IR 05000416/1987001

Insp Rept 50-416/87-01 on 870117-0213.No Violations or Deviations Noted.Major Areas Inspected:Licensee Action on Previous Enforcement Matters,Operational Safety Verification,Maint Observation & Surveillance Observation

ML20207T689
Person / Time
Site:	Grand Gulf
Issue date:	03/09/1987
From:	Butcher R, Dance H, Will Smith NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20207T672	List: IR 05000416/1987001 ML20207T671
References
50-416-87-01, 50-416-87-1, NUDOCS 8703240247
Download: ML20207T689 (14)

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UNITED STATES

[>3 Cfr o , NUCLEAR REGULATORY COMMISSION

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Report No.: 50-416/87-01 Licensee: System Energy Resources, In Jackson, MS 39205 Docket No.: 50-416 License No.: NPF-29 Facility Name: Grand Gulf Inspection Conducted: January 17 - February 13, 1987 Inspectors: cm 79 7 Rt . Butcher, Senior Resident Inspector Date Signed j 9 7 W. F. Smith, Resident ifispector - Date Signed Approved by: A' >

H. C. Dance,' Section Chief Y I7 Date Signed

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Division of Reactor Projects SUMMARY

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Scope: This routine, inspection was conducted by the resident inspectors at the site in the areas of Licensee Action on. Previous Enforcement Matters, Operational Safety Verification, Maintenance Observation, Surveillance Observa-tion, ESF System Walkdown, Reportable Occurrences, Operating Reactor Events, Inspector Followup and Unresolved Items, and Maintenance Program Implementa-tio Results: No violations or deviations were identified.

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REPORT DETAILS Licensee Employees Contacted J. E. Cross, GGNS Site Director

• C, R. Hutchinson, GGNS General Manager

• R. F. Rogers, Manager, Unit 1 Projects

• A. S. McCurdy, Manager, Plant Operations

• J. D. Bailey, Compliance Coordinator

• J. Wright, Manager, Plant Superintendent L. F. Daughtery, Compliance Superintendent D. G. Cupstid, Start-up Supervisor R. H. McAnulty, Electrical Superintendent

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• J. P. Dimmette, Manager, Plant Maintenance W. P. Harris, Compliance Coordinator J. L. Robertson, Licensing Superintendent L. G. Temple, I&C Superintendent J. H. Mueller, Mechanical Superintendent L. B. Moulder, Operations Superintendent J. V. Parrish, Chemistry / Radiation Control Superintendent Other licensee employees contacted included technicians, operators, security force members, and office personne * Attended exit interview Exit Interview The inspection scope and findings were summarized on February 13, 1986, with those persons indicated in paragraph 1 above. The licensee did not identify as proprietary any of the materials provided to or reviewed by the inspectors during this inspection. The licensee had no comment on the following inspection findings:

416/87-01-01, Inspector Followup Item. Correction of human factors concerns in the control room (Paragraph 5).

416/87-01-02, Inspector Followup Ite Followup of actions taken by the licensee to ensure seismic category I electrical panel doors are maintained closed (Paragraph 5).

416/87-01-03, Inspector Followup Item. Resolution of valve and switch labeling problems caused by identification of power source with some motor operated valve numbers (Paragraph 5).

416/87-01-04, Unresolved Ite Determination of seismic qualifica-tion status of Control Rod Hydraulic Unit 36-13 due to missing mounting bolt (Paragraph 5).

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416/87-01-05, Inspector Followup Ite Correction of deficiencies found during LPCI "C" walkdown (Paragraph 8).

416/87-01-06, Inspector Followup Ite Permanent corrective actions to prevent SRV lifts during trip unit calibration checks (Paragraph 10).

416/87-01-07, Inspector Followup Ite Determination of cause and appropriate corrective action related to inadvertent Control Room Fresh Air Unit start in the isolation mode (Paragraph 10). Licensee Action on Previous Enforcement Matters (92702)

(Closed) Violation 416/84-16-0 Failure to audit to the depth required to determine that sufficient qualifications existed in Instrument and Control technicians audit. The inspectors verified the corrective actions committed by the licensee's response to the violation, AECM-84/0364 dated July 13, 1984, were in fact implemented and that documentation existed for the record. The results of the inspection were satisfactor (Closed) Deviation 416/84-16-07. Nuclear Plant Engineering (NPE) failed to implement a previous commitment made in response to violation 416/

82-55-03, which was to revise their implementing procedures to assure that the Plant Safety Review Committee (PSRC) reviewed all 10 CFR 50.59 Safety Evaluations performed by NPE. The inspectors verified that NPE Procedure 01-136,10 CFR 50.59 Safety Evaluations, was issued on December 14, 1984 requiring safety evaluations to be transmitted to the PSRC. In addition, the inspectors reviewed documentation showing that those safety evalua-tions not reviewed by the PSRC were subsequently reviewed. Since the December 1984 issuance of NPE Procedure 01-136, the licensee has made changes and additions to the Administrative Manuals to continue improving controls in this area. Unresolved Items *

One new unresolved item identified during this inspection is discussed in paragraph 5. Operational Safety Verification (71707)

The inspectors kept themselves informed on a daily basis of the overall plant status and any significant safety matters related to plant opera-tions. Daily discussions were held with plant management and various members of the plant operating staff.

• An Unresolved Item is a matter about which more information is required to determine whether it is acceptable or may involve a violation or deviatio .

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The inspectors made frequent visits to the control room such that it was visited at least daily when an inspector was onsite. Observations included instrument readings, setpoints and recordings, status of operating systems, tags and clearances on equipment controls and switches, annunciator alarms, adherence to limiting conditions for operation,

-temporary alterations in affect, daily journals and data sheet entries; control room manning, and access controls. This inspection activity included numerous informal discussions with operators and their super-visor the inspectors were onsite, selected Engineered Safety Weekly, Feature (whenESF) systems were confirmed operabl The confirmation is made by verifying the following: Accessible valve flow path alignment, power supply breaker and fuse status, major component leakage, lubrication, cooling and general condition, and instrumentatio General plant tours were conducted on at least a biweekly basis. Portions of the control building, turbine building, auxiliary building and outside areas were visited. Observations included safety related tagout verifica-tions, shift turnover, sampling program, housekeeping and general plant conditions, fire protection equipment, control of activities in progress, radiation protection controls, physical security, problem identification systems, and containment isolation. The following comments were noted:

there were some control room instruments that offer less than optimum readability from a human factor standpoint. The inspectors expressed concern to the licensee that such instruments can have a negative affect during an emergency by adding unnecessary burdens on the operators because they must go through a conscious thought process over how to read the instrument before they can determine what the instrument is indicatin Examples of some of these instruments are listed belo Some instruments are graduated in unusual divisions, for example, suppression pool temperature instrument M71-R605 shows 70,110,150,190, and 230 degrees F, in 4 degrees subdivisions. Wide range suppression pool level instrument E30-R600 shows 10.5,13,15.5,18, 205, and 23 feet, in 0.15 foot subdivisions. The two diesel generator (DG) field ammeters are different, as well as unusual. DG 11 meter shows 0, 128, 256, 384 and 512 amperes in 12.8 ampere subdivisions while the DG 12 meter shows 0,120, 240, 360, and 480 amperes in 12 ampers subdivisions. The Reactor Core Isolation Cooling pump suction pressure meter shows 0, 15, 30, 45, 60, and 75 psig in 7.5 psig subdivisions with 85 psig at the top of the scale.

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The Standby Liquid Control System storage tank level meter shows 0,1040, 2080, 3120, 4160 and 3200 gallons in 104 gallon subdivisions.

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The wide range reactor vessel level and pressure recorder on the

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operator's console (C34-R614 on P680-28) and the same instrument on the ECCS console (B21-R623B on P601-178) use reverse ink colors. If an i

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operator moves from the P680 panel to the P601 panel, he will find that the red line he was watching for reactor pressure becomes a blue line, and the blue line he was monitoring for reactor level becomes a red lin This can be distracting during an emergenc The High Pressure Core Spray diesel generator (DG 13) fuel nil day tank and storage tank level indicators are located on panel P870-58 about 10 feet away from the DG control console. DG 11 and DG 12 fuel oil day and storage tank level indicators are on their own respective control console On panel P870 the indicating lights for the lower containment personnel airlock being secure /not secure are green / red respectivel The inspectors noted that both the green and the red indicator lights stayed on. The operators initiated a "do not use" sticker (MWO 70493) after the inspectors questioned the operability of the indicator. Erroneous indica-tions should always be identified, even though the operators may already understand that they are erroneou The licensee is currently evaluating many human factors aspects of the control room. The inspectors will follow the program to ensure that concerns such as illustrated by the above examples are being appropriate addressed. This is Inspector Followup Item 416/87-01-0 During a plant tour on January 29, 1987, with Brookhaven National Labora-tory personnel involved in the NRC's Probabilistic Risk Assessment (PPA)

application program, it was noticed that not all the retaining screws in the electrical switchgear breaker panel doors were secured. Typically, one screw was holding the door shut and the other screws were either hanging loose or missing. The licensee evaluated the effect of the loose doors on the operability of the related safety equipment, and concluded that no impact on seismic qualification can be expected due to loose or missing fasteners unless one of the following conditions exists: all fasteners are loose or missing, only one fastener is tightly engaged near a door corner, or not enough fasteners are engaged to resist seismically induced door inertia force Any of the above conditions, according to the licensee, may allow banging i of the doors during a seismic event which could possibly introduce i structural non-linearities not considered in the original seismic evalua-tion. The licensee initiated actions to ensure that all door fasteners on

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safety related electrical panels are fully tightened and issued Mainten-ance Work Orders (MW0s) to repair or replace damaged or missing door fasteners. On subsequent tours the inspectors identified that this i

condition existed in the Diesel General room in that some control panel doors were not secured. The licensee was made aware of this condition and

j corrected the problem. The resident inspectors will review the results of

these actions on future tours. This shall be tracked by Inspector

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Followup Item 416/87-01-02 until the inspectors consider the problem of loose doors corrected.

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On January 20, 1987, while touring the Control Room, the inspectors found the back Control Room door open with a security guard posted at the door to control access. The door handle had been broken, so the door could not be opened from outside the Control Room. There were no Limiting Condi-tions for Operation (LCOs) in effect for the door being in the open position except for a fire protection LC0 that had been in effect for other reason Discussions with the Shift Supervisor (SS) revealed that open as cause for entering an the licensee additional LC did not considerSpecification Technical the door being(TS) 3/4.7.2, Control Room Emergency Filtration System, which would automatically isolate the Control Room upon initiation (i.e., high radiation, loss of coolant accident),

does not address the allowable opening size in the Control room envelop License Condition (LC) 2.C.(38) states that the licensee shall operate Grand Gulf Unit I with an allowable Control room leak rate no to exceed 590 cubic feet per minute (cfm). The open door would allow Control Room inleakage to exceed 590 cfm. Based on discussions with the inspectors the SS immediately initiated an LCO to administratively control the door in case of a Control Room isolation signal. Subsequently on January 22, 1987, Standing Order 87-0005 was issued stating that in case of Control i Room door problems requiring the posting of a guard, the guard will be informed of the requirement to close the door immediately upon orders from Control Room personnel. The doors are to be ordered closed upon receipt of any Control Room isolation signal. Since there does not appear to be any TS related requirement for Control Room opening size except the LC noted above, the inspectors discussed this event with Region II Manage-

ment. Region II Management agreed that if the licensee administratively

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controlled the door such that responsible personnel were made aware of the door condition and were required to take action in case of a Control Room

isolation signal, then the door could be kept open with a guard posted.

! On February 6,1986, while touring the reactor shutdown panel area in the Control Building, the inspectors noted minor discrepancies in panel labeling. The new transfer panel installed by Design Change Package 81/5003 was missing a switch label and had switch labels that did not agree with the Division 1 Reactor Shutdown Panel nor the applicable piping and instrument drawing (P&ID). Specifically, the switch labels for motor '

operated valves E51-F019 and E12-F040 appeared on the transfer panel as E51-F019A and E12-F040A respectively, and the transfer panel switch for E12-F024A did not have a labe The transfer panel labeling problems were promptly corrected by the licensee, but were similar to those identified in NRC Inspection Report 416/86-41 during the High Pressure Core Spray system walkdown. The "A" suffix on the above valves (F019A and F040A) appeared on the P&lD as "-A" thus identifying the power source rather than the valve. Adding the power

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source to the valve numbers on P& ids appears to be causing widespread labeling problems. The licensee is investigating and has committed to correct the problem where it exists. This shall be tracked as Inspector Followup Item 416/87-01-03.

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On February 6, 1987, during a tour of the containment the inspectors found that one fastener required for securing Control Rod Drive Hydraulic Control Unit 36-13 to the upper foundation wide flange beam was missin It was apparent from inspection of the misaligned holes that there never was a fastener installed at _.that location. The licensee was requested to provide the inspectors with documentation approving the installation defect and what effect this condition had on the seismic qualification of the Unit. This shall be Unresolved Item 416/87-01-0 No violation or deviations were identified. Maintenance Observation (62703)

During the report period, the inspectors observed portions of the mainten-ance activities listed below. The observations including a review of the work documents for adequately, adherence to procedure, proper tagouts, adherence to technical specifications, radiological controls, observation of all or part of the actual work and/or retesting in progress, specified retest requirements, and adherence to the appropriate quality control MWO M70285, Replace seal and troubleshoot vibration on Reactor Water Cleanup System Pump MWO 168530, Troubleshoot and repair Offgas System Hydrogen Analyzer MWO 168531, Troubleshoot and repair Offgas System Hydrogen Analyzer MWO 170400, Troubleshoot safety / relief valve high pressure and low-low set trip unit No violations or deviations were identified. Surveillance Observation (61726)

The inspectors observed the performance of portions of the surveillances listed below. The observation included a review of the procedure for technical adequacy, conformance to technical specifications, verification of test instrument calibration, observation of all or part of the actual surveillances, removal from services and return to service of the system or components affected, and review of the date for acceptability based upon the acceptance criteri IC-1C51-SA-0001, Revision 23, Average Power Range Monitor Calibra-tion (Channel B).

06-CH-IN62-V-0051, Revision 21, Off Gas Hydrogen Concentratio .

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06-1C-1821-M-1001, Revision 24, Safety / Relief Valve High Pressure Trip and Low-Low Set Relief Functional Test (Channel F).

06-ME-1M23-V-0001, Revision 25, Containment and Drywell Airlock Seal Leak Tes IC-1C11-M-0003, Revision 26, Scram Discharge Volume High Water Level Float Switches (RPS) Calibratio The inspectors noted that the surveillance procedure left it to the expertise of the technician to decide on what range to set the digital volt meter when connecting to the float switch contacts in accordance with step 5. This would cause unsatisfactory results if the technician chose the wrong scale, resulting in unnecessary repeats of the test or possibly an unwarranted switch adjustmen The procedure should be more specific. The inspectors also noted that inconsistent results can be obtained depending on whether the technician notes the level at which the contacts close, or the level after the float settles out. Due to the float displacement and switch reaction, this choice of taking data can mean the difference between satisfactory and unsatisfactory results. The data for the test above showed readings at or so close to the tolerance limit some adjustment might have been required, but was not made. Step 5.4.6 should specify how the data should be taken. This was discussed with the licensee, who agreed with the concern Prior to the end of this inspection report period, the licensee corrected the procedure by issuance of a change (TCN-2).

No violations or deviations were identified. Engineered Safety Features System Walkdown (71710)

A complete walkdown was conducted on the accessible portions of the Low Pressure Core Injection (LPCI) C System. The walkdown consisted of an inspection and verification, where possible of the required system valve alignment, including valve power available and valve locking where required, instrumentation valved in a functioning electrical and instru-mentation cabinets free from debris, loose materials, jumpers and evidence of rodents, and system free from other degrading condition The system was found to be in a satisfactory condition and appeared ready to perform its safety function if called upo The inspectors noted several minor discrepancies as discussed below:

Revision 31 of Piping and Instrument Drawing (P&lD) M-10858, zone G-6 shows a four inch pipe coming from P&ID M1085A, zone D-6, when in fact it comes from P&ID M-1085C, zone D- The P&ID Legend at GGNS shows "C" next to valves that are normally closed,

"LC" next to valves that are normally locked closed, no symbol next to valves that are normally open, and "LO" next to valves that are normally

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locked ope On P&ID M-1085C, valves E12-FX058, FX026, FX031, and FX032 are indicated as normally open when they should be normally closed as required by System Operating Instruction (501) 04-1-01-E12-1, Revision 35, Residual Heat Removal Syste The above 501, Attachment 1, page 10 requires valves E12-F245 and F281 locked closed but P&ID M-1085C show them "C" (closed).

In the above 501, Attachment I describes E12-F405 as a test connection isolation valve, when in fact it is a drai It does not isolate the test connectio Attachment 1 of the above S0I describes E12-F311 and F324 as drains when they should be the same as E12-F304, which is described as, "C RHR Test Line Test Conn."

The electrical lineup checksheet, page 9 of Attachment III in the above 501, requires the motor operated valve heater breakers to be closed; however, the inspectors found the appropriate breakers danger tagged open since September 17, 1986. This was done in accordance with a September 8, 1986 letter from Nuclear Plant Engineering to the General Plant Manager, stating that motor heaters should be tagged open because the motor heaters were not included in any environmental or seismic testing conducted by Limitorque, and the heaters have been known to cause internal motor operator wiring and limit switch rotor damage due to the heat. This was an immediate action taken by the licensee based on IE Information Notice 86-71, Recent Identified Problems with Limitorque Motor Operators. The licensee intends to issue a design change to disconnect the heaters permanentl The next revision to the 501 should reflect the required breaker positions or delete the breakers from the electrical lineu Correction of the above minor deficiencies shall be tracked under Inspector Followup Item 416/87-01-0 No violations or deviations were identified. ReportableOccurrence(90712and92700)

The below listed event reports were reviewed to determine if the informa-tion provided met the NRC reporting requirements. The determination included adequacy of event description and corrective action taken on planned, existence of potential generic problems and the relative safety significance of each event. Additional inplant reviews and discussions with plant personnel as appropriate were conducted for the reports indicated by an asterisk. The event reports were reviewed using the guidance of the general policy and procedure for NRC enforcement actions, regarding licensee identified violation The following License Event Reports (LERs) are close . _ _ - _ - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _

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I LER NO EVENT 0 ATE EVENT

*86-026-01 July 30, 1906 Control rod drift to full ou i  ;

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*86-045-01 November 18, 1986 Loss of power to Division 2 ESF !

bus due to felled tre ; *86-046 December 18, 1986 Fire watches delinquent due to j personnel illnes ,

LER 86-026-01: On July 30, 1986, a single control rod continued drifting out of the core after receiving a single notch withdrawal command. LER i 86-026 was a volu t ry special report describing the licensee's correction ;

actions. The licensee conducted flush testing on the Control Rod Drive

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(CRD) system on October 31, 1986, and the results were found acceptable.

Two, three minute, high velocity flushes were conducted at two points in the CR0 system and the results measured per GE Specification 22A253 In addition, the licensee re-reviewed all GE Service Information Letters (SILs) from 1 through 360 for applicability to GGNS. Resolution has been  !

obtained for all SILs except as follows:

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SIL 351 recommended changes to the calibration procedures for the RCIC turbine control system. The licensee has committed to revise -

) the applicable procedure (07-5-53-E51 2) prior to the next scheduled !

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performance.

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i SIL 292 and 310 have been reviewed previously but the licensee has l i

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scheduled a major review of all Off-Normal Emergency Procedures  !

(ONEPs) in 1987 and has committed to review Sils 292 and 310 again at

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! The event of LER 86-045-1 was discussed in NRC Inspection Report 416/

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86-21.

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j No violations or deviations were identified, i

j 10. Operating Reactor Events (93702)

l The inspectors reviewed activities associated with the below listed l reactor events. The review included determination of cause, safety ,

j significance, performance of personnel and systems, and corrective actio !

j The inspectors examined instrument recordings, computer printouts, opera-4 tions journal entries, scram reports and had discussions with operations, i maintenance and engineering support personnel as appropriat ;

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Following startup of the slant after the first refueling outage the '

licensee discovered that t1e A turbine steam bypass valve was leaking.

i Nuclear Plant Engineering (NPE) had previously dispositioned Material

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Nonconformance Report (MNCR) 246/85 on May 9,1985, stating that indefi-nite operation with turbine bypass line A isolated was acceptable. MNCR 246/85 had been prepared due to a suspected crack in the A bypass lin Since there is a mixed core configuration following the refueling outage, NPE was requested to re-evaluate the isolation of the A bypass valv NPE's evaluation determined that the A bypass valve could be isolated and the plant would still be in an analyzed condition. The A turbine bypass line was isolated on January 21, 198 On January 23, 1987, at 8:43 a.m. the six low-low set Safety / Relief Valves (SRVs) momentarily lifted while the plant was at full power. The lift was for about 1.9 minutes from the time the first SRV, B21-F051D, opened until the last SRV repeated. There was about a 20 second period when all six SRVs were open. The cause was a voltage spike to the F channel of the SRV high pressure trip / low low set relief logic, which was received during performance of surveillance procedure 06 IC-1021-M-1001, Safety / Relief Valve High Pressure Trip and Low-Low Set Relief Functional Test. The voltage spike occurred while the technician was checking the gross failure trip feature of the 1821-N6688 trip unit, apparently by feedback through the power supply which is common to channels B and F. The operators responded promptly by resetting the low-low set concurrent with the technician removing the test signal, which reseated the SRVs. The reactor operator immediately ramped closed both reactor recirculation flow control valves in the fast detent mode thus reducing power by about 15 percen Plant parameters stabalized at the reduced power level about two minutes af ter the transient subsided. Reactor water level dipped to a low of 20 inches but stabilized at the normal level of 36 inches within four minutes after the incident. The plant responded satisfactorily and as expected. The inspectors reviewed recordings of key parameters indicating plant response as provided by the Emergency Response Facility Information System (ERFIS) only, because the General Electric Transient Analysis Recording System (GETARS) was out of commission for calibration at the time of the incident. As such, precise data was not available, but the data provided was adequate for a conclusion that no apparent anomalies occurred during the transient. The licensee verified there were no faults in the circuits and then revised the above surveillance procedure to require disabling the relief function (the Safety function remained operable) when doing the test in the respective division. The surveill-ance was then satisfactorily completed. The licensee is evaluating the system to determine what other corrective actions should be taken with regard to the interface between channels that allow such voltage spikes to occur. This resolution shall be tracked under Inspector Followup Item 416/87-01-0 On February 3,1987, during power operation, the A Standby Control Room Fresh Air Unit started in the isolation mode from a shutdown conditio The licensee reported that there were no Control Poom or local isolation signals (i.e., primary containment or secondary containment isolation on

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reactor vessel water level low-low Level 2, high drywell pressure, or manual actuation). This event was reported to the NRC in accordance with 10CFR50.72, paragraph (b)(2)asaninadvertentESFactuation. As of the end of this reporting period the licensee had not determined the cause of the actuation; however, it appears that the A chlorine detector might have reacted to a security officer's radio transmitter, causing the actuatio The instrument signal did not lock in, therefore the cause could not be confirme Determination of the cause and followup on appropriate corrective actions shall be tracked as inspector followup item 416/

87-01-0 No violations or deviations were identifie . Inspector Followup and Unresolved Items (92701)

(Closed) Inspector Followup Item 416/83-53-02, Component Labeling of Switches at Local Panel 1H22-P175 was Found Incorrect. The inspectors interviewed operators as to the accuracy of feedwater control system panel 1H22-P175 and the operators stated no problems have been experienced recently. The inspectors compared drawing 9645-J-201.1-N1H22-P175-1,1-1,

Revision K, to the existing panel and found no discrepancie (Closed)InspectorFollowupItem 85-39-03. Failure of electrical terminal strips in the drywell and containment hydrogen monitors. The licensee had '

issued Material Nonconformance Report (MNCR) 00457-84 to document this problem. Licensee Event Report (LER)85-043 documents the licensee's action The temperature control switch setpoints were reset to control at 285* F rather than 300* F. Based on monitoring results with the reduced temperature, no further action is necessary. LER 85-043 was closed in Report 416/86-3 (Closed) Inspector Followup Item 416/86-36-02. Adequacy of Plant Service Water (PSW) Flow to Containment Penetrations. This item was addressed in i Inspection Report 416/86-39 where it was noted that flow based on flow

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measurement appears to be adequate, but this item would remain open until i the licensee completed a special temperature monitoring test to confirm adequate cooling. The test (MWO 168509) was completed on Janaury 19,

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1986. The inspectors reviewed the date which cnnfirmed adequate cooling.

l (Closed) Unresolved item 416/84-16-06. Concern that the licensee's

Nuclear Plant Engineering (NPE) department may not consider FSAR and TS requirements when evaluating deficiencies in plant equipmen The inspectors reviewed current procedural requirements and noted that the NPE Material Nonconformance Report (MNCR) Disposition Form implemented by NPE

Administrative Procedure 01-801, Processing MNCRs, requires a 10 CFR 50.59 safety evaluation applicability review. The FSAR and TS must be consulted in order to perform this review, thus resolving the concer .

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(Closed) P2184-0 Westinghouse Type SA-1 Relay Failure. The inspectors reviewed the licensee's submittals and determined that the licensee did not submit a part 21 report on this item for Unit This item was reported as PRD-84/11 for Unit 2 only. No further action is require (Closed)InspectorFollowupItem 85-03-04. Loss of Reactor Core Isolation Cooling (RCIC) Due to a Spurious High Steam Flow Signal. This item tracked the licensee's action to correct the cause which was to modify the RCIC high steam flow differential pressure detector piping to prevent air from being trapped in the piping and condensing pot. The inspectors have since noted that Design Change Package (DCP) 83/0591-1 was implemented to relocate a vent valve which appeared to be the source of air in the piping. In addition, the licensee adjusted the detector internal damping potentiometer to inhibit spurious signals. The water trap was not removed because of man-rem considerations, however, the licensee has determined through monitoring of the detector output fcr the past year that the problem no longer exists to the extent that any further action is required. The inspectors noted that the DCP was nearly two years old when this problem occurred. Upon questioning the licensee, it was explained that this DCP (among others) reflected the results of earlier system walkdowns by licensee personnel, and that there was insufficient justifi-cation to implement this DCP until the above problem occurre (Closed) Unresolved item 416/86-41-0 Determination of cause and possible system impact of overranging High Pressure Core Spray (HPCS)

instrument E22-PI-R002, condensate bypass pressure gauge. The licensee reviewed past operation of the system and the history of this instrument and concluded that the system had not been overpressurized. When used, this section of HPCS piping pressurizes rapidly. The gauge is an air-filled bourdon tube with no snubber. When the HPCS pump is started or the upstream 10 inch motor operated valve E22-F010 is opened with the pump running, the gauge overshoots due to internal water hamer. This causes the gauge needle to momentarily peg and bend. This has happened on several occasions in the past on local pressure gauges and on other fluid systems. To prevent future damage to E22-PI-R002 and other systems that need not be routinely monitored, the licensee has changed the status of the instrument isolation valve to close, and installed a prominent plastic label alerting the operator to keep the gauge isolated except when obtaining reading No violations or deviations were identifie . MaintenanceProgramimplementation(62700)

The objectives of this inspection are to determine whether the GGNS maintenance program is being implemented in accordance with regulatory requirements, and to determine the ability of the licensee to conduct an effective maintenance program on important plant equipment.

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The inspectors selected records and procedures pertaining to safety-related and non-safety-related equipment failure leading to a plant shutdown, equipment failure leading to reduced capability of a safety related system, and a recurring safety related equipment failur The inspection commenced during this inspection period and will continue though at least one subsequent period. Most of this inspection effort consisted of gathering data from the licensee and interviewing individuals involved with the maintenance activities selected. As of the end of this inspection period, there are no findings to report nor have any conclu-sions been reached.