IR 05000387/1990021

Insp Repts 50-387/90-21 & 50-388/90-21 on 901007-1103.No Violations or Deviations Noted.Major Areas Inspected: Operations,Radiological Controls,Maint/Surveillance Testing, Emergency Preparedness,Security & LERs

ML17157A472
Person / Time
Site:	Susquehanna
Issue date:	12/10/1990
From:	Swetland P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML17157A471	List: IR 05000387/1990021 ML17157A470
References
50-387-90-21, 50-388-90-21, NUDOCS 9012200064
Download: ML17157A472 (61)

Text

U. S. NUCLEAR REGULATORY COMMISSION

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REGION I

Report Nos.

License Nos.

50-387/90-21; 50-388/90-21 NPF-14; NPF-22 Licensee:

Pennsylvania Power and Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Facility Name:

Inspection At:

Susquehanna Steam Electric Station Salem Township, Pennsylvania Inspection Conducted:

October 7, 1990 - November 3, 1990 Inspectors:

G. S. Barber, Senior Resident Inspector, SSES J. R. Stair, Resident Inspector, SSES J. L. Dixon, Reactor Engineer, DRS R. A. McBrearty, Reactor Engineer, DRS P.D. Kaufman, Project Engineer, DRP Approved By:

'w)iot'[0 P. Swetland, Chief Date Reactor Projects Section No. 2A, Ins ection Summa

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radiological controls, maintenance/surveillance testing, emergency preparedness, security, engineering/technical support, safety assessment/quality verification, and Licensee Event Reports, Significant Operating Occurrence Reports, and Open Item Followup.

R~eult:

During this inspection period, the inspectors found that the iicensee's activities were directed toward nuclear and radiation safety.

No violations or deviations were identified.

An Executive Summary is included and provides an overview of specific inspection findings.

't'0 1 2200064 901 21 r PDR ADOCV 05000387 Q

PDR

C'g TABLEOF NTENT I.

EXECUTIVESUMMARY II.

DETAILS SUMMARYOF OPERATIONS......

1 ~ 1 Inspection Activities.........

1.2 Susquehanna Unit 1 Summary...

1.3 Susquehanna Unit 2 Summary...

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2.

PERATIONS o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

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2.1 Inspection Activities.....................,........

2.2 Inspection Findings and Review ofEvents.....,.....,.....

2.2.1 Improper Reactor Water Cleanup Valve Torque Switch Setting Unit 2

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2.2.2 Unplanned Start of "B" SGTS and Ventilation System Isolation

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RADIOLOGICALCONTROLS 3.1 Inspection Activities..................

3.2 Inspection Findings 3.2.1 Contaminated Injured Worker - Unit 1 H

MAINTENANCE/SURVEILLANCE............

4.1 Maintenance and Surveillance Inspection Activity 4.2 Maintenance Observations 4.3 Surveillance Observations.. ~.........

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4.4 Inspection Findings

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EMERGENCY PREPAREDNESS 5.1 Inspection Activity.......

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5.2 Inspection Findings 5.2.1 Transportation Offsite of a

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Contaminated Injured Worker - Unit

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SECURITY 6.1 Inspection Activity ~........

6.2 Inspection Findings 6.2.1 Security System Concern

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'ta ENGINEERING/TECHNICALSUPPORT 7.1 Inspection Activity.............

7.2 Inspection Findings 7.2.1 Prioritization of Work Concern

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7.2.2 Inservice Inspection - Unit 1

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SAFETY ASSESSMENT/QUALITY VERIFICATION 8.1 Licensee Event Reports (LERs), Significant Operating Report (SOORs), and Open Item (OI) Followup...

8.1.1 Licensee Event Reports..............

8.1.2 Significant Operating Occurrence Reports...

8.1.3 Open Items.....................

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Occurrence MANAGEMENTAND EXIT MEETINGS 9.1 Routine Resident Exit and Periodic Meetings.........,...

9.2 Plant Discrepancy Management Meeting................

9.3 Attendance at Management Meetings Conducted By Region Based Inspectors Inspection Reporting

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. EXECUTIVESUMMARY Susquehanna Inspection Reports 50-387/90-21; 50-388/90-21 October 7, 1990 - November 3, 1990

~Oereti ne (30703, 60710, 71707)

The licensee identified that the Unit 2 Reactor Water Cleanup inboard isolation valve's torque switch was improperly set during a previous outage.

Since the switchsetting was set low, the capability to perform its intended safety function was questioned.

Based on low safety significance and compensatory actions taken by the licensee, a temporary waiver of compliance and emergency technical specification amendment were requested and granted in order to prevent a unit shutdown.

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(717tl7j An unplanned start of the "B" Standby Gas Treatment System and a Zone IIIventilation system isolation occurred due to performing certain procedural steps out of sequence.

The cause of this event was attributed to personnel error and inadequate procedural guidance.

The licensee took ap ro riate action to prevent future recurrence.

Individual workers and Health Physics personnel implemented radiological protection program requirements.

Periodic inspector observation noted no inadequacies in the licensee's implementation of the radiological protection program.

A worker received a contaminated puncture wound from what was believed to be a splinter, The wound was cleaned on-site and the individual escorted off-site to a local hospital for removal of the splinter which was thought to have remained in the wound.

No splinter was found and the individual returned to work. The licensee's dose assignment showed that long-term exposure to the worker would be very low. Actions taken by the licensee in response to this event were thorough and conservative.

Maintenance/

urveillance (61726, 62703)

The licensee exercised good control of maintenance and surveillance activities.

No scrams or ESF actuations were attributable to maintenance or surveillance activitie Executive Summary Emer enc Pre aredne s (71707)

A worker received a contaminated splinter in his thumb that required his transport off-site.

The requirements for activating the Emergency Plan for transportation of contaminated injured individuals off-site specified external contamination.

Since the individual was successfully decontaminated externally, no entry into the Emergency Plan was made.

The appropriate 10 CFR 50.72 notification was made.

The licensee's response to this event was commensurate with its safety significance.

~ecurit (71707)

Routine observation of protected area access and egress control showed good control by the licensee.

Review of the licensee's program to upgrade security system hardware determined that identified security deficiencies are being adequately addressed and resolved.

En ineerin /Technical u

(71707, 35702)

Review of the management of engineering work activities determined that they are being performed in accordance with applicable procedures and, that they are being properly prioritized and executed.

The ultrasonic (UT) examination of certain Residual Heat Removal System welds and the calibration of the ultrasonic system equipment was observed.

The results correlated well with previous UT exams and the UT system was found to be calibrated per the procedure.

afe A

e ment/As urance of ualit (90712, 92701, 92720 92700)

Six long-standing open inspection items were reviewed and closed during this inspection.

A Licensee Event Report concerning falsification of firewatch logs by contractor personnel was reviewed.

This event was investigated by the licensee and he found that only four personnel were involved and have been removed from the site.

The NRC, concerned with the depth of the investigation, asked the licensee to perform a more extensive review to determine ifthe condition was more widespread.

The results of this review indicated that the falsification was limited to the cases originally investigated.

The identified cases of potential wrongdoing are still under review and evaluation by the NR DETA~

1.

SUMMARYOF OPERATIONS The purpose of this inspection was to assess licensee activities at Susquehanna Steam Electric Station (SSES) as they related to reactor safety and worker radiation protection.

Within each inspection area, the inspectors documented the specific purpose of the area under review, the scope of inspection activities and findings, along with appropriate conclusions.

This assessment is based on actual observation of licensee activities, interviews with licensee personnel, measurement of radiation levels, independent calculation, and selective review of applicable documents.

Abbreviations are used throughout the text.

Attachment 1 provides a listing of these abbreviations.

1.2 Sus uehanna Unit 1 Summa Unit 1 remained in its fifth refueling outage throughout the inspection period.

Outage related activities continued on the ECCS and ESF systems and components.

The core was fully reloaded and the reactor vessel reassembled.

Condition 4 was reentered on October 30.

The plant was in the process of performing the reactor vessel operational pressure test and closeout of primary containment at the end of the inspection period.

An inadvertent start of the "B" train of the SGTS occurred on October 8, while restoring from load center outage work.

See Section 2.2.3 for details.

An individual working in thereactor cavity injured his thumb and was transported to a local hospital as a contaminated injured person.

See Section 3.2.1 for details.

1.3 u

uehanna nit 2 Summa Unit 2 began the inspection period at full power.

Power was decreased to 80 percent on October 10 due to elevated levels of reactor water conductivity and sulfates.

The licensee determined that the elevated levels were due to resin fines from one of the condensate demineralizers.

The unit remained at this decreased power level pending commencement of the changeout of condensate filter/demineralizers and was restored to full power on October 27. Full power was maintained through the remainder of the period.

No ESF actuations occurred during the period.

2.

OPERATIONS The inspectors verified that the facility was operated safely and in conformance with regulatory requirements, Pennsylvania Power and Light (PP8cL) Company management control was evaluated by direct observation of activities, tours of the facility, interviews and

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discussions with personnel, independent verification of safety system status and Limiting Conditions for Operation, and review of facility records.

These inspection activities were conducted in accordance with NRC inspection procedure 71707.

The inspectors performed 180 hours0.00208 days <br />0.05 hours <br />2.97619e-4 weeks <br />6.849e-5 months <br /> of normal and back shift inspections including deep backshift inspections on: October 15, from 3:00 a.m. to 6:00 a.m.; and, October 26, from 3:00 a.m. to 6:00 a.m.

2.2 In i n Findin s and Review f Even 2.2.1 Im ro er Reactor Water l

nu Valve Tor ue Switch ettin

- Uni 2 NCR 90-0166 documented a nonconforming condition discovered with the Reactor Water Cleanup Inboard Isolation Valve, HV-244F001 on Unit 2.

This NCR stated that the torque switch setting for the valve is set at 1 1/2 instead of 1 3/4. A torque switch setting of 1 3/4 is required to develop enough force to close the valve against a differential pressure of 1000psid.

The licensee's initial operability evaluation concluded that the valve was operable.

This was based, in part, on an interpretation the licensee received in 1987 from the NRC on a similar condition with a HPCI isolation valve.

This operability determination was reviewed during the NRC Maintenance Team Inspection (50-387/90-81).

The inspector questioned the applicability of the previous HPCI interpretation to the RWCU valve.

Based on this NRC concern, the licensee reevaluated the RWCU valve's operability relative to TS 3.6.3 and concluded that the valve was inoperable.

Compliance with this TSAS Action would require isolation of the affected penetration within four (4) hours.

Thus, the licensee considered this action in light of the very limited safety significance of this nonconformance, and also the significant chemistry transient that would occur.

The effects of this TSAS was discussed with the NRC during a conference call on October 23.

During this call, the licensee requested a temporary waiver of compliance for this valve with respect to the actions required by TS 3.6.3.

NRC Region I granted a verbal waiver of compliance which which granted relief from closing valves HV-244F001 and HV-244F004 based on certain compensatory actions.

This waiver was granted by Mr. P.D. Swetland to J.A. Blakeslee at 6:30 p.m. on October 23.

This waiver was followed up by a written request on the part of the licensee on October 24 and the written approval was documented by NRC on October 24.

An emergency TS amendment was granted by NRC on October 31 based on the establishment of certain compensatory actions.

The licensee developed these actions based on a review of operating, abnormal and emergency procedures.

This review determined that adequate compensatory actions could be provided by existing procedures except for some minor changes.

Specifically, the licensee

reviewed existing guidance to the operators in the event of a RWCU suction line break outside of containment and concluded that the existing guidance provides proper direction.

As additional compensatory action, a note was added to procedure AR201-001 to alert the operator to this particular NCR on the HV-244F001 valve, The inspector reviewed the licensee's actions and questioned whether these compensatory actions could be taken in less than 19 minutes.

This action time was determined by the licensee in their safety evaluation to ensure that the off-site doses remained a small fraction of the 10 CFR 100 guidelines.

The licensee assured the inspector that they, were confident that the additional procedural controls were adequate to ensure timely closure of the valve.

The licensee also agreed to reset the torque switch during the next unplanned or planned outage.

A December 14 outage is being planned to correct this and other deficient conditions.

The inspector had no more questions on this issue.

2.2.2 n lanned tart of "B" T

an V ntil i n tern I olation On October 8, with Unit 1 defueled and Unit 2 operating at full power, division 2 of Zone IIIof the ventilation system isolated and the "B" train of SGTS automatically started.

This occurred during restoration of normal power to selected Zone I and Zone IIIventilation dampers following the scheduled 4.16 KV Bus 1D outage.

Within 35 minutes following this occurrence, the normal ventilation flow path was restored and the SGTS returned to standby.

The appropriate ENS call for an ESF actuation was made in accordance with 10 CFR 50.72 (b)(2)(ii) within the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> time requirement.

The licensee attributed the cause of the event to personnel error and procedural deficiencies since the operator performed two procedure steps out of sequence and the procedure did not provide appropriate cable identification.

This resulted in a wire termination error and the Zone IIIlock out relay energized.

The licensee reviewed the event with Operations personnel to re-emphasize procedure adherence and the importance of concise communication.

Maintenance personnel also willbe briefed on the importance of concise communication.

The licensee has agreed to revise the Operating Procedure used for this evolution to permanently identify the original and spare cables.

The licensee plans to have the corrective actions completed by March, 1991.

The inspector reviewed the circumstances surrounding this event and discussed the event with appropriate licensee personnel.

As a result, the inspector. considers the licensee's actions in response to the event appropriate.

No inadequacies were note.

RADIOLOGICALCONTROLS 3.1 In ecti n Activiti PP&L's compliance with the radiological protection program was verified on a periodic basis.

These, inspection activities were conducted in accordance with NRC inspection procedure 71707.

3.2 ~li i

Ch Observations of radiological controls during maintenance activities and plant tours indicated that workers generally obeyed postings and Radiation Work Permit requirements.

No inadequacies were noted during these observations, 3.2.1 n

min ted In'ured W rker -

nit 1 On November 2, the licensee reported that an individual working in the Unit 1 reactor cavity had received a contaminated metal splinter in his thumb and that he was being transported off-site to a medical facility since they could not remove all of the splinter on-site.

The external portion of the splinter was removed and only a small internal piece remained.

The licensee later determined that no splinter actually remained in the wound but that some low levels of residual contamination did remain.

The licensee made the appropriate 10 CFR 50.72 notification within the allotted time interval.

The licensee did not declare an unusual event and activate their emergency plan for this event since the individual was not externally contaminated following initial cleansing of his thumb.

The emergency plan specifically states that the injured person must be externally

'ontaminated.

This situation was discussed with the resident inspectors who concurred with the licensee's actions.

The licensee assessed the individual's dose and found that the long-term exposure would be relatively minor.

The inspector reviewed the licensee's report on the event and discussed the event and dose assignment with appropriate licensee personnel.

The inspector considered the licensee's actions in response to this event appropriate.

No inadequacies were noted.

4.

MAINTENANCE/SURVEILLANCE 4.1

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Maintenance and Surveill nce Ins ecti n Activit On a sampling basis, the inspector observed and/or reviewed selected surveillance and maintenance activities to ensure that specific programmatic elements described below were being met.

Details of this review are documented in the following section.2 Min'nn rv i n The inspector observed and/or reviewed selected maintenance activities to determine that the work was conducted in accordance with approved procedures, regulatory guides, Technical Specifications, and industry codes or standards.

The following items were considered, as applicable, during this review: Limiting Conditions for Operation were met while components or systems were removed from service; required administrative approvals were obtained prior to initiating the work; activities were accomplished using approved procedures and quality control hold points were established where required; functional testing was performed prior to declaring the involved component(s)

operable; activities were accomplished by qualified personnel; radiological controls were implemented; fire protection

, controls were implemented; and the equipment was verified to be properly returned to service.

These observations and/or reviews included:

HPCI turbine preventive maintenance on overspeed trip assembly, performed per WA P01796, on October 25, 1990.

Inspection of the "A" RHR pump seal cooler, performed per WA P04414, on November 2, 1990.

Troubleshooting of Fire Protection OS&Y Valve PA123 for lack of instrument air pressure, performed per WA S07805, on November 2, 1990.

4.3 rveill nce erv i n The inspector observed and/or reviewed the following surveillance tests to determine that the following criteria, ifapplicable to the specific test, were met:

the test conformed to Technical Specification requirements; administrative approvals and tagouts were obtained before initiating the surveillance; testing was accomplished by qualified personnel in accordance with an approved procedure; test instrumentation was calibrated; Limiting Conditions for Operations were met; test data was accurate and complete; removal and restoration of the affected components was properly accomplished; test results met Technical Specification and procedural requirements; deficiencies noted were reviewed and appropriately resolved; and the surveillance was completed at the required frequency.

These observations and/or reviews included:

SE-159-020 LLRTs of CAC Penetrations X80C, X238B, X233 and X91A, performed on October 19.

SE-159-033 LLRTof RWCU Supply Penetration Number X-14, performed on October 1 \\

SE-149-001 TP-055-001 6.

Eighteen'Month RHR System and Logic Functional Test (Division 1), performed on October 25.

CRD Stroke Time Test, performed on October 25.

4.4 In tion Findin The inspector reviewed the listed maintenance and surveillance activities.

The review noted that work was properly released before its commencement; that systems and components were properly tested before being returned to service and that surveillance and maintenance activities were conducted properly by qualified personnel.

Where questionable issues arose, the inspector verified that the licensee took the appropriate action before system/component operability was declared.

No unacceptable conditions were identified.

5.

EMERGENCY PREPAREDNESS 5.5 I~iA The inspector reviewed licensee event notifications and reporting requirements for events that could have required entry into the emergency plan.

5.2 In ection Findin No events were identified that required emergency plan entry.

However, one event was reviewed.

See Section 5.2.1 for details.

5.2.1 T

n rtation ffie f on min t In'ured W rker-nit 1 The licensee decided not to activate the emergency plan for the worker who was contaminated from the splinter discussed in Section 3.2.1.

Since the worker was not externally contaminated when taken to the local hospital, activation of the emergency plan was not required.

The appropriate 10 CFR 50.72 notification to the NRC was made in a timely manner.

The inspector agreed with the licensee's judgement in this determination.

6.

SECURITY 6.1 In ti n Activit PPEcL's implementation of the physical security program was verified on a periodic basis, including the adequacy of staffing, entry control, alarm stations, and physical boundaries.

These inspection activities were conducted in accordance with NRC inspection procedure 7170.2

The inspector reviewed access and egress controls throughout the period.

No unacceptable conditions were noted.

6.2.1

ec rit S stem oncern

A concern was raised that security system hardwar'e problems were not being corrected or budgeted for since the NRC's Regulatory Effectiveness Review (RER) results were good.

An NRC RER was conducted in June 1989 at the Susquehanna Plant.

The NRC's RER report was reviewed by the inspector to determine ifany security findings warranted corrective actions to resolve identified deficiencies.

The results of the RER report disclosed that some findings required licensee action.

The licensee transmitted their corrective actions to the NRC in a letter dated January 25, 1990.

The letter identified the preliminary scope and schedule required to complete their security upgrade program, which is divided into three phases.

The licensee's security upgrade program includes modifications and improvements to correct the identified RER findings.

Some security system hardware modifications have already been completed to correct the RER findings.

The licensee established the security upgrade program after completing their own internal RER in 1988.

However, some security modifications and improvements were not initiated pending the NRC's RER.

The licensee intended to compare and incorporate the NRC's RER findings into their own RER results prior to proceeding with the security upgrade program.

The inspector concluded, that based on review of the NRC's RER report, the licensee's response letter to the NRC's RER, the licensee's security upgrade program, and modifications already implemented to the security system, identified security deficiencies are being adequately addressed and resolved under the licensee's security upgrade program, 7.

ENGINEERING/TECHNICALSUPPORT 7.1 In ecti n Activit The inspector periodically reviewed engineering and technical support activities during this inspection period.

The on-site Technical lTech) section, along with Nuclear Plant Engineering (NPE) in Allentown, provided engineering resolution for problems during the inspection period.

The Tech section generally addressed the short term resolution of problems while NPE scheduled modifications and design changes, as appropriate, to provide long lasting problem correction. The inspector verified that problem resolutions were thorough and addressed at preventing recurrences.

In addition, the inspector reviewed short term actions to ensure that the licensee's actions provided reasonable assurance that safe operation could be maintaine C

7.2 7.2.1 I'riri i ti n of W rk n

rn A concern was raised that no system assignments are made within the Nuclear Plant Engineering (NPE) organization.

In other words, whomever is free works on whatever comes into the group.

The licensee's NPE engineering management work practices is an area that was previously inspected by the NRC during November 1989.

The inspection is documented in NRC inspection report 50-387/89-29 and 50-388/89-32.

Results of this inspection indicate that the licensee's engineering work activities are being executed and prioritized according to Engineering Procedure Manual (EPM) EPM-QA-102, Engineering Work Requests (EWR's).

The prioritization of work activities was found to be acceptable.

As a followup to the original concern, the inspector reviewed EPM-QA-102 to verify that administrative lines of responsibility are clearly identified for determining EWR priority and that the appropriate disciplines are clearly assigned for completing engineering tasks.

The procedure was found to clearly delineate the lines of responsibility necessary to assure engineering work activities are adequately planned, prioritized, and dispositioned.

Based on the previous NRC inspection report and review of the EWR procedure, the inspector concluded that engineering work management practices are functioning per procedure to support the plant and licensee needs.

7.2.2 Inservice In ection -

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The General Electric Company was contracted by the licensee to perform nservice inspection during the current refueling outage.

Ultrasonic examinations were performed using the computerized Ultra Image III "Smart Scan" ultrasonic system.

The inspector observed a portion of the ultrasonic examination of 24" diameter welds DCA-1102-FW11 and DCA-108-1-FW12, valve-to-pipe and pipe-to-elbow respectively, in the Loop B RHR system.

The inspectors observations included the calibration of the ultrasonic system prior to the examination of weld DCA-108-1-FW12.

The calibration was performed using the licensee's calibration block P56.

The previous inspection reports and documentation related to welds DCA-1102-FW1, DCA-108-1-FW12 and HBB-1111-1-A were reviewed.

Current examinations results were compared with previous results and good correlation was noted.

During this inspection, no unacceptable conditions were identified.

8.

SAFETY ASSESSMENT/QUALITY VERIFICATION

-9.

8.1 Licensee Event Reports (LERs), Significant Operating Occurrence Report (SOORs),

and Open Item (Ol) Followup (90712, 92700)

8.1.1 Licen Event Re The inspector reviewed LERs submitted to the NRC office to verify that details of the event were clearly reported, including the accuracy of the description of the cause and the adequacy of corrective action.

The inspector determined whether further information was required from the licensee, whether generic implications were involved, and whether the event warranted onsite followup. The following LERs were'reviewed:

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gni~l

~90-002-00 Firewatch Rounds Not Completed As Required 90-021-00 Division II Zone IIIIsolated and SGTS Started During Restoration of ESS 4.16 KU BUS ID (1A204) Due to Personnel Error. This event was reviewed in Section 2.2.3.

90-022-00 Emergency Diesel Generator "C" Unplanned Automatic Start.

This event was reviewed in Inspection Report 50-387/90-20; 50-388/90-20.

8.1.1.1 n i F

11 wu fLicen ee Event Re For those LERs selected for onsite followup (denoted by asterisks in Detail 8.1.1), the inspector verified that the reporting requirements of 10 CFR 50.73 had been met, that appropriate corrective action had been taken, that the event was adequately reviewed by the licensee, and that continued operation of the facility was conducted in accordance with Technical Specification limits. The. following findings relate to the LERs reviewed on site:

LER

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2-Firewatch Round N t m leted as R uired On January 14, 1990, the licensee was informed by a contractor employee functioning as an hourly roving firewatch that he had not completed his assigned rounds on midnight shift even though he had signed off the firewatch log sheets.

As a result, the licensee performed a review of the previous two weeks of midnight shifts by checking security access points.

This review identified that four of the seven personnel assigned firewatch duties on midnight shift during that period did not complete all of their assigned round Investigation by the licensee determined that the four individuals involved had received adequate training/direction and understood their responsibilities.

Employment of those four individuals was therefore terminated.

It was considered possible that lack of midnight shift supervision contributed to the event and the licensee responded by implementing additional supervisory unannounced backshift inspections.

A watchman key code station system was also implemented to provide a permanent, easily-retrievable record of all firewatch rounds that were completed.

The inspector discussed this event with the licensee and determined that the actions taken by the licensee appear adequate to preclude a recurrence of the event.

Since this event caused a

failure to comply with the requirements of Technical Specifications 3.7.6 and 3.7.7, a violation of the plants Technical Specifications occurred.

This item is unresolved pending NRC review of appropriate regulatory action (UNR 90-20-01) (Common).

8.1.2 i nifi nt eratin c rr n e Re rt SOORs are provided for problem identification and tracking, short and long term corrective actions, and reportability evaluations.

The licensee, uses SOORs to document and bring to closure problems identified that may not warrant an LER.

The inspectors reviewed the following SOORs during the period to ascertain whether:

additional followup inspection effort or other NRC response was warranted; corrective action discussed in the licensee's report appears appropriate; generic issues are assessed; and, prompt notification was made, ifrequired:

Unit 1 51 SOORs inclusive of 1-90-299 through 1-90-349 ((nit 2 14 SOORs inclusive of 2-90-128 through 2-90-141 The following SOORs required inspector followup:

1-90-300, documented an unplanned start of the "B" SGTS.

This event was reviewed in Section 2.2.3.

1-90-343, documented a contaminated injured individual. This event was reviewed in Section 3.2.1 and.1.3 Queen It~em l

NR

-

7/

-2 -

1 Inv lid Post Accident R ctor Buildin Air Tem eratur An

i On December 13, 1988, the licensee made a 10 CFR 50.9 report to the NRC Region I office concerning invalid post accident air temperature analysis used as the basis for their reactor building equipment environmental qualification (EQ) program.

The new analysis concluded that the bulk mean reactor building air temperature peaked at 123 degrees F instead of 104 degrees F.

Thus, exposing safety related equipment in the reactor building to temperatures greater than assumed in the Final Safety Analysis Report (FSAR).

The licensee identified the affected equipment and EQ binders that would be subject to these higher reactor building temperatures.

Licensee Corrective actions relative to the equipment in EQ binders EQDF-33 and 34 were reviewed during an NRC EQ inspection in August 1990.

The inspection revealed that two components (Bailey cards) remained to be relocated to a mild environment.

Results of this inspection are documented in inspection reports 50-387/90-17; 50-388/90-17.

The Unit 1 Bailey card was relocated with installation being completed in October, 1990.

Unit 2's Bailey card is scheduled to be relocated during the fourth refueling outage scheduled to begin on March 9, 1991.

A justification for Interim Operation was developed to establish equipment operability pending final resolution.

Based on the recent NRC EQ inspection results, the licensee's proposed schedule to relocate the Unit 2 Bailey card to a mild environment and justification for interim operations, the inspector determined these actions to be appropriate to resolve the items in binders EQDF-33 and 34.

This item is closed.

8.1.3.2 lo NR

7-

- I FireIn the "B" ff a uard Bed In April 1987, the "B" Offgas Guard Bed experienced a temperature excursion.

Internal temperatures indicated 1100 degrees F and surface temperatures measured 450 degrees F indicating a fire in the offgas charcoal bed.

The licensee identified several potential causes which may have resulted in the "B" Offgas Bed charcoal fire. The probable scenario for the ignition of the absorber bed is that during the purge cycle, the absorber is heated with air flowing at 24 Feet Per Minute at 360 degrees F inlet temperature.

This is a high enough temperature to start oxidation on the carbon.

Iron oxide contamination was found on several samples which lowers the ignition temperature of carbon.

These items coupled with a small amount of platinum contamination may have caused a localized hot spot which resulted in the ignition of the carbon be The licensee has implemented the following corrective actions to prevent recurrence.

The Offgas recombiner catalyst was changed from a ceramic base to a metallic base, per EWR-M-70660, to eliminate potential carryover of platinum dust, Operating procedures were revised to include cautions regarding the possibility for charcoal to be ignited while purging a guard bed with heated air.

Guard bed purge air temperature has been procedurally limited to 250 degrees F.

Guard bed purge criteria was procedurally changed from 8 inch Water Column (WC) differential pressure to 4 inch WC differential pressure.

Humidity sensors and transmitters, with condensation type dewpoint monitors at the guard bed inlet and outlet, were replaced to enable the determination of moisture content of the guard beds.

These corrective actions were proven to be adequate, since no additional fires have occurred in the Offgas charcoal beds.

Therefore, this item is closed.

8.1.3.3 Closed Other 87/87-0 -0 Deficiencies Noted In Peri dic hannel Check urveillance Pr cedures and Labelin In May 1987, several deficiencies were identified with the periodic instrumentation channel check surveillance procedures and component labeling.

The licensee implemented corrective actions to address these deficiencies.

These corrective actions included revising procedures and replacing missing labels on components or instruments.

The following procedures were reviewed to verify that appropriate corrective actions were implemented:

SO-100-006, Revision 9, Shiftly Surveillance Operating Log and SO-100-007, Revision 12, Daily Surveillance Operating Log. The procedures were found to adequately correct the identified discrepancies.

The licensee determined the root cause to be either an incorrect identification of instruments on appropriate panels during the initial development of th'e procedures or typographical errors.

The instrument and component labeling deficiencies were resolved during the licensee's ongoing labeling upgrade program.

This item is considered closed.

I 8.1.3;4 lo ed NV4

-

-1

- 2 mm n F il re Pr erl ecurea N rmall ked Valve The air to diesel engine valve F034058D, a normally locked valve, was found unlocked, although the valve was in its correct position.

The lock and chain were found draped over the valve with the lock unlocked rather than securing the valve as required.

The inspector reviewed licensee actions and verified that appropriate personnel were given additional training on AD-QA-303, "Shift Routine" concerning properly locking valves.

This topic was addressed as part of Supervisor's Operating Agenda training with the last operating shift's training completed on January 1, 1990 and it was verified that all operating personnel received this training on January 30, 1990.

In addition, the inspector sampled a large number of valves in Unit 1 and found them to be properly locked and in the correct position.

This item is close.

8.1.3.5 l sed UNR mm n M IV Jet Im in emen Anal si In May 1986, an inspector was informed of deficiencies related to jet impingement effects on the MSIVs during a recirculation line break.

Licensee analysis showed that the jet impingement forces on the "A" and "D" inboard MSIVs exceeded the forces to which the valves were qualified which was contrary to 10 CFR 50 Appendix A General Design Criteria (GDC) Four.

This nonconformance was documented in NCR 86-0019.

After further evaluation, a Safety Evaluation Report (SER) was written and reviewed by PORC.

Since no unreviewed safety question existed, PORC approved interim operation.

The licensee believed that they could justify their position long-term by providing a "leak-before-break (LBB)

analysis until the Unit 1 Third Refuel Outage and this provided the basis for PORC approving interim operation.

The licensee requested exemption from GDC Four based on LBB in a letter (PLA-2744) to the NRC.

The NRC raised several questions on the scope and assumptions of this methodology.

Thus, the licensee chose to offer an alternative action plan without reliance on LBB analysis.

This analysis currently uses dynamic loading comparisons to evaluate jet impingement effects.

NRC has found this type of analy'sis acceptable at other utilities. The continued acceptability of the licensee's approach is being evaluated by NRC Licensing and its completion status is being tracked by TAC No. 77757 and 77758.

Since the continued acceptability of the licensee's approach is being reviewed by NRC Licensing and involves no direct inspection effort, this item is closed.

Further licensee action may be required to satisfy NRC concerns.

8.1.3.6 l

ed

-21-2 mm n Inad te Fire Pr tecti n for mm ni ti n em Need f r n r l true re Fire A team inspection was conducted in December 1988 to review licensee activities regarding safe'shutdown capabilities in the event of a fire. During the inspection, the team noted that the licensee's alternate shutdown procedure, indicated that radio communication willbe required for use by the operators in the event of a Control Room fire requiring evacuation and shutdown from outside the Control Room.

However, the licensee did not include the communications circuitry within the scope of the Appendix R analysis.

While several communication systems may be available, the ability of one sole communication system to perform its function in the event of a Control Room fire was not conclusively established.

The licensee initiated NCR Number 88-0736 dated December 6, 1988 to resolve this issue.

The licensee evaluated many options in determining how to best provide fire protection for this communication system.

These options included upgrading existing systems or the installation of a new system.

The licensee decided to install a separate and independent system.

Installation for this system for Units 1 and 2 was completed per Design Change Packages 88-3017 and 88-3018 on December 28 and 29, 1988 respectively.

NCR 88-0736 was closed after the installation was determined to fulfillits design requirements.

Therefore, this item is close.

MANAGEMENTAND EXIT MEETINGS 9.1 R

ine Re i ent Exi n

P ri ic M tin The inspector discussed the findings of this inspection with station management throughout and at the conclusion of the inspection period.

Based on NRC Region I review of this report and discussions held with licensee representatives, it was determined that this report does not contain information subject to 10 CFR 2.790 restrictions.

9.2 Pl nt Di cre anc Man emen Meetin A management meeting was conducted between the NRC and the licensee on October 26 at the NRC Region I Office, to discuss licensee initiatives in the area of discrepancy management.

This meeting was held at the licensee's request.

The licensee presented the material provided in Attachment 2 during the meeting.

The licensee's goal for closure of each NCR, EDR, or SOOR willbe one refueling cycle.

The NRC also noted that the licensee plans to reorganize NPE to actively reduce the number of outstanding discrepancies.

They have created a new group of Engineering Projects that reports directly to the manager of NPE.

There are five groups that make up the Engineering Projects organization.

Two of these groups (Discrepancy Management and SSFI Closeout)

willbe assigned the task of reducing the backlog of discrepancies.

The NRC voiced two specific concerns to the licensee during the meeting.

The first concern involved providing adequate staffing for the new organization.

The NRC was concerned with the licensee's use and balance of contractor resources in the new organization.

The second concern involved the licensee's threshold for reporting engineering discrepancies.

The NRC had noted improvements in the reporting of engineering issues, but was not completely satisfied that the licensee was properly reporting these issues.

Further inspections are scheduled to review the licensee's reporting threshold.

9.3 Attendance at Mana ement Meetin nducted B Re ion Ba ed In ector Ins ec ion Re rtin Dates s

$ub~ec

~Rp~rt N )

~Ins ector 10/19 10/19 Maintenance Health Physics 90-81; 90-81 90-22; 90-22 D. Caphton R. Nimitz

ATTACHMENT1 A

reviati n Lis AD ADS ANSI CAC CFR CRD CREOASS DG DX ECCS EDR EP EPA ERT ESF ESS ESW EWR FO FSAR ILRT JIO LCO LER LLRT LOCA LOOP NCR NDI NPE NPO NRC OI PC PCIS PMR PORC QA RCIC RG RHR RHRSW

- Administrative Procedure

- Automatic Depressurization System

- American Nuclear Standards Institute

- Containment Atmosphere Control

- Code of Federal Regulations

- Control Rod Drive

- Control Room Emergency Outside AirSupply System

- Diesel Generator

- Direct Expansion

- Emergency Core Cooling System

- Engineering Discrepancy Report

- Emergency Preparedness

- Electrical Protection Assembly

- Event Review Team

- Engineered Safety Features

- Emergency Switchgear System

-.Engineering Service Water

- Engineering Work Request

- Fuel Oil

- Final Safety Analysis Report

- Integrated Leak Rate Test

- Justifications for Interim Operation

- Limiting Condition for Operation

- Licensee Event Report

- Local Leak Rate Test

- Loss of Coolant Accident

- Loss of Offsite Power

- Non Conformance Report

- Nuclear Department Instruction

- Nuclear Plant Engineering

- Nuclear Plant Operator

- Nuclear Regulatory Commission

- Open Item

- Protective Clothing

- Primary Containment Isolation System

- Plant Modification Request

- Plant Operations Review Committee

- Quality Assurance

- Reactor Core Isolation Cooling

- Regulatory Guide

- Residual Heat Removal

- Residual Heat Removal Service Water

Attachment

RPS RWCU SGTS SI SO SOOR SPING TS TSAS TSC WA 2.

- Reactor Protection System

- Reactor'ater Cleanup

- Standby Gas Treatment System

- Surveillance Procedure, Instrumentation and Control

- Surveillance Procedure, Operations

- Significant Operating Occurrence Report

- Sample Particulate, Iodine, and Noble Gas

- Technical Specifications

- Technical Specification Action Statement

- Technical Support Center

- Work Authorization

~

~

l U.S. NUCLEAR REGULATORY COMMISSION NRC Form 766 (Substitute)

Principal Inspector:

S. BARBER Reviewer:

P.

WETLAND INSPECT R'S REPORT

~EI A

lid d

D cket Ins ection Iayldd=>>

(A)

50- 88/ 0-21 (B)

Penn lv ni P wer &Li ht X I-Insert 2 N rth Ninth treet M - Modify Allent wn P

1

D - Delete R - Release Organization Code I

I:I I

I I '~I From To 10/07/90 11/03/90 Region Office Staff

~Re ion Division Branch X

Resident Inspector(s)

Performance Appr. Team I

B Other B

I* I:

Tyd IA 'yd d

Eid Olyy NRC Form 591 X

Regional Office Letter X 02-Safety 07-Special 12-Shipment/Export 03-Incident 08-Vendor 13-Import 04-Enf.

09-Mat.Acct.

14-Inquiry 05-Mgmt.

10-Plt.Sec.

15-Investigation Audit Ins tion Findin Le er f Re ort Transmi tal Date Zd77</

S&i7 />jr/q~

ABCD NRC Form 591 or Report Sent to iL X X

- Clear

- Violation

- Deviation

- Violation & Deviation Total No. of Violations and

Enforcement Conference

~HI NA Report Contains

Inf rmati n NO

NRC Form 766 (Cont.)

(Substitute)

M D LEINF RMATI N-A Phase/

Direct Module Insp.

N[g

~Hr Percent Q~m~le g Phase/

Module

~ta u No Direct Insp.

,

Percent

~Hr

~Com tete Status 5/30703

5/35702

5/61726

5/62703

5/90712

5/92700

100

60

85

C 5/71707 C

5/92701 (71707)

C 5/92720 C

5/60710 C

C 28.5

17

90

85

D EINF RMATI N-B Phase/

Direct Module Insp.

~N

~Hr 5/71707 36.5 5/90712

5/92720

5/61726

5/92700

Percent Q~m~le g

85

60

Phase/

Module

~tt

~N C

5/92701 (71707)

C 5/35702 C

5/62703 C

C Direct Insp.

~Hr

10 Percent Q~m.delete

~tatu

C

80

gal TANDIN ITEM FILE IN LE D KET ENTRY F RM ARY NRC Form 6 (Substitute)

Docket/Report No.: agggaag21 D 'gi BARBER gggggllg-21

'eport Period: +1/~79/ - ~11 Qg REP RT HO N

F FINDINGS g

i i gg g i: gWBTLANO

/F/D Unitl Unit2 1. Operations

~4" (PLT)~~

2. Rad-Con (RCP)

3. Maintenance 12.0(MNT)~1

~

~

~

4. Surveillance

~12 (SUR)

5. Emerg. Prep.

(EPP)

6. Sec/Safeguards (SPP)

Unitl Unit2 7. Engr. /Tech.

Support (DCM)

8. Safety Asses./

Verification ~7(AOQ)~2 9. Fire Prot./

Housekeeping (FDP)

10. Other(Option)

TOTAL:

~11

~7k,"l umm ta emen racke (12 haracter:

ROUTINE RESIDENT MONTHLY INSPECTION Checklist:

X OI History Updated X

OI Open Items Updated X

766 Complete

~ATER 2515 Schedule Updated Reactor Trip/LER Entries Needed Besides OI Out of Division Concurrences:

ecti n hief/Individual Re Prarah/

i n

~lai ial/Date

t

NRC Form 6 (Cont.)

(Substitute)

TANDIN ITEM FILE IN LE D CKET ENTRY F RM

~NTI D

Item No.

$AL~Pgde Are~gde Re onsi li 387/90-20-01 NON D 'DII 'SIS 01/14/90 P2A

~ri ~~in <~tr BARBER De cri tion (12 character:

FIREWATCH ROUNDS NOT COMPLETED PER TS

~

~

~

~

3.7.6 and 3.7.7.

JlelmN 387/88-20-01 UNR

$AL~PgcLe Are~@de Res on i ii P2A Due D te/Ltr Date~

D te M/ / Event Date*

10/18/90 I

IS

~Ori iaaior:

BARBER De cri i n (12 cha ct r: INVALADPOST ACCIDENTTEMPERATURE ANALYSI TA I

ITEM FILE SIN LE DO KET ENTRY FORM D

NRC Form 6 (Cont.)

(Substitute)

Item No.

387/87-09-01 UNR

~SALP ode

~Area de Res onsi lit P2A D

Ql t II D

~rd dr 10/18/90

~ri inator BARBER De cri tion (123 characters:

FIRE IN THE "B" OFFGAS GUARD BED (UNIT 1).

~item N T~

387/87-09-03 OTH QAL~Pgde

~Area de Res onsiblit P2A Due Dat Ltr Date* D e

M Event D te*

e rt d te/S tu 10/07/90 C

~ri inator BARBER Descri tion (123 ch ra er:

REVIEW OF PERIODIC CHANNELCHECK SURVEILLANCES - PROCEDURAL AND LABELINGDEFICIENCIE T TANDIN ITEM FILE IN LE DO KET ENTRY F RM

~CD IRC NRC Form 6 (Cont.)

(Substitute)

~Iem N Rm

~ALP ode

~Area ode Res onsiblit 387/89-18-02 NV5 D

D/

IC/I D

R~d/d MPS Qrinin~at r BARBER De cri i n

<12 cha c er;

"D" EDG STARTING AIR SUPPLY VALVEFOUND UNLOCKED. (COMMON)

~Item N 387/86-09-03 UNR JALAP

~Ar ~@de Re ni lit P2A d

10/07/90

~ri ~in Lttr'ARBER Descri ti n

<12 character:

MSIV JET IMPINGEMENTANALYSIS. (COMMON)

T TANDIN ITEM FILE IN LED KETENTRY F RM C NTI D

NRC Form 6 (Cont.)

(Substitute)

I~tem N

~T 387/88-21-02 UNR

$AL~P~Le Are~gde Res onsi lit PSS M

.

D IM/ I/E

~d

~ri ~in fear ANDERSON De cri ti n (12 ch racter:

COMMUNICATIONSYSTEM CONCERNS DURING A E

~

CONTROL ROOM FIRE. (COMMON)

~Item N JALAP@@

Are~a@le Re onsi lit R

rt e/Statu Qrigin;Lttr'escri ti n (12 cha t r:

ATTACHNENT 2 Reportability Meeting with NRC Ob ective of Meetin:

To reach agreement on how PP&I. shouId proceed with reporting in the interim untiI new NRC guidance is issued.

A enda

~ Definition of safety significance.

~

Recent PP8cL letter.

Basis for PP8rL position.

~ NRC perspective.

~

Review of past reportability determinations.

~ Control Structure Chilled Water a

MSIV Closure

~ Delta-T concerns'

Other issues

~

Improvements/Changes to PP8'cL proces Safet Si nificance All discrepancies will be initially screened for safety significance using the following questions as guidance.

Does the discrepancy appear to adversel im act a

system or component explicitly listed in the Tech Specs?

Does the discrepancy appear to corn romise the ca abilit of a system or component to perform as described in the FSAR?

Does the discrepancy appear to adversely impact any applicable licensing commitments?

If the answer to any of the above questions is yes, PP&L will expeditiously initiate operability and reportabih'ty evaluati on Determinin Safet Si nificar ce PAL Direction

~

EDRs and SOORs receive thorough review for safety significance.

~

NCRs have not received centralized review for safety significance.

~

A11 discrepancy management programs will involve a thorough review of issues for safety significance using formal established criteria.

Process

~

Deficiencies will be identified, documented, and evaluated for safety significance.'

Operability will be promptly determined and appropriate compensatory actions taken.

Timely action will be taken to resolve issue regardless of operability, with emphasis on safety significance of issue.

=-

~

Safety significance will consieler consequence of failure rather than potential for failure.

~

Conclusions will be thoroughly documented.

~

Reportability will proceed independently from operability determination 'I

Interim Re ortin Process

~

PP8cL will focus on safety.

Our threshold for 50.72/50.73 will be lowered.

~

We will take prompt corrective action.

~ A dedicated team will preside over an improved EDR process.

~

Issues reported under 50.9 will receive same level of attention as issues reported per 50.72/50;73.

~

Issues may begin as 50.9'nd evolve into LERs.

~

NCRs will get more consistent and extensive review for reportability.

~

When ie doubt, PP8cL will, review issues with Senior Resident to

'obtain feedback on appropriate reporting mechanis Re ortabilit Determinations Sub ect How Re orted New Criteria Emerg. Switchgear Room Cooling 50.9 50.72/50.73 USQ: Consequences of Malfunction MSIV Closure Originally Not Reported Voluntary LER 50.72/50.73 USQ: Probability of Malfunction Steam Leak Detection 50.9 50.9: Significance does not meet USQ threshold

Emergency Swi tchgear Room Cooling

~Dt

'uring a DBA, cooling for Vnit

emergency switchgear room is provided by Control Structure Chilled Water instead of Reactor Building Chilled Water.

A new single failure mechanism was discovered that could preclude proper operation of CSCW for ESWGR room cooling following a DBA.

Ori inal Re ortabilit Evaluation Based on the ability to detect, analyze, and react to this postulated single failure, it was concluded that no immediate operability issue existed and that the issue was not reportable under 50.72/50.73.

A 50.9 re ort a

p w s made.

Reevaluation Usin US Criteria

~ Increase the probability of 'ccurrence or consequences of an accident or malfunction evaluated in FSAR?

Yes; subsequently determined that ability to detect and react was inadequate.

Yes; results in potential loss of multiple safety systems.

~ Reduce.

the margin of safety as defined in the basis for Tech

~ Specs?

Yes; potential loss of multiple safety systems reduces margin of safety for long - term cooling function.

~ Create the possibility for a different type of accident?

Reportable per 50.72/50.73

MSIV Closure D~iti Based on NRC Information Notice 88-51, PPErI.

investigated and determined that the inboard MSIVs would not fully close without pneumatic-assist under design containment pressure conditions.

The FSAR indicates that MSIVs will close with pneumatic and, or spring force.

Ori inal Re ortabilit Evaluation Evaluation showed that under DBA conditions, actual containment pressures would be lower than design.

As a result, the valves would actually shut with either pneumatic or spring force.

Reevaluation Usin US Criteria

~ Increase the probability of occurrence or consequences of an accident or malfunction evaluated in FSAR?

Yes; loss of diversity in component design is an apparent increase in probability.

~ Create the possibility for a different type of accident?

No.

~'educe the margin of safety as def ined in the basis for Tech Specs?

No.

Reportable per 50.72/50.7 Steam Leak Detection D~t's a result of continuing reviews resulting from, Reactor Building temperature concerns, numerous steam leak detection issues were raised:

~

Room isolations would occur with 25gpm leaks rather than Sgpm leaks as stated in the FSAR.

~

Backdraft isolation dampers would isolate too early to permit actuation of system isolation due to steam leak. (Later found to be acceptable)

~

Leaks in Main Steam

'Ibnnel can be masked by coolers.

Ori inal Re ortabilit Evaluation Early evaluations were extremely conservative and overstated the potential consequences.

As better information was developed, the consequences were determined to be minimal.

Due to the number of issues and uncertainty, reporting under 50.9 was appropriate.

Reevaluation Usi n US Criteria

~ Increase the probability of occurrence or consequences of an accident or malfunction evaluated in FSAR?

No.

~

Create the possibility for a different type of accident?

No.

~ Reduce the margin of safety as defined in the basis for Tech Specs?

No.

Not reportable per

$ 0. 72/50. 73

Re ortabilit Determinations Sub'ect How Re pried New Criteria Electr ical Distribution Originally 50.9 Later reported as LER 50.72/50.73:

USQ: Consequences of malfunction Degraded Grid 50.9 50.9: Significance does not meet USQ threshold Dynamic Quals Racked-Out Breakers 50.9 50.9: Significance does not meet USQ threshold

Electrical Distribution D~iti In preparation for the EDSFI, several electrical concerns were raised.

Specif ic single failures were found to affect load shedding following certain events.

This had the potential to reduce voltages in one case and to overload diesels in another case.

Ori inal Re ortabilit Evaluation The first concern was evaluated to result in higher loads

'on a

4SOV MCC.

However, calculated voltages were still acceptable.

The second concern could result in higher loads on the diesels following an event.

Additional calculations were needed to assess the impacts of these loads.

This was reported to NRC per 50.9 on 5/25/90.

Subse vent Re ortabilit Evaluation On 7/20/90, PP&L concluded that failure of a 125V DC battery channel coincident with or just prior to a

LOCA and a

LOOP could result in overloading of an emergency diesel.

This independent failure had to occur shortly before the LOCA/LOOP since loss of the battery channel would result in a 2-hour LCO with hot shutdown required within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

If the failure occurred a fraction of a second after the LOCA/LOOP, load shedd ing wou ld proceed normally.

Th is was reported to NRC per 50.72/50.7 Re ortabilit Determinations Sub ect.

How Re orted New Criteria Reactor Building Temperature Diesel Inlet Air Temperature Control of Heavy Loads 5.9 (Original diesel failures vvere 50.72/50.73)

50.9 50.72/50.73:

Consequences of Malfunction.

'0.9:

PP8cL does not believe that this represents the cause of ihe diesel failures.

50.9: Significance does not meet USQ threshold

Re ortabilit Determinations Sub'ect How Re orted New Criteria Degraded Grid Issues 50.9 50.9: Significance does not meet USQ threshold Limitorque MOVs Not reported Not reportable.

Reactor Water Cleanup F001 Originally not reported Later reported as LER 50.72/50.73: PP&L was confused by prior NRC letter

Conclusions

~

Discrepancy management process is rapidly evolving.

~

Improvements will be made in processing of discrepancy documents.

~

Improvements will be made in documenting decisions and their bases.

~

We will focus on safety significance.

~

We will lower our threshold for 50.72/50.73.

~

We will achieve closure of discrepancies within one cycl t

~

0