IR 05000424/1987060

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Insp Rept 50-424/87-60 on 871008-1120.No Violations or Deviations Noted.Major Areas Inspected:Plant Operations, Radiological Controls,Fire Protection,Security,Maint,Quality Programs & Administrative Controls Affecting Quality
ML20237E288
Person / Time
Site: Vogtle Southern Nuclear icon.png
Issue date: 12/16/1987
From: Burger C, Rogge J, Schepens R, Sinkule M
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20237E264 List:
References
50-424-87-60, NUDOCS 8712280251
Download: ML20237E288 (15)
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.e UNITED STATES

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WUCLEAR REGULATORY COMMISSION

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i Report No.: 50-424/87-60 Licensee: Georgia Power Company P. O. Box 4545 Atlanta, GA 30302 Docket No.: 50-424 License No.: NPF-68 Facility Name: Vogtle 1 Inspection Conducted: October 8 - November 20, 1987 Inspectors: $v A Ibct/k WM /2-Date Signed-6[P 7 p CF. Rogge, Senior Resident Inspector kk ff b us y "12./ /G lP 7 !

Date Signed-govR.J.Schepens,ResidentInspector SWw /) babuvo V ( 2llG fil f,b "LC. W. Burge , Resi .ent Inspector Date Signed Approved by: L <4Gdf- [' b) /'

M.'V.' Sinkule, Section Chief Date/ Signed Division of Reactor Projects SUMMARY-Scope: This routine,. unannounced inspection. entailed resident inspection- in .

the following areas: plant operations, radiological' controls, maintenance, surveillance, fi re protection, security, and quality programs and '

administrative controls affecting qualit Results: No violations or deviations were identifie PDR ADOCK 87 h 24

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REPORT DETAILS Persons Contacted Licensee Employees G. Bockhold, Jr. , General Manager Nuclear Operations

  • T. V. Greene, Plant Support Manager
  • R. M. Bellamy, Plant Manager
  • E. M. Dannemiller, Technical Assistant to General Manager C. C. Echert, Technical Assistant to Plart Manager
  • J. E. Swartzwelder, Nuclear Safety & Compliance Manager
  • W. F. Kitchens, Manager Operations R. E. Lide, Engineering Support Supervisor
  • H. Varnadoe, Plant Engineering Supervisor
  • R. E. Spinnatu, ISEG Supervisor C. W. Hayes, Vogtle Quality Assurance Manager
  • G. R. Frederick, Quality Assurance Site Manager - Operations W. E. Mundy, Quality Assurance Audit Supervisor M. A. Griffis, Maintenance Superintendent
  • R. M. Odom, Plant Engineering Supervisor
  • C, L. Cross, Senior Regulatory Specialist S. F. Goff, Regulatory Specialist
  • A. L. Mosbaugh, Assistant Plant Support Manager H. M. Handfinger, Assistant Plant Support Manager F. R. Timmons, Nuclear Security Manager Other licensee employees contacted included craftsmen, technicians, supervision, engineers, operations, maintenance, chemistry, inspectors, and office personne * Attended Exit Interview Exit Interviews (30703)

The inspection scope and findings were summarized on November-20,1987, with those persons indicated in paragraph 1 abov The inspector j described the areas inspected and discussed in detail the inspection !

results. No dissenting comments were received from the licensee. The licensee did not identify as proprietary any of the materials provided to or reviewed by the inspector during this inspection. Region based NRC exit interviews were attended during the inspection period by a resident

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, > Operational Safety Verification (71707)(93702)

The plant began this inspection period.in Power Operation (Mode 1) at=100%

power until October 9 when the _ unit was tripped to complete' a portion of the startup testing program and commence a short outage. The outage proceeded without difficulty until the. number 1 reactor coolant pump motor failed. As a result of the failed motor, Unit I restart was delayed approximately seven days. .The unit entered Hot Standby -(Mode 3) on October 27. Shortly after achieving Mode 3 the residual ' heat removal crosstie valve motor operator failed and the . engineering. walkdowns identified that the re' actor vessel level instrument impingement plates-were not installe These .two problems resulted in further startup -

delays. On Oct60er 31- the unit- entered Startup (Mode 2) and- achieved Mode 1 on November The unit achieved 100% power on November 4. On November 5 the unit tripped on a turbine trip when a vibration sensor was bumped. The unit returned to Mode 1 on November 6 and achieved 100% on November 7. On November 9 the unit- performed the 10% load swing startup test. On November 11 the unit tripped from '100*. feactor. power when the wrong test panel. was used during the performance of a reactor trip breaker test. On November 12 the unit returned.to Mode 1 and~ achieved 90% powe From November 12 througn 17 the unit experienced secondary water chemistry problems which limited power and required the plugging of condenser tube On November 18 the unit was held at 98% power while engineering concerns -

in regard to exceeding the 3411 MWT limit were resolved. On November 19 the unit achieved 100% power. The plant experienced.three ESF actuations; the Control Room Emergency Ventilation System on ' October 26 when a technician improperly reset' the radiation monitors and on November 17 when RE-12116 spiked high, an auxiliary feedwater actuation on November 5 when an operator shut the discharge valve of the running condensate pump'due to improper labeling, and a Containment Ventilation Isolation from RE-2565 on November 9 when the check source did not fully retract. A Notice of-Unusual- Event was reported on November ~ 17- when power was . lost - to meteorological instrument Control Room Activities Control Room tours and observations were performed to verify that facility operations were' being safely conducted within regulator requirements. These inspections consisted of- one. or more . of the following attributes as appropriate at the time of the inspectio Proper Control Room staffing

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Control Room access and operator behavior

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Adherence to approved procedures for activities in progress o

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Adherence to Technical Specification (TS) Limiting Conditions for Operations (LCO)

- Observance of instruments and recorder traces of safety related and important to safety systems for abnormalities

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Review of annunciators alarmed and action in progress to correct

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Control Board walkdowns

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Safety parameter display and the plant safety monitoring system-operability status

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Discussions and interviews with the On-Shift Operations Supervisor, Shift Supervisor, Reactor Operators, and the Shift Technical Advisor to determine the plant status, plans and assess operator knowledge

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Review of the operator logs, unit log and shift turnover sheets No violations or deviations were identifie Facility Activities Facility tours and observations were performed to assess the effectiveness of the administrative controls established by direct observation of plant activities, interviews and discussions with-licensee personnel, independent verification of safety systems status and LCO's, licensee meetings and facility record During these inspections the following objectives were achieved:

(1) Safety System Status (71710) - Confirmation of system operability was obtained by verification that flowpath valve alignment, control and power supply alignments, component conditions, and support systems for the accessible portions of the ESF trains were proper. The inaccessible portions are confi rmed as availability permit Additional in-depth inspection of the Auxiliary Feedwater System was performed to review the system lineup procedure with the plant drawings and as-built configurations, compare valve remote and local indications, and walkdown of hangers, supports, snubbers and electrical equipment interiors. The inspector verified that the lineup was in accordance with license requirements for system operabilit (2) Plant Housekeeping Conditions - Storage of material and components and cleanliness conditions of various areas throughout the facility were observed to determine whether safety and/or fire hazards existe (3) Fire Protectior. - Fire protection activities, staffing and equipment were observed to verify that fire brigade staffing was appropriate and that fire alarms, extinguishing equipment, actuating controls, fire fighting equipment, emergency equipment, and fire barriers were operabl .-

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(4) Radiation Protection (71709) - Radiation protection activities, staffing and equipment were observed to verify proper program implementation. The inspection included review of the , plant program effectiveness. Radiation work permits and personnel-compliance were reviewed during the daily -plant tour Radiation Control Areas (RCAs) were observed to verify proper identification and implementatio (5) Security (71881). - Security controls were observed to verify -

that security barriers were intact', guard forces were on duty, and access to the Protected Area .(PA) was controlled in accordance with the facility security pla Personnel within the PA were observed to verify proper display of badges and that personnel requiring escort. were properly escorted. Personnel-within vital areas were observed to ensure proper authorization for . the area. Equipment operability' and ' proper. compensatory activities were verified on a periodic basi (6) Surveillance (61726)(61700) - Surveillance t'ests were. observed to verify that approved procedures.were being 'used; qualified.'

personnel were conducting the tests; tests were adequate to verify equipment operability; calibrated equipment was utilized;

- and TS requirements were followed. .The inspectors observed portions of the followi_ng surveillance and reviewed completed data against acceptance criteriat Date Sury. N Dep Title 11/3/87 14915-1 Ops QPTR Special Condition Surveil. lance Log 11/4/87 14915-1 Ops Control Rod Insertion._ Limits Special Condition Sury. Log 11/4/87' 14205-1 Ops Plant Emergency Signal; Weekly Operability Test 11/4/87 14805-101 Op Quarterly,. Train B RHR Pump &

Check Valve Inservice Test-11/6/87 14808-10 Ops Quarterly, Train B CCP &

Check Valve Inservice Test '

11/19/87 14030-1 Ops Power Range Calorimetric ,

Channel. Calibration '

11/20/87 14825-108 Ops Quarterly, Train A AFW Valve Inservice Test (7) Maintenance Activities (62703) .- The inspector observed 1 maintenance activities to verify;'that correct equipment clearances were in effect; work requests and' fire prevention work permits, as required, were issued and being followed; quality control personnel were available for P inspection :

activities as required; retesting and? return .'of systems to

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service was prompt and correct; TS requirements were beig followe The maintenance backlog pts reviewep' and noted as

+ consisting of approximately 2,100' hh0's (i.e.. ,' both corrective and preventive) prior to the outdge. Raintenartce had scheduled 249 maintenance work orders to be worked during the outag During the outage the inspector observed that maintenance had actually performed an additional 151 MWO's due to discovery items and 109 MWO's due to the forced outage on the reactor coolant pump motor in addition to the 249 MWO's plarned for a

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t total of 509 MWO's. At the completion'of the outage the outage backlog had been reduced from 506 to fv0 MWO's, however the total MWO backlog had increased slightly from 2,100 to 2,178 MWO' The inspector either observed maintenance activities or r' viewed completed maintenance work puc.kcges for the following

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maintenance activities:

MWO N Dep Work Descr ion 1-87-02793 Elect. Main Perform 1 NATS Procedure and S DCPVIED[J7

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1-87-05326 Elect. Main Invesdqate Problem With Open I dicatfor Light Not Vorking s

1-87-08736 Mech. Main Implerdl1t Design Chaige Package To ,

Pressurizer Level TWrhitntter

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1-87-11815 Maint./ Che Condenser Waterbox B West Tube Leak Check & Plugging F

(8) Outage Activities (71711) - The inspector observed portions of the outage activities to determine management effectiveness in conducting outages. While this was not a refueling outage it did demonstrate the licensee's ability to schedule, prepare, and execute the pla As noted above, at the completion of the

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outage the outage backlog had been reduced from 506 to 300 MWO' During the course of the outage teamwork was evident in *

surfacing new problems and achieving resolution to prevent a new i critical path from develop %g. The planned critical path work J s-

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involving the removal of the temporary steam strainers was achieved ahead of schedulh. The outage work inside containment

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wasperformedwithfewdiffibuities. Two major items did occur -

which hac severe sch'edule impact and resulted in a seven day I restart delay. These items were the motor replacecient on the number one reactor coolant pump and the failed motor on the RHR crosstie valve HV-8716B. Teamwork in resolving both problems resulted in a very coordirsted repair effort. Unit recovery was delayed upon discovery thn. the impingement plates for RVLIS were not installed nor locatable, which required new pieces to be fabricate [

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No violations or deviations pere identifie '

4. Review of Licensee Reports (90712)(90713)(92700)

' In-Off ce R2 view of Periodic and Special Reports ,

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This inspection consists of reviewing the telow Tisivj reports to determine whether the information reported by the licensee is technically adequate and consistent with the inspector knowledge of the material contained w1Ahin the report. Selecteo material within the report is questipedtrandomly to verify accuracy to provide a-reasonable assurance that pther NRC personnel have an appropriate

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document for their activitie Monthly Operating Reports - The report dated October 8, 1987 was ceviewed. The inspector had no significant comments regarding these report Licensee Event Reports (LER's) and Deficiency Cards (DC's)

Licensee Event Reports (LER's) and Deficiency Cards (DC's) were reviewed for potential generic impact, to detect trends, and to determine whether corrective actions appeared appropriate. Events which were reported pursuant to 10 CFR 50.72, were reviewed as they occurred to determine if the technical- specifications and other regulatory requirements were satisfied. In-office review of LER's may result in further followup to verify that the stated corrective actions have been completed, or to identify violations in addition to those described in the LER. Each LER is reviewed for enforcement action in accordance with 10 CFR Part 2, Appendix C. Review of DC's was performed to maintain a realtime status of deficiencies, determine regulatory compliance, follow the licensee corrective actions, and assist as a basis for closure of the LER when reviewe Due to the numerous DC's processed only those DC's which result in enforcement action or further inspector followup with the licensee at the end of the inspection are discussed as listed below. The LER's denoted with an asterisk indicates that reactive inspection occurred at the time of the event prior to receipt of the written repor (1) Deficiency Card reviews:

DC 1-97-261C "DS-416 Reactor Trip Breaker Inspections" This deficitocy dowments the results of the weld inspection During the inspe-tions the NRC resident and vendor branch inspectors were prasen The results of. the inspection were acceptable hv. : ave r the NRC recommended that the shafts be

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replaced in the ior.2 term. These inspections were performed to address the concerns as addressed in Information Notice N .

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DC 1-87-2708 "RVLIS Impingement Covers" On 10/23/87 the impingement cover plates for RVLIS tubing for 1-LX-1310 and 1-LX-1320 were not. installed. In order to correct this problem new plates were fabricated and installe This resulted in a delay in return to powe DC 1-87-2733 " Control Room Isolation Whi'e Resetting Radiation Monitors" This DC describes an unplanned actuation on 10/26/87 when the radiation parameter resetting procedure did 'not call for blocking of the outpu In addition poor communication between operators and the chemistry department was exhibited in that the status of the Control Room Ventilation being reset was not fully understood nor was the nature of work to be performe DC 1-87-2753,1-87-2766,1-87-2846 " Mode 3 Entry Performed without all requirements met" These deficiency cards documented three instances that the licensee identified after the unit entered Mode The three cases were failure to perform IST testing on the A train AFW discharge check valve following maintenance, failure to have the Steam Driven AFW pump steam admission valves open, and failure to perform a functional test of the A train safety injection pump following changeout of the lubrican Each instances had minimal impact as follows: the check valve tested satisfactorily, full secondary steam pressure had not been obtained to support the surveillance testing, and the safety injection pump was tested satisfactoril DC 1-87-2915 " Reactor Trip While Performing OSP 14701-1" This Reactor Trip resulted when the B train auto shunt trip test panel was used during the testing of the A' train breaker. While the procedure directed the operator to the correct test panel no I labeling was in place at the test panel to indicate that the l wrong train was being utilized. During the performance of the undervoltage coil trip test no additional indication existed to indicate that the shunt coil had not been blocked. When the shunt coil trip test was' executed the B train shunt coil energized and the B train reactor trip breaker opened. Since'

the A train SSPS was in test' to support A train reactor trip breaker testing the control room operator had to insert a manual trip to open the A train reactor trip breaker and perform' a manual start of the A train Auxiliary Feedwater Pum DC 1-87-2974 " Missed Surveillance" This deficiency occurred on )

November 16 when a ' room temperature surveillance was not performed due to the floor being painted. The operator NA'd the step which was later identified during a supervisor review and ,

at that time it was noticed that the TS had been misse (2) The following LER's were reviewed and are ready for closure pending verification that the licensee's stated corrective '

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(a) 50-424/87-05, Rev 0-4 "120V AC Voltage Transient Causes ESF Actuations" These LERs describe a plant condition where a voltage transient causes ESF actuations upon energization of the Safety System Sequencer Panel . The inspector noted to the licensee that the final supplemental LER was due on July 30, 1987. The licensee informed the inspector that the LER will be closed on January 1988 once

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the information is received from Westinghous (b) 50-424/87-20, Rev 0 "ESF Actuation Caused by Excessive Leakage Through a Main Feedwater Regulating Valve" The inspector noted to the licensee that the final supplemental LER was due on July 10, 1987. The licensee informed the inspector that the LER will be closed once the final corrective action is performe The LER states that further testing of valve IHV-5139 will be performed when the unit is in Mode 3. The Licensee failed to accomplish this test during the outage but will do the test at the next forced outage or refueling. The final LER will be issued following the tes (c) 50-424/87-56, Rev 0 " Technical Specification Not Met Due To Incomplete Vendor Software For Dose Calculations" This LER describes an event which occurred on September 16, 1987 when it was identified that the cumulative dose calculation program for gaseous releases to the atmosphere for radioiodines did not include isotope I-133 in the software package. The licensee identified this during a data review while preparing the semi-annual radioactive effluent release repor Corrective action includes revising the software and the performance of a functional testing. The inspector has no further questions regarding this repor The following is identified:

50-424/ LIV 87-60-01 " Failure To Implement an Appropriate Surveillance to determine cumulative dose contributions in accordance with the ODCM per TS 4.11.2.3 - LER 87-56" (d) 50-424/87-58, Rev 0 " False Signal From Rad Monitor Leads To Control Room Isolation" This LER describes an event which occurred on September 21, 1987 when the control room isolation occurred due to a false high radiation signal from 1-RE-12116. While no violations resulted from this event the licensee has yet to specify the root cause of the failure in a supplemental report due December 15, 198 (3) The following LER's were reviewed and are considered close (a) 50-424/87-01,. Rev 0 " Incorrect Transmitter Circuit Board Leads to Missing a Required Flow Rate Estimation" This LER was reviewed in NRC Rpt 50-424/87-44 and required

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verification of the corrective action The inspector reviewed procedure 34226-C and the training attendance sheet The following item is identified:

50-424/ LIV 87-60-02 " Failure to Perform required TS Surveillance to Verify c'ompliance with TS 3.3.3.10 -

LER87-01" (b) *50-424/87-02, Rev 0 " Potential Failure of MSIV's to Close

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Following Small Steam Line Break" This LER was reviewed in NRC Rpt 50-424/87-44 and required verification of the corrective action The inspection reviewed the vendor qualification report dated 3-20-87. This report documents that the main steam isolation valves which were supplied l

can remain in the open position for approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />

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while exposed to a 320 degree F environment and retain the capability of closing and stopping steam flow in the syste The inspector noted that the test configuration included the relief valve (4100 psi) and that the hydraulic pressure reached only 3950 psi during the test. DCR 87 VIE 0030 was also reviewed. No further corrective actions are required as a result of the test repor (c) *50-424/87-03, Rev 0 " Restriction of Pipe Movement with Incorrect Penetration Sealant Material" This LER was reviewed in NRC Rpt 50-424/87-44 and corrective action was verified during the course of the event. The inspector has no further question (d) *50-424/87-04, Rev 0 " Containment Isolation Actuations Caused by Faulty Circuit Board" This LER was reviewed in NRC Rpt 50-424/87-44. Corrective action was serified regarding the repair of the f aulty circuit during the course of the even The inspector verified that a new annunciator has been added and 17006-1 response procedure change In addition the inspector noted that the radiation monitors have been removed as an input to containment isolatio (e) *50-424/87-06, Rev 0 "ESF Actuation of Auxiliary Feedwater Due to Inadvertent Trip of the Main Feedwater Pumps" This LER was reviewed in NRC Rpt 50-424/87-44 and corrective action was verified during the course of the event. The inspector notes'that a further corrective action has been the practice of removing the control fuses to the actuation circuit for AFW. This practice has resulted in LER 87-36 when the wrong fuses were pulle (f) *50-424/87-07, Rev 0 "ESF Actuation Caused by Steam Generator Water Level"; *50-424/87-09, Rev 0 "ESF Actuation Caused by Adjustments to Steam Generator Level Control

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Systems"; *50-424/87-10 Rev 0 "RPS Actuation Caused by Adjustments to Steam Generator. Level Control Systems";

  • 50-424/87-18, Rev 0 " Reactor Trip Caused by Faulty Bistable Circuit Board"; *50-424/87-24, Rev 0 " Procedure-Inadequacy Causes Auxiliary Feedwater Actuation";
  • 50-424/87-25, Rev 0 " Reactor Trip Due to Startup Test Procedure Inadequacy"; *50-424/87-27, Rev 0 " Reactor Trip Caused by Inadvertent Closure of MSIV During . Maintenance";
  • 50-424/87-30, Rev 0 " Lightning Causes Reactor Trip Due to Incorrectly Grounded Current Transformer"; *50-424/87-31, Rev 0 " Auxiliary Feedwater System Actuation During Startup Test Due to Procedure Inadequacy"; *50-424/87-34, Rev 0

" Reactor Trip Due to Failure of Main Feedwater Pump Discharge Check Valve"; *50-424/87-35, Rev 0 " Faulty Main Feedwater Pump Turbine Hydraulic Tubing Connection Leads to Reactor Trip"; *50-424/87-36, Rev 0 " Auxiliary Feedwater Actuation Circuitry Inoperable Due to Personnel Error";

  • 50-424/87-39, Rev 0 " Pressure Transmitter Failure Causes ESF Actuation on Steam Generator Hi-Hi Water Level";
  • 50-424/87-41, Rev 0 " Reactor Trip Due to Improperly Calibrated Field Current Transducers"; *50-424/87-50, Rev 0

" Reactor Trip Caused- by Instrument Technician's Error".

These LERs were reviewed in NRC Rpt 50-424/87-38 and NRC i Rpt 50-424/87-44 with corrective action verified during the course of the event Additional NRC concerns were addressed in several management meetings regarding the control of Steam Generator water level. Improved system performance resulted from increased operator experience and additional system tunin (g) *50-424/87-11, Rev 0 " Trip due to Lo-Lo Steam Generator Level" This LER was reviewed in NRC Rpt 50-424/87-44. The inspector noted that the corrective. actions included temporary markings on the site glass and an engineering evaluation to determine further correction actio The inspector questioned the final status of these two actions and was informed that no further actions were necessar (h) *50-424/87-13, Rev 0 "Feedwater System Valve Malfunctions Result in Reactor Trip" This LER was reviewed in NRC Rpt 50-424/87-44 and at the time of the event. MWO 1-87-4987 was reviewed to verify proper reassembl LER 87-34 describes a repeat failure of the same check valve and describes further corrective actio !

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(i) *50-424/87-15, Rev 0 " Inadvertent Steam Dump Operation Results in ESF Actuation' This LER was reviewed in NRC Rpt 50-424/87-44 and at the time of the event. Training was verified regarding the connection of test rack The inspector noted to the licensee that the LER implies that the steam header pressure control loop was tested after the '

event to ensure its proper operation was part of the corrective action, when in fact the only testing was as part of the, power ascension test phas The licensee has j not been responsive in revising the LE (j) *50-424/87-19, Rev 0 " Control Room Isolation Due to Signal -

From Toxic Gas Monitors"; *50-424/87-28, Rev 0 " Control i Room Isolations Caused by Spurious Signals From Toxic Gas Monitor" Procedure 24537-1 and 24538-1 were reviewed to verify that monthly calibration checks were implemente It was noted that the licensee is not required to have operable monitors since chlorine is removed from the sit The licensee is pursuing a TS change to raise the setpoint from 2 to 5 ppm to eliminate spurious actuations and - then return chlorine onsite ]

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(k) 50-424/87-21, Rev 0 " Control Room Isolation Initiated by Radiation Monitor Loss of Power" The final corrective actions for this problem will be discussed along with the resolution of LER87-0 (1) *50-424/87-23, Rev 0 "RHR System Minimum Flow Requirement  ;

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Potentially Not Met Due to Partially Closed Valves" This LER was reviewed in NRC Rpt 50-424/87-31 and resulted in ,,

the identification of a Severity Level III' Violation 50-424/87-31-0 Procedure 14460-1 was verified to have the changes and the preventive maintenance sheets indicate the calibration frequency to be every six month The corrective MW0s were also reviewe (m) *50-424/87-32, Rev 0 " Operator Error Leads to a Reactor Trip on Source Range High Flux" Procedure 12003-1 was reviewed to verify the requirement for a ICRR plot and a reactor engineer. Procedure 14940-1- was reviewed for to verify incorporation of correct boron worth and that the procedure will be performed by a reactor engineer. The training plan and simulator changes were reviewe (n) *50-424/87-33, Rev 0 " Reactor Trip on Steam Generator Lo-Lo Level While Transferring Feedwater Flow" Procedure 12004-1 was reviewed to verify that the correct power levels were indicated for transferring from the Bypass Feedwater regulating valve to the Main Feedwater regulating valv a

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(o) *50-424/87-37, Rev 0 " Failure to Meet Technical *

Specification Action Statement Due to Procedural Inadequacy" Procedure 00150-C was reviewed to verify the additional guidance was incorporate The inspector interviewed the NSSS engineering supervisor to determine the results of the LLRT performed during the outage. The results indicated that while degradation was noted the valve was within the acceptance criteri The inspector determined that no actual TS violation had occurred since the valve was inoperable due to the potential that the leakage was high. This event served in identifying a procedural system weaknes (p) *50-424/87-38, Rev 0 " Manual Reactor Trips Due To Overly Conservative Annunciator Response Procedure" Procedure 17010-1 was reviewed to verify that the response procedure has been revised to place DRPI in. the Data A or Data B to regain rod position indication pri'or to a manual tri (q) *50-424/87-42, Rev 0 " Boron Concentration Exceeds Tec Spec. Limiting Condition of Operation Time Limit" The tickler sheet was reviewed to show the correct TS limit The memorandum regarding surveillance was also reviewe This item is identified as follows:

50-424/ LIV 87-60-03 " Failure to Adequately Perform required TS Surveillance to Verify compliance with TS 3.1.2.6.b -

LER87-42" (r) *50-424/87-43, Rev 0 " Improper Performance of Containment Pressure Surveillance Due to Personnel Error" Procedure 14000-1 was reviewed to verify that the computer point was included in the procedur This item is identified as follows:

504424/ LIV 87-60-04 " Failure to Adequately Perform required TS Surveillance to Verify compliance with TS 3.6.1.4 -

LER87-43" (s) 50-424/87-46, Rev 0 " Waste Gas Decay Tank Not Sampled Within Technical Specifications Time Limit" The memorandum regarding surveillance was reviewedf Corrective actions include the establishment of fixed tim This item was identified in NRC report 50-424/87-49 as an LI (t) *50-424/87-57, Rev 1 " Procedure Deficiency Results in Failure to Trip Overtemperature Delta T Reactor Trip Bistable". This LER describes an event which occurred on August 8,1987 when the shift failed to place one of four

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required bistables in trip. The error was identified on August 9,1987 during a control panel walkdown. The root cause was a procedural deficiency in specifying the correct bistables to trip. The inspector noted that the failure mode consisted of the pressure instrument drifting high about 40 psi and not a . total failure high. At the inspectors request engineering performed a calculation to show the effect that this pressure drift would have on the setpoin This calculation showed that even with this error the setpoint was within the 6.6% total allowanc The procedure was reviewed and the corrective actions have been completed. The inspector also noted that the LER was submitted late due to an improper review of the deficiency car Both items above represent violations of NRC

, requirements where the licensee has met the criteria for no i citatio To track these items the following are.

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50-424/ LIV 87-60-05 " Failure to Place the OTDT Trip-Bistables in the Trip Condition per TS 3.3.1 Item 7 - LER 87-57" and 50-424/ LIV 87-60-06 " Failure to Submit an LER Within 30 Days After The Discovery of the Event per 10 CFR 50.73(a)(1) - LER 87-57" 5. Management Meetings (30302B)

On October 21, 1987, an enforcement conference was held to discuss the results of NRC report 50-424/87-5 On Ncvember 9,1987, a site tour was given to the Director, Office of Nuclear Reactor Regulation (NRR), Thomas Murley and the Associate Director for Inspection & Technical Assessment, Richard Starostecki by the resident inspector Following the tour, two meetings were conducted with the licensee. The first meeting was held with the Unit 1 operations personnel and the second meeting was held with the Unit 2 construction personne On November 10, 1987, the fourth onsite meeting with the licensee was held regarding the performance of the uni '

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