IR 05000440/1988003

Insp Rept 50-440/88-03 on 871231-880222.Violations Noted. Major Areas Inspected:Previous Insp Items,Operational Safety,Nonroutine Events,Maint,Surveillance,Ler,Allegations, Physical Security & Radiological Controls

ML20150C080
Person / Time
Site:	Perry
Issue date:	03/04/1988
From:	Knop R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20150C067	List: IR 05000440/1988003 ML20150C064 ML20150C072
References
50-440-88-03, 50-440-88-3, NUDOCS 8803170318
Download: ML20150C080 (18)

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d U.S. NUCLEAR REGULATORY COMMISSION

REGION III

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Report No. 50-440/88003(DRP)

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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: December 31, 1987 through February 22, 1988 Inspectors: K. A. Connaughton "

G. F. O'Dwyer Approved By . nop, Chi )

p actor Projects Branch 3 'Date Inspection Summary Inspection on December 31, 1987 through February 22, 1988 (Report N /88003(DRP))

Areas Inspected: Routine unannounced inspection by resident inspectors of ,

previous inspection items, operational safety, nonroutine events, maintenance, l surveillance, engineered safety features, containment closeout, licensee event

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reports, allegations, onsite review committee activities, physical security, i and radiological control Plant status meetings between licensee and NRC ,

regional management personnel were conducted on February 4 and 19, 198 {

Results: Of the 12 areas inspected, two violations were identified in the i operational safety area (failure to take required actions for inoperable ESW loop A radiation monitor - Paragraph 3.b. , and; failure to supplement special :

report for diesel generator failure as required - Paragraph 3.c.); and one i violation was identified in the licensee event report area (failure to take l required actions for an inoperable Turbine Building / Heater Bay vent radiation monitor - Paragraph 9). Additionally, two other violations were identified in

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I the second area; (failure to perform channel checks for source range neutron ]

monitors and control room radiation monitor - Paragraph 9, and; failure to !

maintain RHR head spray containment isolation valve deenerized while inoperable due to lapsed surveillance - Paragraph 9). .However, in accordance ,

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with 10 CFR 2, Appendix C, Section V.A., a Notice of Violation was not issue l During this inspection, the licensee completed a planned maintenance outage '

approximately four weeks in duration. The next scheduled outage is expected i to occur in October 198 !

8803170318 880303 5  !

PDR ADOCK C5000440 O l PDR

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. r DETAILS

. Persons Contacted

1. Alvin Kaplan, Vice President, Nuclear Group '

+#* C. M. Shuster, Director, Nuclear Engineering Department (NED)

B. D. Walrath, Manager, Engineering Projects Support Section (NED)

D. R. Green, Manager, Electrical Design Section (hED)

l K. R. Pech, Manager, Mechanical Design Section (NED)

H. D. Lyster, General Manager, Perry Plant Operations Department (PP00)

+#* R. A. Stratman, Manager, Operations Section, (PP00) i R. P. Jadgchew, Manager, Instrumentation and Controls-Section (PP00)

A. F. Silakoski, Operations Section-(PP00) ,

1. • V. K Higaki, Manager, Outage Planning Section (PP00) ,

+#* F. R. Stead, Director, Perry Plant Technical Department (PPTD)-  ;

1. • W. R. Kanda, Manager, Technical Section (PPTD)

S. F. Kensicki, Technical Superintendent (PPTD)

P. A. Russ, Licensing and Compliance Section (PPTD)

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T. L. Heatherly, Licensing and Compliance Section (PPTD)

G. S. Cashell, Licensing and Compliance Section (PPTD)

L. L. Vanderhorst, Radiation Protection Section (PPTD)

• 1. E. M. Buzzelli, Manager, Licensing and Compliance Section (PPTD)

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

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S. J. Wojton, Manager, Radiation Protection Section (PPTO)

1. E. Riley, Director, Nuclear Quality Assurance Department (NQAD)

T. A. Boss, Supervisor, Quality Audit Unit (NQAD)

R. H. Simmons, Operational Quality Section (NQAD)

W. E. Coleman, Manager, Operations Quelity Section (NQAD)

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D. J. Takas, Manager, Mechanical Maintenance Quality Section (NQAD)

• Denotes those attending the exit meeting held on February 22,.198 # Denotes those attending the February 4, 1988 plant status meetin + Denotes those attending the February 19, 1988 plant status meetin . Licensee Action on Previous Inspection Findings (92701, 92702) (Closed) Open Item (440/85089-01(DRP)): Resolution of underdrain system preoperational test discrepancies. During the performance of ,

underdrain system preoperational test OP72-P002, one piezometer,

1. 17, was identified as acting more slowly than the other unit Test records indicated that this piezometer lagged the other piezometers by approximately six hours during the test by less than one foot. The preoperational test procedure called for ,

the underdrain system to be capable of reaching an equilibrium ground water level of less than 568 fee Due to the sluggish response of piezometer #17, at the end of the 48-hour drawdown cycle specified in the test procedure, piezometer #17 had a reading

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of 568.16 feet while all other units were less than 568 feet.

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A deficiency report (NTS-5526) was issued to document and resolve this conditio The resolution of this deficiency was to "use as is."

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Based upon a review of preoperational. test data which demonstrated that by continuing the drawdown cycle, ground water level at piezometer #17 could be drawn down below the 568 foot elevation, the 48-hour drawdown cycle initially specified in the test procedure was determined to be overly prescriptive for the purposes of meeting test objective During the test, there appeared to be no significant differences in manhole water level Pump operation appeared to be within design requirement During the portion of the test requiring establishment of

. equilibrium conditions and increasing flow in 50 gpm increments, an equilibrium was established at 50 gpm input, but when the flow was increased to 100 gpm, west side water level reached elevation 570 feet. This part of the test was stopped upon reaching the 570 foot level per step 3.k.(3) in the procedur Since flow increments did not reach the combined flows suggested in the procedure, the licensee intended to perform additional testing at several flow rate Further review of this matter by the licensee oetermined that the equilibrium condition established at the 50 gpm step provided conclusive data concerning the ability of the porous concrete mat to conduct flow in the west to east direction and, therefore, test objectives were satisfie The inspector noted during the review of this item that the licensee had provided the NRC staff with an analysis in FSAR Section 10.4.5 which demonstrated that under design basis conditions (the postulated failure of the circulating water lines for both units), the high level gravity drain system alone could prevent exceeding the design basis ground water level of 590 feet. Based upon NRC staff review of this analysis documented in Perry SSER #8, technical specifications were not required for underdrain system operability and reactor shutdown as a function of water level rise in the plant underdrain system. The inspector has no further concerns regarding these matters, (0 pen) Violation (440/87004-01(DRP)): Untimely corrective action for diesel generator control air system leaks resulting in inoperable Division 1 and Division 2 diesel generators. Previous to this inspection period, the inspector reviewed the licensee's response letter dated May 20, 1987. The inspector's review

concluded that corrective actions specified in the response sad the timetable for implementation were adequate. During this inspection period, by letter dated January 27, 1988, the licensee reported the status of previously specified corrective actions and a revision to the timetable for the completion of certain corrective action Specifically, an engineering evaluation to establish a better defined service life for the control air solenoid valves was to be completed by December 30,- 1987 The January 27, 1988 letter indicated that this evaluation was not yet complete and proposed that the evaluation be completed prior to startup following the first refueling outag A design change was initiated for the j

diesel generator building ventilation system to cecrease the average L ambient temperature in the vicinity of the diesel generator control

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pane According to the January 27, 1988 letter,stnisdesigncbange '

had been finalized ar)d approved to modify the diesel generator building louver operation'and add an exhaust fan which will be operated when the diesel generator is not running to allow the room ambient temperature to be decrease This design change was <

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proposed to be implemented prior to startup following the first refueling outag In order to determine the acceptability of the revised schedule for control air solencid valve service life determination, the inspector performed a review of corrective actions taken to date, control air solenoid valve longevity, and interim actions to-ensure continued control air system reliabilit Licensee contact with the control air solenoid valve vendor determined that the most common cause of valve failure was degradation of the Buna-N poppet material due to exposure to excessively high temperatures. The valves which failed in February 1987 had been installed in the control panel for 8 to 10 years and were continuously energized for approximately 2 year All such valves on the Division 1 and Division 2 diesel generators were replaced with valves which utilize Viton poppet material which was more resistent to high temperatures. The ambient temperature in the diesel generator rooms had been reduced by' operation of existing diesel generator room ventilation equipment under administrative control. Finally, a once per-shift visual inspection of pneumatic control components was being performed and will continue until the first refueling outag Based upon the inspector's review of the foregoing, the inspector determined that the corrective action implementation schedule specified in the January 27, 1988 letter was acceptable. This matter will remain open pending inspector verification that outstanding corrective actions have been satisfactorily accomplishe (Closed) Open Item (440/87012-01(DRP)): Correction of mislabeled and unlabeled valves. The inspector reviewed Valve Lineup Instruction (VLI)-P45 for the Emergency' Service Water (ESW) System, Revision 2, and found that the licensee had corrected VLI-P45 to:

specify the proper location for valve IP45-F644 in accordance with isometric drawing, 0-P45-522, Revision 3; correctly designate valves 1P45-F648 and 1P45-F645 as ESW Supply Header Vent Valves A & B; and correctly designate the Emergency Closed Cooling (ECC) Heat ,

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Exchanger A ESW Cutlet Bypass valve as IP45-F5466A instead of F5446A. The inspector observed that the licensee labeled the following ESW valves: F530A, F637A, F5466A, F636A, F551A, F639A, F638A, and F501A. Based on the foregoing, the inspector has no further concerns regaroing this matte . Operational Safety Verification (71707) General The inspectors observed control room operations, reviewed applicable logs, and conducted discussions with control room operators during_

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this inspection perio The inspectors verified the operability of selected emergency systems, reviewed tag-out records and verified tracking of Limiting Conditions for Operation associated with-affected component Tours of the intermediate,- auxiliary, reactor, and turbine buildings were conducted to observe plant equipment l

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conditions-including potential fire hazards, fluid leaks, and excessive vibrations, and.to verify that maintenance requests had been initiated i'or certain pieces of equipment in-need of maintenance. 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 procedure ' Inoperable Emergency Service Water Discharge Radiation Monitor Due to High Background Radiation Levels During routine control room observations and panel reviews on January 5, 1988, the inspector observed that emergency service water loop A radiation monitor 1017-K604 was in a High-High alarm stat Initial discussions with operating personnel disclosed that the monitor had been in an alarm state for approximately 2 days though sampling and analysis of emergency service water effluent indicated no detectable activity. The inspector contacted licensee chemistry personnel to determine the reason for the High-High alarm conditio The inspector was informed that the alarm condition was received'

following initiation of shutdown cooling and was due to the buildup  !

of corrosion products in an unshielded stagnant low point in RHR '

system piping in the vicinity of the radiation detecto The inspector discussed this matter with the Operations Section

~ Manager and pointed out that the High-High Alarm condition caused by ..

high background radiation levels rendered the alarm function l associated with the radiation monitor incapable of alerting -)

operators to an increase in activity in emergency service water j effluen Following this discussion, the licensee declared radiation monitor 1017-K604 inoperable and implemented technical specification 3.3.7.9 required actions. Technical Specification

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3.3.7.9 required that during effluent releases via the emergenc service water loop A with the emergency service water loop A radiation monitor inoperable, grab samples were to be collected and analyzed for gross radioa9tivity once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> at a limit of detection of at least 10 microcuries per millilite Review of licensee actions from the time the monitor was initially rendered inoperable on January 3,1988 through January 5,1988 when the monitor was declared inoperable disclosed that grab samples were l

obtained and analyzed for gross activity.at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; however, the analysis procedure, which was mandated by a l

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non-Technical Specification administrative requirement, did not yield the required limit of detectio Failure to declare the ESW Loop A radiation monitor inoperable and implement technical specification required actions is a violation (440/88003-01(ORP)).

c. Diesel Generator Surveillance Technical Specification 4.8.1.2. requires that with the number of diesel generator failures greater than 1 in the last 20 valid-tests or greater than 4 in the last 100 valid tests, the surveillance test frequency is to be increased from once per 31 days to once per 7 day Testing is to be continued at the higher frequency until such time as the foregoing criteria is satisfie An alternative to the increased test frequency is provided as follows:

"For the purposes of determining the required test frequency, the previous test failure count may be reduced to zero if a complete diesel overhaul to like-new condition is completed, provided that the overhaul including appropriate post maintenance operation and testing is specifically approved by the manufacturer and if acceptable reliability has been demonstrated. The reliability criterion shall be the successful completion of 14 consecutive tests in a single serie Ten of these tests shall be in accordance with the routine surveillance requirement 4.8.1.1. and 4.8.1.1.2.a.5., four tests in accordance with the 184-day .

testing requirement of surveillance requirements 4.8.1.1.2.a.4 and 4.8.1.1.a.5. If this criterion is not satisfied during the first series of tests any alternate criterion to be used to transvalue the failure count to zero requires NRC approval."

As of the time of this inspection, the Division 2 diesel generator had experienced six failures in the last 100 valid tests and testing i was being conducted once per 7 day Four of the 6 test failures were attributed to electro pneumatic components in the cor. trol air syste As a result of these failures, and in addition to the increased frequency of surveillance testing, the licensee performed extensive modifications to the Division 2 diesel generator control air system to eliminate the failed components from the design while retaining all original control features through the use of electronic The licensee was satisfied that the cause of the 4 control air related failures had been eliminated and that the control air system had been restored to a like-new conditio Previous to this inspection, the licensee had approached the inspector as well as representatives from the NRC Office of Nuclear Reactor Regulation concerning the elimination of control air related failures from the failure count based upon the foregoing modification and restoraticn activities. The inspector informed the licensee that such an approach, with adequate technical justification, may prevent unnecessary wear and tear on the Division

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2 diesel generator resulting from the incr w 4 surveillance test :

frequenc Apparently, the licensee received similar feedback from NRC personnel in the the Office of Nuclear Reactor Regulatio However, during the discussions with the inspector, and apparently in the discussions with NRR personnel, the permissibility of such an approach under current technical specification requirements, from a f purely legal standpoint, was not explicitly discusse The licensee, however, proceeded under the assumption that the approach was permissible under current technical specification The licensee eliminated the 4 control air-related failures from the failure count and, following the weekly surveillance test of the Division 2 diesel generator conducted on January 13, 1988, began -

implementation of a monthly surveillance test frequency. The inspector became aware of this on January 22, 1988, and informed the-licensee that, as worded, current technical specifications do not provide for partial elimination of failures from the failure count based upon partial overhaul of the diesel generators. As a result, the licensee reinstituted a weekly surveil;ance test schedul ,

Further inspector review of this matter disclosed that, aside from this misapplication of technical specifications, 2 of the 14 '

successful diesel generator tests which the licensee had intended to utilize to demonstrate diesel generator generator reliability for the purposes of eliminating control air system-related failures were, in fact, performed prior to completion of modification and restoration activitie The licensee performed a review of plant equipment status during the approximately 48-hour period over which the Division 2 diesel generator was inoperable due to a lapsed surveillance interva Based upon the licensee's review and independent verification by the inspector, the inspector was satisfied that this nisapplication of technical specification provisions did not result in the violation of a technical specification limiting condition for operatio In light of the foregoing misunderstandings concerning diesel )

generator surveillance requirements, the inspector performed an i indepth review of diesel generator surveillance tuchnical ,

specifications and diesel generator surveillance test history. As a l result of this review, the inspector determined that on February 27,  !

1987 the licensee had accumulated 7 valid failures within the last l 100 valid tests when totaled for all three divisional diesel 1 Technical specification 4.8.1.1.3 required that the

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i diesel generator failures be reported to the NRC pursuant to technical specification 6.9.2 within thirty days and, if the number of failures in the last 100 valid tests on per nuclear unit basis l was greater than or equal to 7, the report was required to be supplemented to include the additional information recommended in 1 l regulatory position C.3.8 of Regulatory Guide 1.108, Revision 1, I i August 1977. A special report concerning the February 27, 1987

) failure, transmitted by letter dated March 27, 1987, did not contain the required supplemental information. Failure to provide the information required by technical specification 4.8.1.1.3 within  !

thirty days is a violation (440/88003-02(ORP)). l l

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Following identification of this matter to the licensee, the licensee submitted the required information by special report dated February 5, 198 Cognizant licensee personnel have been made aware of this reporting requirement in order to ensure future complianc .

The inspector is satisfied with the foregoing corrective actions and this matter is considered close . Followup of Nonroutine Events at Operating Power Reactors (93702)

On January 23, 1988 at approximately 7:00 p.m. while in cold shutdown, the licensee deenergized the B Reactor Protection System (RPS) bus to perform calibrations and functional testing of associated equipment protection assemblies. Upon deenergization of the bus, the inboard main >

steam isolation valves (MSIVs) opened. Following this occurrence, the inspector was dispatched to the site to assess plant status and to evaluate the results of the licensee's initial investigative effort The licensee reviewed applicable electrical design drawings and

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procedures in use at the time of the occurrence, and determined that the unexpected opening of the inboard main steam isolation valves was in accordance with desig Prior to deenergizing RPS bus R. the MSIV control switches were in the closed position and no MSIV isolation signals were present. With the inboard MSIV control switches in the closed position, relays associated with each control switch were energized from RPS bus B. When energized, the relays interrupted power from RPS bus B to the 8 MSIV pilot solenoids and interrupted power from RPS bus A to the A MSIV pilot solenoids. Both pilot solenoids were required to be deenergized to maintain the MSIVs in the closed position. Upon deenergization of RPS bus B, power was lost to the B MSIV pilot solenoids and associated logic. However, the relays

, which interrupt P.PS bus A power to the A inboard MSIV solenoids also deenergized providing power to the A inboard MSIV pilot solenoids and causing the inboard MSIVs to open. An analagous situation existed for the outboard MSIVs. Deenergization of the A RPS bus with the outboard MSIV control switches in the closed position and the MSIV logic not satisfied would have resn'ted in the outboard MSIVs opening. The inspector witnessed licensee recreation of the above circumstances and observed the MSIVs to ope ,

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aspect of MSIV control circuit design was of minimal safety significance  ;

in that the MSIVs would have closed upon loss of both RPS buses and upon receipt of an MSIV signal. If an isolation signal was present, the valves would not reopen upon loss of power to either RPS bus. If there was no isolation signal present and either the outboard or inboard valves had opened on loss of the A or B RPS buses, the valves would have immediately closed upon restoration of power to the RPS bus or upon receipt of an isolation signa The foregoing information was presented to NRC Region III management during a discussion with licensee management on January 25, 1988. As a result of this discussion, the licensee agreed to revise applicable ,

operating procedures to prevent MSIV opening during a planned RPS bus l deenergization or transfe Additionally, procedures utilized to restore

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plant systems following receipt of an isolation signal were to be revised to prevent opening of the MSIVs if plant conditions required them to remain closed. The inspector verified that these procedural revisions were implemented prior to plant startup on January 27, 1988. The licensee also agreed to provide a documented commitment to eliminate this aspect of the MSIV control circuitry design by a design change and to provide a design implementation schedule. By letter dated February 12, 1988, the licensee provided a commitment to implement a design change prior to startup after first refueling which will utilize direct contacts from new control room MSIV position switches to interrupt power to the MSIV pilot solenoid valves when the MSIV position switches are placed in the closed position. This change will eliminate the dependence of the MSIV manual closing logic on power availability. Inspector verification of licensee actions relative to the foregoing commitment, will be tracked as an open item (440/88003-03(DRP)). Monthly Maintenance Observation (62703)

Station maintenance activities of safety related systems and components described 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 initiating the work; activities were accomplished using approved procedures and were inspected as applicable; quality control records were maintained; activities were accomplished by qualified personnel; parts and materials used were properly certified; radiological controls were unnecessary; and, fire prevention controls were implemente During the period February 16 through February 22, 1988, the inspector observed significant portions of maintenance activities authorized by Work Order 8/-4078 which consistea primarily of the relocation of temperature elements 1M43-N010C and 1M43-N2108 for the Division 2 Diesel Generator Room Ventilation System. On the last day of the inspection period (February 22, 1988), the work order remained open. Inspection of the remainder of the work and documentation associated with this work order will be accomplished in a future inspection perio No violations or deviations were identifie . Monthly Surveillance Observation (61726)

On ~ January 13, January 27, and February 13, 1988 the inspector observed various portions of technical specifications required surveillances:

SVI-R43-T1318, Revision 2, "Diesel Generator (DG) Start and Load Division 2;" SVI-P53-T7312, Revision 2, "' Jpper Containment Airlock Seal Pneumatic System Leak Test," and 5VI-E12-T0161-A, Revision 1, "Emergency Core Cooling System / Low Pressure Core Injection A Discharge Pressure High Channel A Functional for 1E12-N655A." i

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. . v For the above mentioned surveillances the inspector verified that testing was performed in accordance with procedures, that test instrumentation was calibrated, that limiting conditions for operation were met, that removal and-restoration of the affected components were accomplished, that test results conformed with technical specifications and procedure requirements and were reviewed by personnel other than the individual directing the test, and that any deficiencies identified during the testing were properly reviewed and resolved by appropriate management personne No violations or deviations were identifie . Engineered Safety Feature (ESF) Walkdown (71710)

During this inspection period, the inspector performed a detailed walkdown of train "A" and train "B" and common components of the Standby Liquid Control (SLC) System. The system walkdown was conducted using Valve Lineup Instruction (VLI)-C41, Revision Prior to conducting the walkdown, the. inspector verified VLI-C41 against controlled Piping and Instrumentation Diagrams (P& ids) for the SLC system. No significant discrepancies were identified as a result of this verificatio During the system walkdown, the inspector directly observed equipment conditions to verify that hangers and supports were made up properly; appropriate levels of cleanliness were being maintained; piping insulation, heaters, and air circulation systems were installed and operational; valves in the' system were installed in accordance with applicable P& ids and did not exhibit gross packing leakage, bent stems, missing handwheels, or 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 indication, the inspector verified that valves in the system flow path were in the correct positions as required by VLI-C41; that where required, power was available to the valves; valves required to be locked in position were locked; and, that pipe caps and blank flanges were installed as require !

No violations or deviations were identified.

, Containment Closecut Inspection (61715)

On January 26 and 27, 1988, the inspector verified the proper positioning of the following isolatinn valves associated with containment penetrations:

Outboard Inboard Test Connection (if and) Penetration (Locked and Capped) l P11-F060 P11-F545 P11-F539 P108 i

G41-F145 G41-F140 P301 1 l

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M51-F110 M51-F090 P302 G41-F10v G41-F522 G41-F528 P203 G50-F277 G50-F272 P420 G33-F034 G33-F028 P424-C11-F083 C11-F122 P204' ,

P57-F015B P57-F5248 P116 M14-F090 M14-F085 M14-F610 V314 M14-F205 V314 M14-F200 V314 017-F071A D17-F071B P201 i D17-F079A 017-F0798 P201 The inspector also observed various portions of Surveillance Instruction (SVI)-P53-T7312, "Upper Containment Airlock Seal Pneumatic System Leak Test," as documented in Paragraph 6. of this repor 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 LER 87049-LL Deenergization of Reactor Protection System Bus During Performance of a Surveillance Instruction Results in Unexpected Shutdown Cooling Isolation Valve Closure i

LER 87050-LL Intermediate Range Monitor Electrical Noise Results in Reactor Protection System Actuation LER 87053-LL Control Room Rounds Log Deficiency Results in Technical Specification Violation Due to Surveillance Requirements Not Being Performed LER 87060 LL Personnel Errors Result in Inoperable Effluent Radiation

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Monitor and Technical Specification Violation LER 87062-LL Containment Isolation Valve Energized in Viclation of Technical Specification LER 87063-LL Reactor Core Isolation Cooling System Isolation Upon Returning Leak Detection Bypass Switch to NORMAL due to Indeterminate Cause

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LER 87064-LL Feedwater Pump Overspeed Caused by a Deficient Periodic Test Instruction Resulted in'a Reactor Scram, High Pressure Core Spray Injection and an Unusual Event LER 87065-LL Bad Connection in Isolation Logic Results in Unexpected Isolation of a Main Steam Drain Line Inboard Isolation Valve During Surveillance Testing

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LER 88002-L Reactor Scram Results From Intermediate Range Neutron Monitors Upscale Trip Due to Excessive Feedwater Flow with Manual Control of a Turbine Driven Feedwater Pump LER 88003-LL Failure to Declare Radiation Monitor Inoperable Results in Violation of Technical Specification Action Statement Requirements LER 88005-LL Personnel Error During Reactor Protection System Bus Power Supply Transfer Results In An Residual Heat Removal Shutdown Cooling Isolation LER 88008-LL Equipment Testability Deficiency Results in Partia .

Nuclear Steam Supply Shutoff System Balance of Plant Isolation LER 87053 reported failures to perform channel checks for the control room radiation monitor and source range neutron monitors as required by  !

technical specifications 4.3.7.1 and 4.3.7.6.1, on July 13 and 14, 1987, while in cold shutdown. Upon discovery of the missed channel checks, the  ;

channel checks were performed satisfactorily indicating that the '

instruments were capable of performing their intended function )

Licensee investigation disclosed that the cause for the missing channel ,

checks was due to operator misunderstanding of a plant rounds instruction  !

utilized to perform technical specification rounds, including subject l channel checks while in cold shutdown or refueling. A new revision of

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the plant rounds instruction was being used for the first time on

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July 13, 1987. The instruction had been revised to provide far the l documentation of an entire week of daily readings on a single lo Footnotes which were intended to clarify when readings were required were unclear. The footnotes were interpreted by operators to mean that the control room radiation monitor ano. source range neutron monitor readings were not require As a result of this occurrence, the plant rounds instructions for cold shutdown and refueling were revised to avoid further confusio Additionally, the plant rounds instructions for Operational Conditions 1,

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2, and 3 were also reviewed and revised to avoid similar confusio !

Operators involved with this event were counseled on the need for greater '

attention to detail and to resolve procedural ambiguitics prior to the performance of procedural actions. Failure to perform channel checks of the control room radiation monitoc and source range neutron monitors at i the intervals specified by technical specifications 4.3,7.1 and 4.3.7.6.1, respectively, is a vioDation (440/88003-04(DRP)). This violation meets the tests of 10 CFR 2. Appendix C,Section V.a.

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consequently, no notice of violation will be issued and this matter is considered close I LER 87060 documented an event on August 21, 1987 in which the Turbine Building / Heater Bay vent radiation monitor was rendered inoperable due to improper implementation of a tagout for the Turbine Building / Heater Bay vent isokinetic sample pump. Instead of tagging out the isokinetic sample pump, operators tagged out and secured the Turbine Building / Heater Bay vent radiation monitor. Upon securing the Turbine Building / Heater '

Bay vent radiation monitor, the control room received an alarm on the airborne radiation monitoring system panel labeled "TB/HB Vent Air Rad Mon Flow Low." Control room operators cleared the annunciator but did not respond as required by the associated alarm response instructio Cperators mistakenly identified the annunciator as being associated with the expected isokinetic sample pump out-of-service condition. Had the tagout been properly implemented, the control room would have received an annunciator on the airborne radiation monitor panel labeled "TB/HB Vent Gas Sample / Stack Flow Low." As a result of the operator's inability to

distinguish between the two annunciators, the inoperability of the t

Turbire Building / Heater Bay vent radiation monitor went unrecognized for approximately 19 hours2.199074e-4 days <br />0.00528 hours <br />3.141534e-5 weeks <br />7.2295e-6 months <br /> without the actions required by technical '

specification 3.3.7.10 being performe Required actions which were not performed included: the acquisition and

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, analysis of noble gas grab samples; the continuous collection of particulate and iodine samples with alternate sampling equipment within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of monitor inoperability; and, the performance of sample flow estimates at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The licensee noted in this LER that a similar event occurred on April 13, 1986, which was reported via LER 86004. In that event, the Unit 1 plant vent radiation monitor was rendered inoperable by a tagout for the Unit 2 plant vent radiation monitor due to the opening of a shared power supply circuit breaker. In that event, operators similarly failed to recognize control room annunciation which signaled the inoperable condition of the Unit 1 plant l vent radiation monitor and did not respond as required. Corrective ,

actions fcr that event included personnel training on the airborne radiation ;

monitoring system design operating procedures and proper response to plant annunciators. As a result of that event, a Notice of Violation was issued (440/86011-02(DRP)). Failure to perform actions required by  !

technical specification 3.3.7.10 with the Turbine Building / Heater Bay i vent radiation monitor inoperable is, in part, repetitive of violation (440/86011-02(DRP)) and is a violation (440/88003-05(DRP)).

LER 87062 was submitted to report an event which occurred on August 28, 1987 in which the residual heat removal (RHR) head spray containment

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isolation valve 1E12-F023 was ener0ized for a period of approximately 48

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hours over which time the valve was inoperable pending the performance of required retests following modifications to the valve control circuitr Licensee investigation into this event determined the root cause to be personnel error. The Unit Supervisor authorizing clearance of the tagout was not aware of the impact that reenergizing the valve had on technical specification requirements for containment integrity. Technical specifications required the valve to be closed and deenergized until r completion of valve stroke and isolation time verification. Throughout i

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tne event, the plant was in Operational Condition Throughout the event, an isolation signal was present to prevent the  ;

valve from opening. By design, the isolation signal is present whenever  :

reactor pressure is greater than 135 psig to protect the RHR system from  :

an overpressure condition. The operator involved with this event was counseled regarding the need for greater attention to detail when reviewing work packages for impact on technical specification i requirements. The head spray containment isolation valve was satisfactorily retested on September 23, 1987. Failure to maintain the 1 residual heat removal head spray containment isolation valve deenergized .

while inoperable is contrary to technical specification 3.6.4 and is a '

violation (440/88003-06(DRP)). This violation ~ meets the tests of 10 CFR

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2, Appendix C,Section V.a, consequently, no notice of violation will be  :

) issued and this matter is considered close LER 88003 reported an event involving failures to take technical specification required actions for an inoperable emergency service water

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loop A discharge radiation monitor. Inspector review and inspector l j- findings relative to this event are discussed in paragraph 3b of this '

s repor ;

LER 88005 describes an event involving a residual heat removal (RHR)  ;

system shutdown cooling isolation during restoration of plant equipment ,

following a transfer of reactor protection system (RPS) bus A from its *

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normal power supply to its alternate suppl Tripping of the isolation

logic during the break-before-make bus transfer was anticipate Operators were directed to rack out the motor supply circuit breakers for l the shutdown cooling isolation valves to provent a loss of shutdown, cooling. Following the RPS bus transfer, operators were directed to ,

rack in the circuit breakers prior to having reset one of- two trains of

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logic associated with the high reactor pressure RHR isolation functio As a result, the isolation valves closed immediately upon racking in ,

their respective motor supply ciruit breaker The event was reported as required pursuant to 10 CFR 50.72(b)(2)(ii) as an actuation of an engineered safety feature that was not part of a preplanned sequenc Subsequent licensee evaluations of the event determined that it was not i reportable pursuant to 10 CFR 50.72 and 50.73 as an unplanned engineered 1 i safety feature actuatio In arriving at this determination, the j licensee reasoned that, since the RPS bus transfer was itself a ,

preplanned evolution involving procedures which acknowledged the tripping j

of the shutdown isolation logic as an expected response, the shutdown cooling isolation was part of a preplanned sequence. The inspector sharply disagreed with this determination based upon the overwhelming evidence that operating personnel controlling the bus transfer evolution neither planned nor expected to actuate the shutdown cooling isolation function. To the contrary, operating personnel controlling the evolution took affirmative steps to prevent such an isolation but were unsuccessful.

j Following discussions with the licensee regarding this matter, the

licensee issued this LER. While the LER provided a complete description

! of the event, root causes, and corrective acttons to prevent recurrence, l

) the licensee maintained that this event was ;,ot reportable pursuant to 10 l l

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CFR 50.72 or 10 CFR 50.7 The licensee specified in the LER that it was i submitted as a voluntary report. In light of the confusion displayed by

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the licensee regarding the notion of an event being part of a preplanned  !

sequence, the inspector will examine the handling of future engineered safety feature actuations to determine whether or not the licensee's evaluation of this event represents an adverse change in reporting  ;

philosophy and to verify compliance with 10 CFR 50.73. This matter is an unresolved item (440/88003-07(DRP)). l f

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1 Allegation Followup

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(Closed) Allegation RIII 87-A-0112: "Feedwater pumps for the reactor were

leaking and CEI was not going to fix these feedwater pumps until the i plant was to be scheduled for down time."

The inspector. interviewed licensee personnel and obtained a summary 4 listing of work orders generated by the licensee's computer-based Perry Plant Maintenance Informat'on System for work orders involving leaks on the feedwater system including the reactor feedwater pumps and feedwater -

pump turbines. The interviews and work order summary reviews were conducted to ascertain whether or not safety significant leakage had occurred on the feedwater pumps or feedwater pump turbines without timely and appropriate licensee corrective actio ;

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The inspector determined that between January,1987 and August 1987, the time this allegation was received, a number of leaks had been identified on or around the feedwater pumps, feedwater pump turbines, seal water 3 supply piping, and adjacent small bore instrument vent and drain pipin "

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Leaks directly associated with the feedwater pumps and feedwater pump turbines generally consisted of shaft seal leakage. Seal water i system leakage was generally confined to threaded or flanged joints. In i accordance with the licensee's administrative controls for corrective maintenance, work orders for these leaks were evaluated for impact on i equipment operability and reliab?llty and were subject to prioritization l and scheduling based upon the results of these evaluations. A number of the leaks were evaluated and determined not to have immediate impact on .

feedwater pump operability and moreover, plant safety. As a result, ,

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plant operation continued with the existing leakage. Such leakage was 4 collected in the floor drain system and transferred tr> liquid ,

I radioactive waste treatment facilities for processin l l Given the lack of specificity contained in this allegation, the inspector j was unable to identify the particular leak or leaks of concer The inspector was able to substantiete that instances of feedwater leakage on l i or around the feedwater pumps had occurred during plant operatio Based l

, upon the nature of the leaks identified during the inspector's review, l however, the inspector was unable to substantiate that safety significant leakage on the feedwater pumps occurred kithout timely and appropriate corrective actions by the license The faspector noted that subsequent to the riceipt of this allegation in October 1987, the licensee

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identified a po%ntially significaut leak on the B main feedwater pump j casin ' nwing

. the Identification of this leak, the B feedwater pump

was ret servi e and twained out of service during plant i

opera' '

lanucr/ 190 During a planned outage in January l

1988, 'a punp wat r mlaced in its entirety with an identical j 19

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feedwater pump from Unit This allegation is considered close .

t 11. Onsite Review Committee,(40700)

The inspectors reviewed the minutes of the Plant Operations Review Committee (PORC) meetings No.87-221 through 87-260 and 88-018 through 88-021, conducted prior to and during the inspection period to verify conformance with PNPP procedures and regulatory requirements. These  !

observations and examinations included PORC membership, quorum at PORC.

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meetings, and PORC activitie No violations or deviations were-identifie . Physical Security Procedures For The Resident Inspector (71881)

During this inspection period, the inspectors observed / reviewed selected licensee activities for conformance with the approved physical security ,

plan. The inspectors reviewed security personnel staffing levels and verified that individuals authorized by the physical security plan to direct security activities were provided for each shif The inspectors observed that access control measures, including search equipment, protected area and vital area barriers, and security door locking devices were operational and in use. The inspectors observed that personnel and packages entering the protected area were properly searched in accordance with licensee procedures. The inspectors observed that persons granted access to the site were badged to indicate whether or not they had unescorted or escorted access authorizaior,. Finally, by direct observation the inspectors determined that the effectiveness of detection '

assessment aids was maintained by the absence of obstructions in the isolation zone, adequate illumination of the protected area and protected '

area barrier, and operable video surveillance equipmen i No violations or deviations were identifie ;

1 Radiological Protection Procedures For The Resident Inspector (71709) '

Through discussions with licensee management, supervisory, and health physics personnel, and observation of licensee work planning activities, the inspectors determined that licensee personnel were aware of the ALARA

, program and that ALARA considerations were routinely considered in the 1 planning of activities which involved occupational radiation exposur The inspectors also determined through monthly Plant Status Meetings such

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as the one described in Paragraph 14. of this report and review of the .

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licensee's internally generated Monthly Performance Reports, that the status of meeting ALARA goals and objectives is periodically assessed and ,

disseminated to affected plant personne l

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During the course of routine inspection activities conducted during this ,

inspection period, the inspectors accessed plant areas requiring a '

radiation work permit (RWP). The inspectors reviewed the radiation work permits and verified that, in accordance with licensee procedures, the '

. RWPs contained a description of activities authorized, radiation levels, P

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contamination levels, protective clothing requirements, dosimetry requirements, health physics coverage requirements, expiration dates, and required review and approval signatures. The RWPs were determined to be current and readily available for employee review. Work activities observed by the inspectors were conducted in accordance with RWP requirement Inspector observation of personnel within the radiologically controlled area determined that personnel monitoring equipment was properly utilized and that dosimeter readings were recorded as required upon leaving the radiologically controlled area. Personnel exiting the radiologically controlled area were observed to properly utilize personal contamination monitors. Posting of radiation areas, contaminated areas, and labeling of containers holding radioactive material was determined to be in conformance with NRC regulations and licensee procedure No violations or deviations were identifie . Plant Status Meetings (30702)

On February 4,1988, at the Perry site and on February 19, 1988, at the NRC Regional Offices, NRC management met with CEI management to discuss the current status of the plant, recent events, and licensee initiatives to improve the quality of plant operating and maintenance activitie These meetings are being held on a periodic (initially monthly) basi . Violations For Which A "Notice of Violation" Will Not Be Issued The NRC uses the Notice of Violation as a standard method for formalizing the existence of a violation of a legally binding requiremen However, because the NRC wants to encourage and support licensee's initiatives for self-identification and correction of problems, the NRC will not generally issue a Notice of Violation for a violation that meets the tests of 10 CFR 2, Aopendix C,Section V.A. These tests are: 1) the violation was identified by the licensee; 2) the violation would be categorized as Severity Level IV or V; 3) the violation was reported to the NRC, if required; 4) the violation will be corrected, including measures to prevent recurrence, within a reasonable time period; and 5)

it was not a violation that could reasonably be expected to have been prevented by the licensee's corrective action for a previous violatio Violations of regulatory requirements identified during the inspection for which a Notice of Violation will not be issued are discussed in Paragraph . 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 bot An open inspection item disclosed during the inspection is discussed in Paragraph .

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1 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 deviatio An unresolved item is identified in Paragraph . Exit Interviews (30703)

The inspectors met with the licensee representatives denoted in Paragraph 1 throughout the inspection period and on February 22, 1988. The inspector summarized the scope and results of the inspection and discussed the likely content of the inspection repor The licensee did not indicate that any of the information disclosed during the inspection could be considered proprietary in natur :

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