IR 05000440/1987003

Insp Rept 50-440/87-03 on 870117-0303.Violations Noted: Failure to Establish Documented Safety Evaluations & Failure to Provide NRC Notification of RCIC Sys Inoperability within Time Limits Specified in 10CFR50.72

ML20206A174
Person / Time
Site:	Perry
Issue date:	04/01/1987
From:	Knop R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20206A137	List: IR 05000440/1987003 ML20206A133 ML20206A146
References
50-440-87-03, 50-440-87-3, IEIN-87-008, IEIN-87-8, NUDOCS 8704070439
Download: ML20206A174 (19)

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, U.S. NUCLEAR REGULATORY COMMISSION

REGION III

Report No. 50-440/87003(DRP)

Docket No. 50-440 License No. NPF-58 Licensee: Cleveland Electric Illuminating Company Post Office Box 5000 Cleveland, OH 44101 Facility Name: Perry Nuclear Power Plant, Unit 1 Inspection At: Perry Site, Perry, OH Inspection Conducted: January 17 through March 3,1987 Inspectors: K. A. Connaughton G. F. O'Dwyer P. R. Wohld di'C dby Approved By: R. C. Knop, Chief b Reactor Projects Section IB Date Inspection Summary Inspection on January 17 through March 3, 1987 (Report No. 50-440/87003(DRP))

Areas Inspected:

Routine unannounced inspection by resident inspectors of previous inspection items, IE Information Notice 87-08, operational safety, onsite followup of events at operating reactors, engineered safety feature (ESF) walkdown, Licensee Event Reports, monthly surveillance observation, monthly maintenance observation, startup test witnessing, and onsite review committee activitie Results: Of the ten areas inspected, two violations were identified in one ,

area (failure to establish documented safety evaluations required by 10 CFR 50.59-Paragraph 4c) and (failure to provide NRC notification of RCIC system inoperability within time limits specified in 10 CFR 50.72-Paragraph 4d).

Plant status meetings were conducted between licensee and NRC Region III management personnel on January 21, and February 26, 1987. At the close of this inspection period the facility was preparing for a reactor startup and resumption of Test Condition 1 startup testing,

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, DETAILS Persons Contacted A. Kaplan, Vice President, Nuclear Operations Division

• 1. C. M. Shuster, Manager, Nuclear Engineering Department (NED)

• D. R. Green, General Supervising Engineer, EDS (NED)

• D. J. Nealon, Electrical Engineer (NED)

• 1. M. D. Lyster, Manager, Perry Plant Operations Department (PP0D)

• R. A. Stratman, General Supervising Engineer, Operations Section, (PP0D)

R. P. Jadgchew, General Supervising Engineer, Instrumentation and Controls Section (PPOD)

G. R. Anderson, Instrumentation and Controls, (PP0D)

1. L. R. Teichman, Maintenance Planning, (PP0D)

• D. J. Takacs, General Supervisor, Maintenance (PP00)

1. M. W. Gmyrek, Senior Operations Coordinator (PPOD)

• D. R. Workman, NOV Field Engineer (PP0D)

• R. L. Scherman, Lead Supervisor (PP0D)

1. G. Chasko, Operations Engineer (PP0D)

• G. Karayianopoulos, M0V Field Engineer (PP0D)

1. D. J. Meinke, Supervisor, Lead I&C (PPOD)

F. R. Stead, Manager, Perry Plant Technical Department (PPTD)

1. E. M. Buzzelli, General Supervising Engineer (PPTD)

5. F. Kensicki, Technical Superintendent (PPTD)

• D. C. Jones, Licensing and Compliance Section (PPTD)

• 1. B. S. Ferrell, Licensing Engineer (PPTD)

• R. A. Newkirk, Technical Section (PPTD)

L. L. Vanderhorst, Plant Health Physics (PPTD)

• G. G. Rhoads, Operations Engineer (PPTD)

• 1. G. A. Dunn, Compliance Engineer (PPTD)

1. V. J. Concel, Mechanical Systems Lead (PPTD)

• E. Riley, Manager (NQAD)

• 1. B. D. Walrath, General Supervising Engineer, Operational Quality Section (NQAD)

• 1. V. K. Higaki, Maintenance and Modification Quality Section, (NQAD)

• W. G. Morris, Quality Engineer (NQAD)

The inspectors also contacted and interviewed other licensee and contractor personne * Denotes those attending the exit meeting held on February 24, 1987.

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1. Denotes those attending the exit meeting held on March 3, 198 . Licensee Action on Previous Inspection Findings (92701, 92702) (Closed) Violation (440/86020-02)(DRP)): Inadequate administrative controls for the use of temporary space heaters. The inspector reviewed the licensee's response letter dated November 10, 1986, which specified procedural revisions implemented by the licensee to control the use of temporary space heaters and to require

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. engineering evaluation prior to the use of external heat sources (space heaters) in the Off-Gas System. The inspector verified by review of Plant Administrative Procedure (PAP)-0508, "PNPP Operating Rules and Practices," Revision 3, dated August 15, 1986, that procedural requirements had been established for consulting with the licensee's safety and fire protection organizations prior to the placement of temporary heaters in facilities other than offices, control rooms, locker rooms, wash rooms, laboratories, or lunch rooms. The inspector verified by review of System Operating Instruction (S01)-N64, "Off-Gas System (Unit 1)," Revision 2, dated June 26, 1986, that procedural requirements had been established for the conduct of the engineering evaluations prior to the use of any-external heat source in the off-gas system. Based upon these reviews, the inspector has no further concerns in this are (Closed) Violation (440/86025-01(DRP)): Failure to accomplish activities affecting quality in accordance with instructions and procedures. The inspector reviewed the licensee's response letter dated December 5,1986, which specified immediate and supplemental corrective actions taken in response to each of the cited procedural violations. Each violation resulted in unplanned engineered safety feature actuations. The inspector determined that immediate corrective actions by operating personnel to restore affected equipment were appropriate to the circumstances. Additionally, the inspector verified by discussions with licensee instrumentation and controls, and operations supervisory personnel that personnel counseling and training conducted in response to these violations addressed root causes and was administered to all personnel involved in similar procedural evolutions. The inspector verified the effectiveness of licensee corrective actions by review of licensee event reports (LERs) issued subsequent to the cited violations and prior to the close of this inspection period. The inspector did not identify any similar occurrences. Based upon the inspector's reviews, the inspector has no further concerns in this are c. (Closed) Unresolved Item (440/86028-01(DRP)): Licensee action on G.E. recomendation for safety related instrument channel functional testing described in Service Information Letter (SIL) 445. In response to this item, the licensee revised Instrument Administra-tive Procedure (IAP)-0503, " Plant Instrument Calibration and Maintenance," on January 23, 1987. Curing this inspection period, the inspector reviewed the subject IAP and determined that it incorporated the G.E. recommendation for the performance of channel functional testing following the completion of maintenance activities on source range monitor or intermediate range monitor channels. Based upon this and previous licensee actions regarding G.E. SIL 445, the inspector is satisfied that the issues raised in SIL 445 have been adequately addresse d. (Closed) Violation (440/86028-02(DRP)): Inadequate instructions prsnded for the installation of temporary heat tracing on safety related leak detection system flow instrument sensing lines. The

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. inspector verified by review of licensee training records, that following the events which led to this violation, the licensee provided training to appropriate craft personnel in the installation and operation of heat tracing. A course entitled, " Fundamentals of Heat Tracing," was administered to 9 individuals between October 6, 1986 and October 10, 1986. The inspector reviewed General Welding Instruction (GWI)-0005, " General Post-Weld Heat Treatment Require-ments," Revision 1, dated January 16, 1987, and verified that the procedure had been revised to extend its applicability to all heat tracing applications. Additionally, the procedure was enhanced to clarify thermocouple installation and to require a sketch detailing the arrangement of heating elements and thermocouples. By review of Purchase Requisition No. 6745-86-65 and Receiving Report N PPR070-R01, the inspector verified that the licensee had procured additional heat tracing equipment with improved temperature control characteristics as described in the licensee's January 22, 1987 response letter. Based upon these reviews, the inspector concluded '

that licensee actions were adequate and in accordance with the licensee's response lette (Closed) Open Item (440/86031-02): Operation of Gama-10 Portal Monitors in the walk-through mode. In response to this item and in order to increase portal monitor sensitivity, the licensee began operating all Gamma-10 portal monitors in the " time-out" mode on December 12, 1986. This licensee action was verified by the inspector during routine inspection activities requiring inspector use of the portal monitors prior to exiting the radiologically controlled and protected area (Closed) Open Item (440/86033-01)(DRP)): Technical Specification change request did not include an upper bound on the actuation setpoint for the containment vacuum breaker isolation valves. By letter dated February 10, 1987, the licensee revised proposed Amendment 1 to the Perry Unit 1 Operating License to include an upper bound on the containment vacuum breaker isolation valve differential pressure actuation setpoint of less than or equal to 1 0.160 psid (containment to outside containment). This upper bound will ensure that the containment isolation function of the valves l will not be impaired as a result of non-conservative (excessively )

high) containment to atmosphere differential pressure actuation setpoint . Inspection and Enforcement Information Notice (IEN) 87-08(92701)

(Closed) IEN 87-08, " Degraded Motor Leads in Limitorque DC Motor Operators." The inspector verified by discussions with licensee personnel and document review, that the licensee had received the subject IEN, reviewed it for applicability, and taken necessary actions to address the identified concern. Licensee review of all installed Limitorque DC Motcr Operators identified two operators employing DC motors with non-environmentally qualified defective motor leads. The motor operators in question were utilized on Reactor Core Isolation

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Cooling system minimum flow valve IE51-F019 and Lube Oil Cooler supply valve 1E51-F046. On February 14, 1987 the licensee installed qualified replacement DC motors in operators associated with these valve Facility operation prior to this date with the unqualified DC motors is a potentially enforceable unresolved item which will be further reviewed by the NRC Region III Office's, Division of Reactor Safety (440/87003-01)(DRS)).

4. Operational Safety Verification (71707) General The inspectors observed control rcom operations, reviewed applicable logs, and conducted discussions with control room operators during this inspection period. The inspectors verified the operability of selected emergency systems, reviewed tag-out records and verified tracking of Limiting Conditions for Operation associated with affected components. Tours of the intermediate, auxiliary, reactor, and turbine buildings were conducted to observe plant equipment conditions including potential fire hazards, fluid leaks, and excessive vibrations, and to verify that maintenance requests had been initiated for certain pieces of equipment in need of mainten-ance. The inspectors by observation and direct interview verified that the physical security plan was being implemented in accordance with the station security pla The inspector observed plant housekeeping / cleanliness conditions and verified implementation of radiation protection control These reviews and observations were conducted to verify that facility operations were in conformance with the requirements established under technical specifications,10 CFR, and administrative procedures, "oluntary Entries Into Technical Specification Limiting Conditions for Operations Action Statements During this inspection period, the inspector reviewed the sequence of events occurring on January 9,1987, involving multiple voluntary entries into the action statements associated with the technical specification Limiting Condition for Operation for the Leak Detection Syste On January 9, 1987, while in Operational Condition 2, the licensee attempted to repair flange leaks on the reactcr water cleanup system regenerative heat exchangers. The repair process involved the injection of a sealant compound into specially designed collar Work Order numbers 86-12426 and 86-16360 provided instructions for the repair activity and specifically required that the reactor water cleanup leak detection isolation function be bypassed during injection of the sealant compound. The isolation function was to be bypassed to prevent system isolation while the system was aligned to transfer reactor coolant and any sealant compound entering the system to the liquid radwaste syste ..

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. The isolation function was bypassed as permitted by Technical Specifications and in accordance with licensee administrative procedures which permitted bypassing the isolation function for up to one hour. Prior to exceeding the one hour. time limit, operators restored the isolation function to service though the repair work was not complete.- After obtaining a revised estimate of the repair completion time, operators again bypassed the isolation function and imposed a new one hour time limit for restoration of the isolation function, zThis sequence of events was repeated several times as the repair work continued to b; delayed beyond estimated times of completion. As a result, the isolation function was bypassed 6 times for a total of 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and 21 minutes over a period of 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and 36 minute The inspector was concerned that while the letter of the Technical Specifications was met, the repeated voluntary entry into the Technical Specification Action Statements for the reactor water

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cleanup system isolation function inoperability over the time frame in question did not meet Technical Specification intent. The inspector discussed this matter with licensee management to get a more detailed understanding of the rationale employed by operating personnel involved in the'above sequence of events. The.1,icensee

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acknowledged the inspector's concern and concurred that in hindsight the sequence of events was inconsistent with the intended use of the flexibility afforded by the provisions contained in the Technical Specification Limiting Condition for Operation. The inspector was informed that the decisions to make multiple use of the Technical Specification provisions for leak detection system inoperability were prompted by multiple reports from personnel directly involved in the repair work, that the repair work was near completion. This matter was further discussed with the licensee during the Management Meeting documented in Paragraph 12 of this report.'

As a result of these discussions, the licensee agreed to revise administrative procedures to provide guidance on voluntary entry into Technical Specification Limiting Conditions for Operation-consistent with Technical Specification intent. These actions will be tracked as an open item (440/87003-02(DRP)).

c. 10 CFR 50.59 Reviews for Temporary Alterations During a routine Control Room tour and Panel walkdown conducted by the inspector on February 2, 1987,.the inspector noted that process radiation monitor CD17K0821A'was inoperable and that an information tag was hung referencing Tag Order (temporary alteration) number 16-566. Subsequent inspector review of the Tag Order Log disclosed--

that on July 24, 1986, the plant underdrain system pump trip interlock with radiation monitor OD17K0821A had been defeated by determination of a wire in the radiation monitor /underdrain pump interlock circuitry. Because the radiation monitor /underdrain pump trip interlock is described in the FSAR, the inspector attempted to determine whether or not the licensee had conducted a safety

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evaluation of the tag order in accordance with the requirements of 10 CFR 50.59 prior to implementation. The inspector contacted the responsible system engineer and was informed that the required evaluation had not been performed due to a misinterpretation of licensee administrative procedure The inspector reviewed the following Plant Administrative Procedures (PAPS) which governed the conduct of safety evaluations required by 10 CFR 50.59 for temporary alterations:

PAP-0205 " Operability of Plant Systems," Revision 3, dated February 3,1986 PAP-0305 " Safety Evaluations," Revision 4, dated February 10, 1986 PAP-1402 " Control of Lifted Leads, Jumpers, Temporary Electrical Devices, and Mechanical Foreign Items," Revision 4, dated January 30, 1787 PAP-1402 directed that temporary alterations affecting plant systems required to be operable per PAP-0205 have safety reviews performed per PAP-0305. Licensee personnel implemented this requirement by

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referring to PAP-0205, Attachment 1, "PNPP Systems Requiring l Independent Verification." If temporary alterations were judged to affect systems or portions of systems listed in PAP-0205, attachment 1, they were evaluated in accordance with the requirements of PAP-0305. If the temporary alterations did not affect systems or portions of systems listed in PAP-0205, Attachment 1, they did not necessarily receive reviews in accordance with PAP-030 Inspector review of PAP-0205, Attachment 1, determined that it did !

not include all plant design features described in the FSAR and, therefore, subject to the provisions of 10 CFR 50.59. Following discussions with the licensee concerning the failure to conduct the l

safety evaluation prior to implementation of temporary alteration !

! which defeated the process radiation monitor /underdrain pump trip interlock, the licensee conducted reviews of all active tag orders ;

in accordance with PAP-0305. From these reviews, 11 additional temporary alterations were identified which should have received

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safety evaluations required by 10 CFR 50.59 prior to implementatio The identified failures to conduct safety evaluations required by 10 CFR 50.59 for changes to the facility as described in the FSAR, is a violation (440/87003-03(DRP)).

In response to the identified violation, the licensee conducted safety evaluations for all 12 temporary alterations which constituted changes to the facility as described in the FSA These reviews which were conducted between February 2, and February 5, 1987, determined that none of the temporary alterations involved an unreviewed safety questio )

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In order to prevent recurrence of this type of violation, the-licensee revised PAP-1402 on February 6, 1987 to require that

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all tag orders be reviewed in accordance with PAP-0305 prior to implementation. The inspector determined that the foregoing licensee actions were adequate to address both the effects and

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the root cause of the identified violatio Reportability of Reactor Core Isolation Cooling System Inoperability Pursuant to 10 CFR 50.72

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On February _ 10, 1987, the inspector was contacted by the Operations

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- Shift Supervisor and informed that the Reactor Core Isolation Cooling (RCIC) System had failed a pump operability test, and that the licensee was preparing to make a four-hour non-emergency NRC l Notification required by 10 CFR 50.72 for inoperability of a single

train safety syste Earlier.that day, the inspector had noted that the RCIC system was inoperable. The inspector, therefore, questioned the Shift Supervisor as to whether or not the test failure necessitated an

additional NRC notification. Following that discussion, the

• inspector _ reviewed the operability status of the RCIC. system for the previous week. The inspector determined that the RCIC system had been declared inoperable on February 8, 1987, due to a failure i

of valve 1E51-F063 to stroke and that its status remained unchanged

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and informed him of the results of this review. The Shift i

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Supervisor had performed a similar review and determined that the  :

RCIC system had been declared inoperable on February 8,1987, at j 11:32 p.m. Additionally, NRC notification of RCIC system  !

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i inoperability had not been performed within the four hour time j limit specified in 10 CFR 50.72. At approximately 3:25 p.m. on

February 10, 1987, approximately 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> beyond the'four hour time j

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limit, the licensee provided the required notification. Failure to report Reactor Core Isolation Cooling System inoperability within the four time interval specified in 10 CFR 50.72 is a violation j (440/87003-04(DRP)).

Onsite Followup of Non-Routine Events at Operating Power Reactors (93702)

Unplanned Orderly Reactor Shutdown Due to Questionable Seismic

, Qualification of Leak Detection Temperature Senser Installations t On February 5,1987, the' licensee conducted an unplanned, orderly shutdown following an initial determi u tion that 23 leak detection system temperature instruments identified in Nonconformance Report

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No. NEDS 1447 were inoperable due to indeterminate seismic

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qualification as installed.

! Nonconformance Report No. NEDS 1447 documented that the temperature

elements were not rigidly attached to their supports because

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mounting threads on the temperature element assemblies did not fully engage their supports. The Nonconformance Report had previously i been dispositioned as " rework /use as is." This had been interpreted l by operating personnel as meaning that the' nonconforming condition j should be eventually corrected but that in the interim, the.

, operability of the temperature instrumentation was unimpaired.

On February 4, 1987, the Nonconformance Report was re-reviewed by

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disposition was obtained from cognizant engineering personne Engineering personnel explained.that the stated disposition was meant to convey the fact that the temperature elements, themselves,

were not damaged by the nonconforming condition and, therefore, did not need to be replaced. However, in order to assure seismic qualification of the affected temperature elements, rework was

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specified. Based upon the foregoing clarification, operatin personnel declared the temperature instrumentation inoperable and initiated a plant shutdown at 12:30 p.m. on February 5,198 At approximately 2:15 p.m. the same day, the licensee received the results of an evaluation of the _ identified nonconforming condition conducted by the architect / engineer, Gilbert Associates, Inc. The evaluation concluded that seismic qualification of the temperature instruments had not been impaired.

, tihile shutdown between February 5 and'7, 1987, the licensee

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conducted a re-review of all 299 open Nonconformance Reports concerning installed safety related items. The re-review was conducted to determine whether or not any of the nonconforming conditions affected equipment operability and to identify any corrective actions required prior to plant startup., In instances where the nonconforming conditions' impact on equipment operability was questionable, the nonconforming conditions were correcte prior to plant startup on February 7,1987. Subsequent licensee

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evaluations of all of these nonconforming conditions determined {

that none of them had impaired the operability of affected equipmen {

Inspector review of this matter included a review of Nonconformance Report No. NEDS 1447, the documented "use as is" disposition received from Gilbert Associates, Inc., and the results of the licensee's re-review of safety related Nonconformance Reports documented under Condition Report 87-0075. Based upon these reviews, the inspector determined that licensee actions regarding this matter were appropriat No violations or deviations were identified, Reactor Scram Due to Spurious low Reactor Water Level Signal During Instrument Calibration On February 13, 1987, at approximately 10:38 P.M. while in operational condition 2, with reactor power at 6%, a reactor scram occurred due to scram discharge volume high water level. The scram

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. discharge volume high water level resulted from an alternate rod insertion initiated by Division 2 of the redundant reactivity control syste Instrumentation and control technicians had completed a calibration of reactor level indicator IB21-R0098 which shared common reactor vessel level instrument taps with both Division 2 redundant reactivity control system level transmitters. During instrument restoration, opening of the low side instrument isolation valve for the level indicator resulted in a pressure transient in the common instrument piping of sufficient magnitude to cause the Division 2 redundant reactivity control system level transmitters to generate low level-2 trip signals. The low level-2 trip signals resulted in an alternate rod insertion and tripping of the reactor recirculation pump Following the occurrence, the redundant reactivity control system and reactor protection systems were reset and the licensee commenced a cooldown and depressurization to place the unit in cold shutdow On February 14, 1987, the inspector reviewed the licensee's post scram evaluation documented in Scram Evaluation Report No. 87-0 The scram evaluation which included a chronological sequence of events; written statements from onshift operating personnel and instrumentation and control technicians; the Sequence of Events Log; and Time History Plots of pertinent plant parameters obtained from the process computer. Based upon the inspector's review, the inspector determined that the root cause of the scram had been identified and that plant response was as expecte In order to prevent similar occurrences in the future, the licensee had initiated a design change to relocate redundant reactivity control system level transmitters such that level transmitters associated with the same system divirion do not share common level instrument sensing lines. The licensee intends to perform this modification during a future scheduled plant outag No violations or deviations were identifia Reactor Vessel Level Instrumentation Anomoly On February 16, 1987 during the performance of Reactor Core Isolation Cooling (RCIC) system testing, in which the system was providing ficw from the condensate storage to the reactor vessel, reactor vessel level instrumentation associated with reference leg 1821-D004A trended offscale high beginning approximately 5 minutes after initiation of RCIC flow to the reactor vesse The licensee declared the instruments inoperable and took Technical Specification j required action Approximately 55 minutes later, similar behavior of level instruments associated with reference leg 1821-0004B was observe . . ._ -, . - - - . . - -, -. - .-.

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In the following 6 minutes, the licensee decided to terminate the RCIC test and RCIC flow to the reactor vessel was secured. Within seconds, level instrument readings from affected instruments returned onscale and were in agreement with all other level instruments. The RCIC system was placed in secured status and the level instruments were declared operable. Subsequent testing was conducted on February 18, 1987 to determine if there existed a threshold for RCIC flow to the reactor vessel below which the level instrumentation was unaffected. The testing indicated that the level instrumentation was affected at RCIC flows above 140 gp Licensee evaluation of the RCIC/ reactor level instrumentation interaction conducted in consultation with General Electric, concluded that during RCIC injection via the reactor vessel head spray nozzle resulted in an accumulation of water in the instrument sensing lines associated with the reactor water level instrumenta-tion reference legs. The resultant level instrument error corresponded with the elevation difference between the reference leg condensing chambers and the instrument sensing line reactor vessel penetration As a result of the RCIC/ reactor level instrumentation interaction, the licensee requested and received an emergency Technical Specification change on February 26, 1987, which permitted disabling of the RCIC automatic injection function for 30 days in order to permit completion of startup testing associated with Test Condition 1 (32% power plateau). Prior to expiration of the 30-day time interval, the licensee was expected to provide the results and conclusions of additional evaluations of this matter and propose a schedule for corrective actions to eliminate the RCIC/ reactor level instrumentation interactio No violations or deviations were identifie Divisions 1 and 2 Standby Diesel Generator Inoperability On February 27, 1987 at approximately 5:08 P.M., while in Operational Condition 1, the Division 1 Standby Diesel Generator failed a special test to prove that the diesel generator could start and stabilize within 10 seconds using only 1 of 4 air start admission valve Lipon startup, the diesel generator experienced a loss of control air due to a leaking control air valve. Personnel secured from the special test and attempted to prove the diesel generator operable by starting the diesel with normal redundant air start capabilities. Again, there was a loss of control air and the Division 1 diesel generator was declared inoperabl In accordance with technical specification Action Statements, the Division 2 diesel generator was started to demonstrate operability, but it also failed due to a leaking control air valve and was declared inoperable at approximately 7:05 P.M., February 27, 198 _ _ _

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The Division'3 diesel generator was' tested and' demonstrated operable. Both Divisions 1 and 2 diesel generators leaking control air valves were replaced. Following the maintenance,-successful maintenance tests were performed on the Division 1:and 2 diesel generators. Operability of all three diesel generators was established by 3:00 A.M.', on February 28, 1987. Restoration of the diesel generators to operable status was accomplished within the time _ limits specified in Technical Specifications for continued plant operatio Subsequent to these occurrences,_cn March ~2, 1987 the inspector was verbally informed by licensee personnel .that the leaking control air valves had been previously identified and documented via maintenance

. Work Requests. The inspector will review this matter during a subsequent inspection to further assess circumstances surrounding the diesel generator' failures and the timeliness of licensee corrective actions for the identified leaking valves. This matter is an unresolved item (440/87003-05(DRP)). Reactor Scram ar.d High Pressure Core Spray System Injection Due to Loss of Feedwater At approximately 3:20 A.M. on March 2,1987, a reactor scram occurred due to low reactor water level (low level 3) which.resulted from a trip of the "A" turbine driven feedwater pump. At the time of the occurrence, the unit was operating at 28% power with feedwater being supplied via "B" and "D" feedwater booster pumps and "A" turbine driven feedwater pum Operators had been dispatched to investigate a faulty level switch on the hot surge tank which was causing erroneous hot surge tank low-level annunciation in the Control Room. The level switch mechanism was believed to be stuck in the-actuated position. _0perating personnel attempted to clear the condition by tapping on the level switch housing, but instead of tapping the hot surge tank level switch associated with the alarming annunciator, they tapped the level switch which generated a feedwater booster pump trip signa The operating feedwater booster pumps tripped and the "C" feedwater booster pump auto-started approximately 1 second later. The operating feedwater pump tripped due -to a trip interlock with the feedwater booster pump supply breakers which was momentarily made u Following the feedwater pump. trip, reactor water level decreased to low level 2 and the high pressure core spray system auto-initiated and supplied water to the reactor vessel. Feedwater booster pump

"D" was manually restarted, the motor driven feedwater pump was manually started, reactor water level was restored, and the high pressure core spray system was secured within the following-4 minutes. The reactor core isolation cooling system received an autostart signal, but did not initiate due to a failure of the inboard steam supply isolation-valve IE51-F063 to ope .- -. .- - .- - .- . . . .- -

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As a result of the unplanned high pressure core spray initiation

{ and injection into the vessel, the licensee declared.an Unusual

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Event in accordance with Emergency Plan Instruction (EPI)-A1.

! Following completion of initial and termination notifications of j the counties, the state, the Coast Guard, and the NRC, the Unusual

Event was terminated at 4

35 A.M.

. The inspector developed the foregoing information by interview of '

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onshift operating personnel and review of the licensee's post scram 1 -evaluation documented in Scram Evaluation Report No. 87-02. The licensee attributed the reactor scram to personnel error and more j specifically, a failure to verify the-identity of the hot. surge tank

, -level switch associated with the control room annunciator. The inspector agreed with the' licensee's determination of root cause and

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restart action items. At the close of the inspection, the plant remained shut down pending completion of plant startup

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prerequisite No violations or deviations were identifie . Engineered Safety Feature (ESF) System Walkdown (71710)

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i During this inspection period, the inspector performed a detailed walkdown of the Division 2 Emergency Diesel Generator (D/G), its starting i air subsystem, and its fuel oil subsystem. This system walkdown was t conducted utilizing Valve Lineup Instruction (VLI)-R45, " Division 1 and 2 i Diesel Generator Fuel Oil System (Unit 1)," Revision 2, dated January 30,

! 1986 and VLI-R47, " Division 1 and 2 Diesel Generator Starting Air System j (Unit 1)," Revision 3, dated September 11, 1986. The inspector also i

referred to the Standby Readiness Checklist and other appropriate portions of System Operating Instruction (S01)-R43, " Division 1 and 2 i

Diesel Generator System (Unit 1)," Revision 4, dated March 14, 198 During the inspection, the inspector verified VLI-R44 and VLI-R45 against controlled Piping and Instrumentation Diagrams (P&ID) for the starting

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air and fuel oil systems, respectively; The inspector verified most of these P& ids against the actual system as-built configuration. No-discrepancies were identified as a result of these verification '

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During the sys'em walkdown, the inspector directly observed equipment conditions to verify that appropriate levels of cleanliness were being

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maintained; valves in the systems were installed correctly and did not exhibit gross packing leakage, bent stems, missing handwheels, or

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improper labeling; and that major system components were properly labeled and exhibited no leakage. The inspector verified that instrumentation associated with the system was properly installed, functioning, and.that

significant process parameter values were consistent with normal expected values. By direct visual observation or observation of remote position L

indication, the inspector verified that valves in the flow paths of the i systems were in the correct positions as required by VLI-R44, VLI-R45,

and S01-R43; that where required, power was available to the valves; valves required to be locked in position were locked; and that pipe caps i and blank flanges were installed as required. Similarly, the inspector

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verified proper breaker and switch positions at local electrical boards and indications on control boards in accordance with S0I-R4 '

At the time of the system walkdown, the Division 2 Emergency Diesel Generator had been declared operable and was being maintained in standby readiness. During the walkdown on February 10, 1987, the inspector noted that the discharge valves (IR45-F552B and IR45-F504B, respectively) for Fuel Oil Trans'er pumps 1 and 2 for Division 2 did not appear fully open as required by VLI-R45. These valves and the corresponding discharge valves for the Division 1 Fuel Oil Transfer pumps appeared to be about 3 turns open instead of being fully 15 turns ope Upon the inspector's request, a Perry Plant Attendant (PPA) verified the positions of the Division 2 discharge valves and reported them to the Control Room to be approximately 2-1/2 turns open instead of fully ope Subsequently, the PPA reported to the Control Room that the Division 1 fuel oil transfer pump discharge valves were fully closed. Following each discovery, the valves were fully opened. The licensee initially determined that the Diesel Generators had been inoperable and reported the matter to the NRC via the Emergency Notification System (ENS) phone in accordance with 10 CFR 50.72 (B)(1)(II). Subsequently, the PPA who had reported the Division 1 valves as being fully closed informed licensee management that the valves were perhaps a turn or more ope This was approximately consistent with the resident inspector's observatio The licensee determined by document review that the Division 1 and 2 valves were left in a throttled position following inservice testing of the fuel oil transfer pumps conducted on January 27, 1986, and February 2,1987, respectively. On February 10, 1987, the licensee conducted pump flow tests and concluded that more than adequate flow was available to support diesel generator operation with the discharge valves throttled between 1-1/4 and 1-3/4 open. Inspector review of the pump flow data obtained by these tests and diesel generator fuel consumption data obtained during preoperational testing confirmed the licensee's conclusion. On February 11, 1987, valve positions for each of the four valves were recreated based on visual observation made on February 10, 1987, by the resident inspector. In each case it was determined that the valve was open sufficiently to pass required pump flows . Pump test procedures were revised to require that the valves be returned to the full open position upon completion of inservice testin No violations or deviations were identifie . Licensee Event Reports Followup (92700)

Through direct observations, discussions with licensee personnel, and review of records, the following event reports were reviewed to determine that reportability requirements were fulfilled, immediate corrective action was accomplished, and corrective action to prevent recurrence had been accomplished in accordance with technical specification *

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. LER 86051-0 failure to Document Design Change Results in RHR Valve Inoperabilit LER 86052-0 Failure to Test Relay Contacts Results in Technical Specification Violatio LER 86054-0 During Initial Startup Testing, Equipment and Design Problems Cause RCIC Isolation e LER 86055-0 Isolated Instrument Results in PressuriTransient and Reactor Scra LER 86064-0 Personnel Error Causes Inoperability of Leak Detection i System Division LER 86066-0 Reactor Core Isolation Cooling System Isolation Due to

Indeterminate Caus LER 86069-0 Personnel Error Causes Control Rod Technical Specification Violatio LER 86070-0 Conservative Design Setpoints Result in Actuations of the Drywell Vacuum Relief Syste LER 86072-0 Design Deficiency Causes Loss of Reactor Protection System Bu LER 86077-0 Personnel Error Results in Violation of Primary Containment Integrit LER 87003-0, " Failure of Valve IE51-F063 to Stroke," was also reviewed during this inspection. On January 10, 1987 (and on a number of subsequent occasions), Reactor Core Isolation Cooling (RCIC) system inboard containment isolation valve 1E51-F063 failed to open; and, on at least one occasion, it failed to close. The valve is a 10 inch, Borg-Warner, flexible wedge, rising stem gate valve with a Limitorque, SMB-1 motor-operator, with a 4.3 HP, 125 VDC moto The inspector reviewed the licensee's evaluation of the failures and troubleshooting activities to determine the root cause. It appears that the valve closure failure may have been unrelated to the opening failures, having occurred shortly after a valve packing tightening to correct a packing leak at reactor pressure. Efforts to determine the opening failure causes have included valve cycling under various thermal conditions and system stator pressures. Motor Operated Valve Analysis and Test System (M0 VATS) diagnostic testing has been performed; valve

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and valve operator vendor representatives have examined the equipment, including inspection after disassembly of both the valve and operator; and, physical examination during handwheel operation has been performe As yet, the problem has not been isolate l

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While the M0 VATS data show some anomolous conditions, they do not reveal the problem. There is some question about the validity of the data because of hich temperature conditions that may affect the equipment during testing. Handwheel operation, using a torque wrench, indicated that approximately 215 foot-pounds of torque was required to open the valve immediately after an open stroke failure. This is excessively high compared to 70 foot-pounds handwheel torque determined necessary to open the close switch in the close direction. Such a difference would suggest high inertial loading on closure with the motor; however, MOVATS data fails to support thi To address the problem, the licensee intends at this time to: Increase the motor-operator thrust capability by conversion to a more powerful, 480 VAC motor that is capable of operating the valve against a differential pressure of 1177 ps . Reduce inertial loads to the valve disc by changing from an SMB-1 to an SB-1 model operato . Change the valve logic to normally open, as originally designed, to reduce the probability of RCIC loss because of failure to open.

d Continue testing after operator replacement to further evaluate valve operating condition While reviewing valve problems, the inspector noted that valve packing controls may be inadequate, apparently resulting in one valve failure to close. Further review of this area by the licensee and subsequent inspector review is considered an open item (440/87003-06)(DRS).

In the past, the licensee had considered the affect of system static pressure acting on the valve stem cross section (stem piston affect) as negligible; however, depending on reactor system pressure, the difference between opening and closing thrust requirements for valve 1E51-F063 can be 10,000 to 12,000 pounds. The piston affect alone caused the valve motor running current to go from 12 amperes when closing at zero static pressure to 44 amperes against full reactor pressure. The licensee indicated that the stem piston affect will be factored into diagnostic testing in the futur While the licensee has increased the wiring ampacity from the battery to the valve motor, in response to earlier problems identified, it is not apparent to the inspector that all power supply problems were resolve A valve motor curve, supplied by the licensee, indicates that maximum torque is provided by the motor at 172 amperes locked rotor current, supplied by 125 VDC. Test data obtained indicated that for 130 VDC battery voltage, the locked rotor current to the motor was 124 amp (Calculations indicate that the voltage to the valve in this condition would be 90 volts.) Hence, if torque is proportional to the square of the current, only 52 percent of the maximum torque on the motor curve would be available for valve operation. This is even with the battery

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voltage 5 volts above nominal rating for the motor. Pending the resolution of

.(440/87003-07) questions in this area, this is considered an open. item (DRS).

This LER will remain open pending successful corrective action by the licensee and further inspector revie . Monthly Surveillance Observation-(61726)

The inspector observed the performance of various portions of SVI-E51-T2001, " Reactor Core Isolation Cooling (RCIC) System Pump an Valve Operability Test," Revision 6. At approximately 11:05 a.m. on January 22, 1987, the first attempted performance resulted in only 400 gpm returning to the Condensate Storage Tank (CST) but the RCIC pump discharge pressure was greater than the simulated design reactor pressure would be, therefore, operating personnel opened the test return valve more to match the pump discharge pressure to design pressure. .The surveillance requires RCIC to achieve 700 gpm flow. Before a new attempt could be commenced, it was found necessary to pump down the suppression pool because the level was high. Other plant activities and an eleven day maintenance outage interrupted the surveillanc At approximately 2:55 P.M. on February 8, 1987, the inspector observed on Control Room panel P601, that the RCIC pump achieved over 700 gpm within 18.4 seconds with a RCIC pump discharge pressure of.approximately 1,150 psig thereby meeting the acceptance criteria for the RCIC turbine and pump quick start portion of SVI-E51-T2001, Revision At approximately 3:40 p.m. a Division 1 RCIC system isolation occurred due to a low indication on the Residual Heat Removal (RHR)/RCIC steam line flow instrument. This terminated surveillance testing and prevented its completion at that time. The calibrations of all Measuring and Test Equipment (M&TE) used in the surveillance test were current. The proper revisions of the test were used, licensee review and approval of the test were correct, comunications were adequate, test crew staffing was adequate and they were knowledgeable about the test. All Limiting Conditions for Operation (LCOs) were me No violations or deviations were identifie . Monthly Maintenance Observation (62703)

Station maintenance activities of safety related systems and components listed below were observed / reviewed to ascertain that they were conducted in accordance with approved procedures, regulatory guides and industry codes or standards and in conformance with technical specification The following items were considered during this review: the limiting conditions for operation were met while components or systems were removed from service; approvals were obtained prior to ir.itiating the work; activities were accomplished using approved procedures and were inspected as applicable; quality control records.were maintained;

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activities were accomplished by qualified personnel; parts and materials used were properly certified; radiological controls were implemented; and, fire prevention controls were implemente The following maintenance activities were observed / reviewed:

. Repair of leaks in the air supply lines associated with the lower personnel airlock inflatable seals conducted under Work Order No and 87-515 on February 3, 198 .

Motor Operated Valve Analysis and Testing System (M0 VATS) testing of Reactor Core Isolation Cooling System inboard steam supply isolation valve on January 22, 198 No violations or deviations were identifie . Startup Test Witnessing (72302)

On February 25, 1987, the inspector witnessed various portions of Startup Test Instruction (STI-C51-011, " Local Power Range Monitor (LPPJ1)

Calibration," Revision The procedure of appropriate revision was in use by all members of a sufficiently staffed test crew. All test prerequisites and initial conditions were met except those postponed which were reviewed and approved in accordance with procedural requirements of the test program. Properly calibrated plant equipment was used to take measurements (ERIS), on a real time basis. The test procedure was technically adequate. Crew actions and coordination appeared sufficien Plant response was analyzed as correct, in a rapid manner on a real time basis and all test data was collected for final analysis by proper personnel. The inspector ensured that for the sections of the test that were observed, the test acceptance criteria was met as asserted by licensee's preliminary test results evaluation. The inspector ensured that the requirements of the Limiting Conditions for Operation (LCO) of Technical Specification 3/4.10.2 were adhered to during the testing. The inspector concurred with licensee personnel's preliminary evaluation that the acceptance criteria appeared to be satisfie No violations or deviations were identifie . Onsite Review Committee (40700)

The inspectors reviewed the minutes of the Plant Operations Review Committee (PORC) meetings No.86-249, 253, 257, 258, and 87-001 through 87-027 conducted prior to and during the inspection period to verify confonnance with PNPP procedures and regulatory requirements. These observations and examinations included PORC membership, quorum at PORC meetings, and PORC activitie No violations of regulatory requirements or deviations from commitments were identified in this are ,

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12. Plant Status Management Meetings (30702)

On January 21, and February 26, 1987, NRC management met with CEI management to discuss the current status of the plant and recent events. The January 21 meeting was conducted at the Perry sit The February 26 meeting was conducted at the NRC's Region III offic These meetings are being held on a periodic (initially monthly) basi The meetings each included discussions of: the status of the plant; recent Licensee Event Reports (LERs); corrective actions taken or planned to be taken to preclude repetition; and, the schedule for future evaluations. The February 26 meeting also included a discussion of technical specification intent with regard to voluntary entry into Action Statements associated with Limiting Conditions for Operatio . Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether it is an acceptable item, a violation or a deviation. An unresolved item is identified in Paragraphs 3 and 5 . Open Inspection Items Open inspection items are matters which have been discussed with the licensee, which will be reviewed further by the inspector, and which involve some action on the part of the NRC or licensee or both. Open inspection items disclosed during the inspection are discussed in Paragraphs 4b and . Exit Interviews (30703)

The inspectors met with the licensee representatives denoted in Paragraph 1 throughout-the inspection period and on February 24, 1987 and March 3, 1987. The inspector summarized the scope and results of the inspection and discussed the likely content of the inspection report. The licensee did not indicate that any of the information disclosed during the inspection could be considered proprietary in natur I

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