IR 05000245/1989014

Insp Rept 50-245/89-14 on 890616-0717.No Unsafe Conditions Noted.Major Areas Inspected:Plant Operations,Physical Security,Control Room Exercise Equipment & NRC Temporary Instruction 2513/93

ML20248A751
Person / Time
Site:	Millstone
Issue date:	07/28/1989
From:	Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20248A744	List: IR 05000245/1989014 ML20248A740
References
50-245-89-14, GL-88-14, NUDOCS 8908090019
Download: ML20248A751 (11)

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-U.S.' NUCLEAR REGULATORY COMMISSION

' REGION I'

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. INSPECTION SUMMARY =

Report:N .50-245/89-14--

' Docket No'. 50-245 License N DPR-21 O l

Licenseei NortheastNuclearEnerhyCompany Facility: '

Millstone Nuclear Power Station, Unit.1 g ' Inspection At: Waterford, Connecticut

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Dates: June 16 to July 17, 1989

'Inspectorst . Michael Boyle, Project Manager, NRC:NRR Lynn Kolonauski, Resident Inspector, MP-1

' William'Raymond, Senior' Resident Inspector-F Approved byi

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. /gMcCabe, Chief, Re toft Projects Section 4A ' Da'te Inspection Summary: Inspection rom 6/16/89 - 7/17/89 (Report 50-245/89-14)

Areas Inspected: Routine NRC resident inspection of previously identified items, plant operations, physical security, control . room exercise equipment, NRC.. Temporary Instruction 2515/93, temporary repair of the service water header

' to the reactor. building closed cooling water heat exchangers, maintenance and surveillance activities, licensee event ~ reports'and committee activitie The inspection involved 67. inspection bour Four backshift hours, including one deep backshift' hour, were conducte .Results: The inspection identified no unsafe plant conditions. Followup is planned for: (i) the issuance of a 10 CFR part 21 report on the dimensional changes ma'de to ITT Grinnel snubbers (Detail 10.0) and (ii) LER 88-10-01 (De-taii.10.0).

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8908090019 890728 PDR ADOCK 05000245 g PDC

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TABLE OF CONTENTS j PAGE 1.0 Persons Contacted.................................................... I 2.0 S umma ry o f Fa c i l i ty Act i v i ti e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 3.0 Previous Inspection Findings (71707).................................. 1 3.1 (Closed) UNR 88-12-02, "Non-QA Work Order Used for Vital AC Motor Generator Set".......................................... 1 3.2 (0 pen) UNR 89-02-02, " Post Accident Operability of Reactor Building to Torus Vacuum Breaker Butterfly Valves 1-AC-3A and 3B"....................................................... 2 4.0 Facility Tours and Operational Status Reviews (71707)................ 3 4.1 Safety System Operability....................................... 3 4.2 Plant Incident Reports.......................................... 4 5.0 Control. Room Exerci se Equipment (71707) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 6.0 NRC Temporary Instruction (TI) 2515/93, " Diesel Generator Fuel Oil Quality Assurance" (25593).............. ................... ...... 4 7.0 Temporary Repair to Service Water (SW) Header to Reactor Building Closed Cooling Water (RBCCW) Heat Exchangers (71707)............... 5 8.0 Maintenance (62703)...................... ........................... C 9.0 Surveillance (61726).................. ............................... 7

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10. 0 Li cen see Event Report s ( 92700) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 11.0 Plant Operations Review Committee (40500).............. ............. 9 12.0 Management Meetings (30703)......... ............... ............... 9 l

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l DETAILS 1.0 Persons Contacted J. Stetz, Unit 1 Superintendent R. Palmieri, Operations Supervisor P. Prezkop, Instrumentation and Controls Supervisor N. Bergh, Maintenance Supervisor W. Vogel, Engineering Supervisor M. Brennan, Health Physics Supervisor The inspectors also contacted other members of the Operations, Instrumen-tation and Control, Maintenance, Engineering, and Health Physics depart-ment .0 Summary of Facility Activities Millstone 1 was at full power at the start of the inspection period. On June 17 at 8:12 p.m. , the licensee commenced a downpower to 81% in re-sponse to low discharge pressure and motor amperage indications for the

"C" condensate pump. Maintenance personnel removed the pump, inspected its internals, and found only a small amount of foreign material. After pump reinstallation, discharge pressure and motor amperage still indicated low during a retest on June 21. The licensee then postulated that the "C" condensate pump suction valve (1-CN-1C) had failed close On June 28 at 2:58 p.m., the licensee commenced a controlled shutdown in order to drain the hotwell for repair of 1-CN-1 Cold shutdown was achieved on June 29 at 6:32 a.m.. The licensee discovered that the valve stem had broken. Because of the short term unavailability of replacement parts, the licensee chose to remove rather than repair the valve inter-nal Startup commenced on July I at 6:51 a.m., the reactor was critical at 8:19 a.m. , and full power was reached on July 2 at 4:32 p.m. Full power was maintained for the remainder of the inspection except for a short down-power to 80% on July 6 to conduct turbine stop valve testin .0 Status of Previous Inspection Findings 3.1 (Closed) UNR 88-12-02, "Non-QA Work Order Used for Vital AC Motor Generator Set" In 1984, the Production Maintenance Management System (PMMS) errone-ously indicated that the Millstone 1 Vitcl AC motor generator (MG)

set was not under Quality Assurance (QA) program controls. The lic-ensee discovered and corrected this error in June 1987. However, an automated work order (AWO M1 86 11155) written for the VAC MG set prior to the correction was later implemer.ted as non-QA. Millstone 1 Engineering wrote a nonconformance report to address this discrepancy

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(NCR 188-055). The work included replacing an upper sprocket, re-

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working a lower sprocket, and installing a new sprocket belt. The replacement parts were classified as commercial grade items, but dedication was accomplished by a successful retest. In addition, the operability of the VAC MG set was maintained because it has been in continuous service during power operations. The inspector had no further questions. This item is close .2 (0 pen) UNR 89-02-02, " Post Accident Operability of Reactor Building to Torus Vacuum Breaker Butterfly Valves 1-AC-3A and 3B" At Millstone 1, a design basis loss of coolant accident (LOCA) with a loss of Instrument Air as the single failure would leave a single check valve (1-AC-2A or 28) for primary containment isolation because vacuum breakers 1-AC-3A and 3B would fail open. These valves are designed to fail open so the vacuum relief function would be main-tained at the sacrifice of containment isolation. NRC IR 50-245/

89-02 listed five separate inspector concerns related to this issu One of those oncerns is addressed her The inspector questioned the safety / risk benefit of seismically qualifying the air supply to 1-AC-3A and 38. The licensee evaluated this issue in accordance with the Integrated Safety Assessment Pro-gram (ISAP Topic 1.108). Major evaluation assumptions include: Any proposed modification would completely eliminate the leakage potential from the vacuum relief lines without impacting vacuum relief capabilit . The leakage through check valves 1-AC-2A and 2B was assumed to be the average total leakage obtained during the five most re-cent local leak rate tests for the associated primary contain-ment penetratio . For core melt sequences which resulted in containment failure at or before the time of core melt, no benefit would be gained by eliminating leakage through the vacuum relief line . The failure probability of Instrument Air is negligible compared to the failure of the 120 V Instrument AC system and either loss l would cause 1-AC-3A and 3B to fail ope j l In an accident scenario with a loss of normal power (LNP) and l failure of the Gas Turbine Generator, Instrument Air was assumed I to fai j l

The evaluation methodology considered three failure mechanisms: the  !

valve closed but normal leakage occurred; the valve failed to seat; or the internal disc rupture Initiating events leading to core melt / damage were: internal initiating events, fires, and seismic l

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events. .The evaluation resulted in'a maximum total'public exposure savings of 3.95 man-rem over the remaining unit life. _In reality, this savings is not fully ' achievable because assumption . number.1 is i not attainable. Additionally, the evaluation assigned this topic an ISAP score of 181/ year; this-number' expresses the. relative worth of individual-plant modifications from a public safety perspective. In comparison, the torus strainer changeouts (see IR 50-245/89-08) were rated'as 342,000/ year. The results indicate.that this. issue has low-safety significanc As indicated in assumption number 4, it is notaole that the reli-ability of.the 120V Instrument AC. system.is more important than that H of the Instrument Air syste Either loss would cause 1-AC-3A and 3B l to fail. open. Four other inspector concerns related to this issue remain unresolved; these are listed in NRC IR 50-245/89-02. This item remains open pending further NRC review of what licensee actions may.be warranted after. receipt of the licensee's response to NRC Generic Letter 88-14, " Instrument Air Supply System Problems Affect-ing Safety'-Related Equipmnt."

4.0 ' Facility Tours and Operational Status Reviews The inspector reviewed control indications for proper functioning, cor-relation between channels, and conformance with Technical Specifications (TS). -The inspector verified proper control room manning and discussed alarm conditions in effect and alarms received with the operators and found them to be cognizant of plant conditions and indication The in-spector observed prompt and appropriate operator response to of fnormal and changing plant conditions. Shift turnovers were found to be thorough and in conformance with ACP 6.12 " Shift Relief Procedure." Operating logs and Plant Incident Reports (PIRs) were reviewed for accuracy and adherence to station procedures. The inspectors conducted backshift inspections of the control room and found all shift personnel to be alert and attentive to their duties. During plant tours, posting, control, and the'use of personnel monitoring devices for radiation, contamination, and high radi-ation areas were inspected. Plant housekeeping controls were observed, including control of flammable and other hazardous materials. The inspec-tors also verified proper implementation of selected aspects of the sta-tion security program, including site access controls, personnel searches, compensatory measures, adequacy of physical barriers, and guard force re-sponse to alarms and degraded conditions. No unacceptable conditions were identified. The inspectors also addressed the following activitie .1 Safety System Operability Standby emergency systems were inspected to determine system oper-ability and readiness for automatic initiation. The following sys-tems were reviewed: feedwater coolant injection, automatic pressure relie/, low pressure coolant injection, emergency service water, core spray, standby gas treatment, and standby liquid contro The status

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of the control rod drive hydraulic control units, emergency diesel .

generator, gas turbine, station batteries, and isolation condenser- L was also inspected. '.The reviews. considered (as applicable) proper positioning of major flow path valves, operable normal and emergency power sources, proper operation of indications and controls, and pro-

.per cooling and lubrication. References used for the review included the Updated Final-Safety Analysis Report, and system diagrams and operating procedures. The inspectors identified no inadequacie .2 Plant Incident Reports Selected plant incident . reports (PIRs) were reviewed to (i) determine the significance of the events, (ii) review the licensee's evaluation of the events, (iii) verify the licensee's response. and corrective actions, and (iv) verify whether the licensee reported the events in accordance with applicable requirements. The.following PIRs were .

reviewed; significant events are described elsewhere in this report I as referenced: 1-89-14, 1-89-22 and 1-89-25 (Detail 10.0),1-89-37 (Detail 10.0), 1-89-48, 1-89-49 (Detail 8.0), 1-89-51, 1-89-53 to 1-89-5 No inadequacies were identifie .0 Control Room Exercise Equipment The licensee asked whether physical exercise equipment could be used by onshift operators. This matter was being evaluated by the licensee as a potential initiative to improve employee health. The inspector stated that using exercise equipment within the hearing of control room operators (e.g., within the control room) did not appear acceptable because of the l distraction potential. Subsequent NRC review noted that use of exei.'se equipment onshift does not fall within the professional duties of operat-ing personnel and therefore does not appear acceptable for individuals fulfilling NRC watchstanding requirements. For individuals on watch and in excess of NRC watchstanding requirements,.no prohibition against use of exercise equipment onshift was identified, provided that there is no re-L sulting inadequacy in the ability of the shift to perform required safety l functions. The inspector recommended that, if the licensee considered the use of exercise equipment by watchstanders to be worthwhile, a license amendment be sought to specifically authorize and constrain such activit .0 NRC Temporary Instruction (TI) 2515/93, " Quality Assurance for Diesel Generator (DG) Fuel Oil - Multiplant Action Item A-15" TI 2515/93 requires the inspector to verify that the licensee has included DG fuel oil in his quality assurar.ce program. The inspector reviewed Ap-pendix A of the licensee's Quality Assurance Topical Report (Revision II)

and noted the following entry:

"The following systems, structures, and components of a nuclear power plant, including their foundations and supports, are rated as Category The pertinent quality assurunce requirements of 10 CFR 50 Appendix B

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should be-applied, as a minimum,'to all quality activities affecting the

. safety function of these systems, structures, and components as listed ,

'. below and to other items and services specifically identified by Northeast I Utilities'in-each Final Safety Analysis Report addressing Section 3.2.1 of Regulatory Guide 1.70." The inspector verified that emergency diesel generator fuel oil is listed in Appendix A.under.the heading'"Consum-ables',"'and concluded that the licensee meets the-requirements of Multi-plant Action Item A-1 .0 = Temporary Repair to Service Water (SW) Header to Reactor Building Closed ~

Cooling-Water (RBCCW) Heat Exchangers On June 14, maintenance personnel were preparing to paint the SW header to the RBCCW heat exchangers. .They planned to.use a needle gun to remove existing paint, but a through-wall break developed when the gun was ap-plied to the buttweld located immediately upstream of valve 1-SW-107. The break was not isolable without removing the SW system from service. The pipe.'is 24 inch diameter carbon steel Class 3 piping; the inner surface is epoxy-coate The licensee conducted ultrasonic examinations of the en-tire weld and surrounding area. The results indicated that the break was approximately 0.5 by 0.375 inches and was concise as no wall thinning was present'around the break. No additional weak areas were found. The con-ciseness of the flaw lead the licensee to conclude'that the break was caused by a construction defect, rather than erosion or corrosio The licensee initially installed a lap patch using a piece of rubber and a metal ban This patch was replaced with an interim, noncode repair (per ASME Section XI). On June 20, the licensee welded a four by two inch car-bon steel patch to the pipe; a rubber gasket was attached to the inner surface of the patch to limit SW 1eakage during installation. The inspec-tor reviewed the automated work order (AWO) M1 89 07031 and identified no inadequacies. Retesting consisted of a liquid penetrant test and opera-tional pressure test; the licensee chose not to conduct a hydrostatic test per ASME Section XI because the repair was not a permanent Section XI repair. . The licensee will conduct a permanent repair that will restore the piping to its original design configuration (B31.1-1967) during the 1991 refueling outag On June 22, NRC held a conference call with the licensee to discuss this specific noncoda repair and the licensee's program for future noncode re-pair The NRC requested two specific actions from the licensee: first, that he submit a formal written safety evaluation which documents the ac-ceptability of the repair and addresses brittle fracture and SW system operability; and second, that he provide a detailed description of his Class 3 piping repair program. The inspectors will continua to follow the licensee's actions related to future noncode repair ;

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L8 .0 ' Maintenance-The inspectors observed and reviewed selected aspects of the following safety-related maintenance, including procedural adherence, obtaining re-quired ' administrative approvals and tagouts prior to work initiation, pro-per quality assurance and personnel protection measures, and verification:

of proper system restoration and retest prior to.its' return to servic No inadequacies were identifie '

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Temporary Repair to SW Header to RBCCW Heat Exchangers on June 20

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Removal of Valve Wedge from "C" Condensate Suction Valve 1-CN-IC, on-June 30 On June 17 at 8:10 p.m., control room operators noted that the offgas recombiner isolated. While investigating the cause, the operators observed low indications for the "C" condensate pump discharge pres-sure and motor amperag At 8:12 p.m., the licensee commenced a power reduction to 81% to maintain power within' the capacity of two condensate pumps. At full power, Millstone 1 operates with three condensate, three condensate booster, and two feedwater pump Maintenance removed and insr:cted the internals of the "C" condensate pump; a small amount of of foreign material was found. The licensee reinstalled the pump, but it again indicated low motor amperage and discharge pressure during a June 21 retest. The licensee then pos-tulated that the "C" condensate suction valve (1-CN-1C, a'36. inch manual gate valve) had failed closed. A shutdown commenced on June 28 in order to drain the hotwell for investigation of 1-CN-1C in-ternal The licensee discovered that the valve stem had a clean break ap-proximately ten inches from its key fit to the valve wedge. The valve stem was the original and had beer in service for the past nineteen years. The licensee plans to have the stem analyzed for failure mechanism determinatio Because of the short term unavailability of replacement parts, the licensee decided to remove the valve internals rather than rejoin the valve wedge to a new stem. The inspector observed this evolution and reviewed the associated work orders (AW0s M1 89-07399 and 07401) and jumper bypass evaluation (1-89-45). No inadequacies were note CN-1C is normally full open during power operations; its purpose is to isolate the condensate pump for corrective maintenance. In the event of a "C" condensate pump failure during the current cycle, the licensee would either operate at 80% power or shutdown to drain the hotwell because the pump is no longer isolable for repairs. The licensee was confident in the reliability of the "C" condensate pump

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because itLwas rebuilt during the recent refueling ' outage and was inspected again in June, as described above. The inspector had no further question .0 Su,rveillance The inspector observed.and reviewed selected aspects of the following'sur-veillance for conduct in accordance with current approved procedures, for test result compliance with administrative requirements and technical specifications, and for deficiency-correction in accordance with admini-strative requirements. The inspector noted that the surveillance. teams displayed thorough coordination and adherence to procedures. No inade-quacies were identifie SP408J, " Condenser Low Vacuum Scram Functional Test", on July 12 10.0 Licensee Event Reports The following Licensee Event Reports (LERs) were reviewed to assess LER accuracy, the adequacy of' corrective actions, compliance with 10 CFR re-porting requirements and to determine if there were genoric implications or if further information was required. The events are discussed below or elsewhere, as referenced. The inspectors found the events to be accu-rately described in sufficient detai No inadequacies were note LER 86-18-02, " Reactor Protection System (RPS) Initiation from Inter-mediate Range Monitor (IRM) Noise Spike" On May 26, 1986, while Millstone I was shutdown, an RPS actuation occurre due to noise spikes on IRMs 12 and 16. Source Range Monitor (SRM) 21 was being. withdrawn at the time of the scram; the licensee determined that-SRM drive relay chattering caused IRM spiking. The update report de-scribes the licensee's discovery of pitted contacts in the SRM/IRM drive relays. These have been replaced with arc suppression relay LER 89-07-00, " Hydraulic Snubber Failures" In accordance with TS 4.6.I.1, all safety-related mechanical and hydraulic snubbers were visually inspected during the refueling outage. On April 11, the licensee determined that all had passed their visual examinations with the exception of one hydraulic snubber (HSS-85). HSS-85-had no vis-ible fluid in its reservoir; it was manufactured by ITT Grinnell and was located on the "A" SRV blowdown lin Upon disassembly, the licensee discovered cracks in the snubber's plastic j reservoir. The licensee contacted ITT Grinell and learned that dimen- '

sional changes had been made to the end plates and plastic reservoir cylinders on later model snubbers. Since the Millstone I snubbers pre-dated this change, reservoirs purchased for rebuilt snubbers cracked be-cause the tolerance between the raised flange on the snubber end plate and l

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the cylinder inner diameter was less than the minimum design tolerance (0.006 versus 0.015 inches). The licensee replaced all safety-related snubbers susceptible to cracking with new snubbers procured from ITT Grinnel. Because the failure cause for HSS-85 was established and re-

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medied for all ITT Grinnel snubbers, the failure of HSS-85 will not neces-

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sitate a shorter snubber inspection interval per TS Table 4.6. The inspector asked the licensee if ITT Crinnel had issued a 10 CFR Part 21 report to alert other licensees to the snubber failure potential. The licensee was not aware of such a Part 21 report. This issue remains un-resolved pending the issuance of a Part 21 report by either ITT Grinnel or the licensee (UNR 89-14-01).

TS 4.6.1.3 requires a variety of functional tests. One inoperable hy-draulic snubber (HSS-44) was identified in the first ten percent sampl No failures were identified in the additional five percent sample. Two failures (HSS-23 and HSS-27) were identified among those snubbers tested  !

because they failed functional testing during the previous refueling out- '

age. All failed snubbers were rebuilt and satisfactorily reteste In reviewing LER 89-07, the inspector noted that the licensee had not up-dated LER 88-10, which inaccurately describes personnel error as the root cause for a failed snubber discovered in November 1988.Section V of LER 89-07 describes the actual failure mechanism which involves the dimen-sional inadequacies described above. The inspector will review the up-dated LER upon issuance (UNR 89-14-02).

LER 89-11-00, " Safety Relief Valve Setpoint Drift" On May 11 during the Cycle 12 refueling outage, Wyle Laboratories provided the licensee with bench test results for the six Main Steam Safety Relief Valves (SRVs) which had been in service during Cycle 1 The resuits in-dicated that four of the six SRVs failed to actuate within the tolerance (+/- 1%) allowed by TS 4.6.E.1. The SRVs are two-stage Target Rock model 7567F. The results are listed below:

Pilot Assembly Desired As Found Percent Above Serial Number Setpoint Setpoint TS Limit 1036 1095 psig 1290 psig 16.8%

'038 1110 1124 0.2%

1041 1125 1138 0.1%

1077 1125 1177 3.6%

The licensee determined that the reactor coolant pressure boundary safety limit (1375 psig) would not have been exceeded during a Cycle 12 transient by applying General Electric (GE) analysis EAS-84-0787, which was de-veloped in July 1987 for Cycle 11. The analysis concluded that reactor pressure would not exceed 1375 psig during a limiting transient if one SRV failed to open and the remaining valves opened at pressures five percent i

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above their desired setpoints. In comparison, the failures postulated by the study exceed the actual failures listed above. Although GE performed EAS-84-0787 for Cycle 11, it can be applied to Cycle 12 because the Cycle 12 overpressurization event analysis resulted in a peak pressure less than the same limiting analysis performed for Cycle 11 (1268 psig versus 1270 psi 9).

The SRVs were rebuilt, set to their desired setpoints, and were satisfac-torily retested. The BWR Owners Group Subcommittee on SRV Setpoint Drift postulates that mild oxidic bonding between the pilot disc and seat causes setpoint drift. Testing conducted under subcommittee direction found that alternate pilot disc materials are expected to resolve the bonding prob-le The licensee installed the alternate pilot disc materials in one valve (S/N 1035) during the 1989 outag The licensee will evaluate modi-fication of the remaining valves if satisfactory bench test results are obtained during the next refueling outage, which is scheduled in Winter 1991. The inspector had no further question LER 89-12-01, " Loss of Normal Power (LNP) While Switching Reserve Station Service Transformer" This event was reviewed in IR 50-245/89-08, De. ail The licensee dis-covered an error in the LER description of the unit's electrical con-figuration at the time of the LNP. The update report corrects this erro .

LER 89-14-00, " Required Shutdown Due to "A" Recirculation Pump Seal Failure" (IR 50-245/89-12 Detail 4.4)

LER 89-15-00, " Condenser Low Vacuum Scram" (IR 50-245/89-12 Detail 4.5)

11.0 Plant Operations Review Committee The inspector attended several Plant Operations Review Committee (PORC)

meetings during the period and verified that Technical Specification 6. requirements for committee quorum were met. The meeting agenda included reviews of plant design changes, licensee event reports, procedure revi-sions, and new procedures. The inspector noted that committee performance was in compliance with TS 6.5.1 and and that frank discussions and probing questions were encouraged. No inadequacies were identifie .0 Management Meetings The inspectors held periodic meetings with station management to discuss findings during the period. A summary of findings was also discussed at the conclusion of the inspectio No proprietary information was covered during the inspection. The inspectors provided no written material to the license i

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