plant. The committee, formed by a group of state legisla-tors, conducted hearings throughout the month on issues ranging from emergency planning to the health effects of radiation. Boston Edison management was also represente Representatives of state and local governments, civil defense agencies and citizens groups have also testi-fied before the committe On May 1, 1987, the licensee met with NRC management in the regional offices to discuss a recent NRC violation concerning the surveillance progra A management meeting was also held on May 7, 1987 to discuss the 1986 Systematic Assessment of Licensee Performance (SALP) report. Representa-tives from the NRC and Boston Edison attended the meeting, as did local public officials and members of the news medi .o I'

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Attachment I to Inspection Report 50-293/87-18 i

Person- Lontacted R. Bird, Senior Vice President - Nuclear

• K. Roberts, Station Manager D. Swanson, Nuclear Engineering Department Manager N. Brosee, Maintenance Section Head'

T. Sowdon, Radiological Section Head N. Gannon, Chief Radiological Engineer J. Seery, Technical Section Head P. Mastrangelo, Chief Operating Engineer R. Sherry, Chief Maintenance Engineer C. Higgins, Security Group Leader F. Wozniak, Fire Protection Group Leader

• Senior licensee representative present at the exit meeting.