IR 05000387/1998011

Insp Repts 50-387/98-11 & 50-388/98-11 on 981013-1123.No Violations Noted.Major Areas Inspected:Operations,Maint, Engineering & Plant Support,Including Assess Controls to Radiologically Controlled Areas

ML17164A937
Person / Time
Site:	Susquehanna
Issue date:	12/24/1998
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML17164A936	List: IR 05000387/1998011 ML17164A935
References
50-387-98-11, 50-388-98-11, NUDOCS 9901050095
Download: ML17164A937 (46)

Text

U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Docket Nos:

License Nos:

50-387, 50-388 NPF-14, NPF-22 Report No.

50-387/98-1 1, 50-388/98-1

Licensee:

Pennsylvania Power and Light Company 2 North Ninth Street Allentown, Pennsylvania 19101 Facility:

Susquehanna Steam Electric Station Location:

P.O. Box 35 Berwick, PA 18603-0035 Dates:

October 13, 1998 through November 23, 1998 Inspectors:

S. Hansell, Senior Resident Inspector S. Barr, Senior Resident Inspector J. Richmond, Resident Inspector A. Blarney, Resident Inspector J. McFadden, Radiation Specialist S. Dennis, Operations Engineer Approved by:

Clifford Anderson, Chief Projects Branch 4 Division of Reactor Projects 990i050095 981224 PDR ADOCK 05000387

PDR

EXECUTIVE SUMMARY Susquehanna Steam Electric Station (SSES), Units 1 5 2 NRC Inspection Report 50-387/98-11, 50-388/98-11 This integrated inspection included aspects of Pennsylvania Power and Light Company's (PPRL's) operations, maintenance, engineering, and plant support at SSES.

The report covers a 6-week period of resident inspection; in addition, it includes the results of an announced inspection by a regional radiation specialist.

~Oerations Plant operator building rounds were thorough and properly maintained.

Communications between control room and field operators were good, operations personnel were knowledgeable of their responsibilities and control board awareness was good.

However, the inspectors observed inconsistencies in the numb r of identified leaking safety relief valves on multiple Plant Control Operator and Unit Supervisor turnover sheets during the week of October 19, 1998, and operations personnel were unsuccessful in starting the "A"Turbine Building Filter Exhaust Fan because they were unaware that a blocking permit was still in effect on the fan dampers'ir supply. (Sections 01.1 and 01.3)

The observed operator performance during the October 9, 1998, startup was good.

Changes made to plant procedures in response to a previous event were effective in minimizing operator distractions and resulted in improved control of core reactivity.

However, operators were challenged by minor equipment problems and discrepancies between training and actual operation of nuclear instrumentation.

(Section 02.1)

PPS.L identified several problems associated with the implementation of Technical Specifications (TS), including a missed surveillance test (drywell floor drain sump level instrumentation),

a PPSL determination that prior surveillances for the source range monitors were not adequately performed, and an instance where a more conservative TS Interpretation was not recognized.

In each case, the inspectors concluded PPRL took prompt and effective initial corrective actions. (Section 03.1)

PPSL identified that four eighteen month logic system functional feedwater/main turbine trip system actuation instrumentation tests, required by Technical Specifications, were missed.

PPSL's corrective actions, including procedure and programmatic actions, were good. This non-repetitive, licensee identified violation is being treated as a non-cited violation, consistent with Section VII.B.1 of the NRC Enforcement Policy.

LER 50-387/98-010is closed.

(Section 08.1)

Executive Summary (cont'd)

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The failure to perform a monthly channel functional test on four Suppression Pool Temperature Monitoring System alarm relays was a violation of Technical Specification 4.6.2.1.c.2, functions 3a, b, c, and d surveillance requirements.

This licensee-identified and corrected violation is being treated as a Non-Cited Violation, consistent with Section VII.B.1 of the NRC Enforcement Policy.

LER 50-387/98-012 is closed.

(Section 08.1)

The failure to establish a continuous fire watch within one hour of halon system inoperability was a violation of Technical Specification 3.7.6.4 action "a" requirements.

This non-repetitive licensee-identified and corrected violation is being treated as a Non-Cited Violation, consistent with Section VII.B.1 of the NRC Enforcement Policy.

LER 50-387/98-013is closed.

(Section 08.1)

PP&L identified, in a Licensee Event Report, that a one hour fire watch'was established, instead of a continuous fire watch, as required by Technical Specifications.

PP&L's corrective actions, including procedure and programmatic actions, were good.

This non-repetitive, licensee identified and corrected violation is being treated as a non-cited violation, consistent with Section VII.B.1 of the NRC Enforcement Policy.

LER 50-387/98-015is closed.

(Section 08.1)

Maintenance The inspectors concluded that the PP&L method to assess plant risk for on-line/emergent work, by reviewing the work in accordance with the current revision of the Susquehanna Team Manual and quality assurance procedures, meets the intent of the maintenance rule. (Section M1.2)

Operators and maintenance technicians responded properly to numerous equipment problems.

Previous analysis and completed corrective actions have not been fully effective at preventing recurrence of some equipment problems.

Examples included repetitive instances of Unit 2 containment instrument gas header pressure loss due to a stuck open check valve, "C" emergency diesel failure to start, and a repetitive steam leak on reactor feedwater pump turbine steam supply valve HV-12710C requiring a second on-line leak seal repair.

However, PP&L management's proactive response to the RHRSW pump shaft failure aggressively resolved this potential common cause failure. (Section M2.1)

~En ineerin

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The inspectors concluded that the lack of an adequate feedwater loop seal'existed since initial plant startup.

Once the lack of an adequate loop seal and other related issues were documented in condition reports, PP&.L performed thorough operability determinations.

PP&L's completed and planned corrective actions were well planned and thorough.

(Section E8.1)

Executive Summary (cont'd)

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PPRL identified inconsistencies within the FSAR and Technical Specifications for the feedwater containment boundary penetrations.

These inconsistencies resulted in PP5L not requesting an exemption from 10 CFR 50 Appendix J testing.

PPRL's completed corrective actions and scheduled corrective actions were thorough and complete.

This non-repetitive, PP&L identified and corrected violation is being treated as a Non-Cited Violation, consistent with Section VII.B.1 of the NRC Enforcement Policy. (Section E8.1)

Plant Su ort

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PPSL implemented effective radiological controls at SSES.

Access controls to radiologically controlled areas were effective, appropriate occupational exposure monitoring devices were provided and used, personnel occupational exposure was maintained within applicable regulatory limits and As-Low-As-Reasonably-Achievable (ALARA),and the radiation work permit program was implemented properly for control of radiological work. (Section R1.1)

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PPtkL implemented overall effective surveys, monitoring, and control of radioactive materials and contamination.

The surveys, monitoring, and controls were performed with calibrated and properly used devices.

Personnel and area contamination rates were properly tracked and trended.

(Section R1.2)

PP5L's self-identification and corrective action processes in the area of radiation protection were effective: Quality Assurance surveillances, corporate assessments, and self-assessments continued to be effective in identifying, at a low threshold, deficiencies and improvement opportunities.

Corrective actions were implemented for findings.

(Section R7)

TABLEOF CONTENTS EXECUTIVE SUMMARY TABLEOF CONTENTS

. v Summary of Plant Status I. Operations

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Conduct of Operations........ ~.........

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01.1 Unit Operations and Operator Activities

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01.2 Operational Safety System Alignment 01.3 Operations Shift Activities Operational Status of Facilities and Equipment 02.1 Unit 1 Plant Startup following Reactor Scram Operations Procedures and Documentation.......

03.1 Technical Specification Problems.........

Miscellaneous Operations Issues

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08.2 Followup of Open Items............

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M1 Conduct of Maintenance

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M1.1 Pre-Planned Maintenance and Surveillance ActivityReview M1.2 On-Line/Emergent Work - Risk Assessment...........

M2 Maintenance and Material Condition of Facilities and Equipment

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E8 Miscellaneous Engineering Issues.............

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V. Plant Support I

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I R1 Radiological Protection and Chemistry Controls.......

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R1.1 Radiological Controls - External and Internal Exposure R1.2 Radioactive Materials, Contamination, Surveys, and Monitoring R1.3 ALARAProgram R1.4 Radiation Protection Program Changes.................

R7 Quality Assurance in Radiological Protection Activities..... ~.....

S2 Status. of Security Facilities and Equipment...................

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~ 2 1 ATTACHMENT Attachment 1 - Inspection Procedures Used

- Items Opened, Closed, and Discussed

- List of Acronyms Used

Re ort Details Summar of Plant Status Susquehanna Steam Electric Station (SSES) Unit 1 conducted a plant startup at the beginning of this inspection period.

Reactor criticality was achieved on October 10, but reactor power was held at approximately 12% until October 15, to resolve a problem with the main generator synchronizing breaker.

On October 16, the unit reached full power.

On November 19, power was reduced to 98% to perform control rod scram time testing, following on-line Hydraulic Control Unit (HCU) maintenance.

Power was returned to 100%

on November 19 and remained at 100% power for the rest of the inspection period.

SSES Unit 2 maintained 100% power throughout the inspection period, except for the following power reductions.

On October 23, power was reduced to 83% to perform on-line maintenance of 12 HCUs and control rod scram time testing; power was returned to 100% on October 26.

On October 30, power was reduced to 65% to perform maintenance on isophase bus duct fans, then increased to 85% to perform on-line maintenance of 12 HCUs and control rod scram time testing; power was returned to 100%

on November 2.

On November 6, power was reduced to 83% to perform on-line maintenance of 6 HCUs and control rod scram time testing; power was returned to 100%

on November 8.

On November 15, power was reduced to 85% for a control rod pattern sequence exchange, then returned to 100%.

The unit remained at 100% for the rest of the inspection period.

I. 0 erations

Conduct of Operations

'1.1 Unit 0 erations and 0 erator Activities a.

Ins ection Sco e 71707 Routine activities of plant control operators (PCOs), nuclear plant operators (NPOs),

unit supervisors (USs), and shift supervisors (SSs) were observed.

b.

Observations and Findin s The inspectors determined that routine operator activities were properly prescribed, communicated and, in general, well performed in accordance with SSES operation department procedures.

Communication between PCOs and NPOs was observed to be of good quality.

Shift turnovers were observed to be generally detailed and complete, with one exception.

During a Unit 1 shift turnover on October 23, 1998, the inspectors observed inconsistencies in the number of identified weeping safety relief valves

'Topical headings such as 01, M8, etc., are used in accordance with the NRC standardized reactor inspection report outline.

Individuai reports are not expected to addross all outline topic (SRVs) on multiple PCO and US turnover sheets.

The SS and shift technical advisor (STA) stated six SRVs were identified as leaking (tail pipe temperature >200 degrees) since the last Unit 1 startup.

A review of turnover sheets for the week of October 19 found that PCOs had incorrectly identified only five leaking SRVs, while the USs had incorrectly identified only 4 leaking SRVs. This inconsistency in turnover sheets occurred over multiple shifts. This issue was discussed with operations supervision and corrected.

C.

Conclusions PP&L conducted plant operations in accordance with SSES procedures, operator activities were, generally well performed and communicated.

Shift turnovers were observed to be detailed and complete, with one exception.

The inspectors observed inconsistencies in the number of identified leaking safety relief valves on multiple PCO and US turnover sheets during the week of October 19, 1998.

01.2 0 erational Safet S stem Ali nment ao Ins ection Sco e 71707 b.

The Inspectors conducted walkdown inspections of selected safety systems, and observed equipment alignment and operability.

Observations and Findin s PP&L performed system operations in accordance with procedures, and effective controls were implemented for safe plant operation.

Overall equipment operability, material condition, and housekeeping conditions were good.

During plant tours, the alignment and operability of selected safety systems, engineered safety features, and on-site power sources were verified. A partial walkdown of the following systems was performed:

Unit Common Emergency Diesel Generators (EDG)

Unit 1 & 2 Control Rod Drive (CRD) Hydraulic Control Units (HCUs)

Unit 2 Residual Heat Removal System (RHR)

Unit 1 Core Spray Unit 1 Engineered Safety Features Instrument Racks The inspectors identified several minor housekeeping and material condition items, including a small condensate leak at the suction of the "1C" reactor feedwater pump. These items did not affect system operability and were communicated to PP&L management for review.

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Conclusions PP&L performed safety system operations in accordance with SSES procedures, and established effective equipment alignment and operabilit i 01.3 0 erations Shift Activities ao Ins ection Sco e 71707 The inspectors observed selected operations activities to determine whether approved procedures were in use, details were adequate, Technical Specifications (TS) were satisfied, shift turnovers were comprehensive, activities were performed by knowledgeable personnel, and surveillance testing was appropriately completed and documented.

Specifically, the inspectors observed portions of the following activities:

Unit 2 - Reactor Building operator rounds, November 17; Unit 2 - Reactor Steam Dome Pressure surveillance test - SI-258-303, November 17; Unit 2 - HPCI aux lube oil system functional test - OP-252-001, November 17; Common - "A" SBGT surveillance test - SO-070-001A, November 18; Unit 2 - RCIC tagout - November 18; Unit 1 - Reactor Building operator rounds, November 19; Unit 1 - Turbine Building operator rounds, November 20; Units 1 and 2 - morning shift turnovers, November 17, 18, and 19; Unit 1 - Turbine Building Filtered Exhaust System "A" return to service-November 20; b.

Observations and Findin s The inspectors determined that the plant operator building rounds in the turbine, reactor, and control areas were thorough and building log readings were complete and properly maintained.

Additionally, communications between control room and field operators were good during the observed evolutions, operations personnel were knowledgeable of their responsibilities, and high pressure coolant injection (HPCI), standby gas treatment (SBGT) and reactor steam dome pressure surveillance paperwork was completed accurately.

Control room operators board awareness was good, particularly during l&C,testing. Technical Specifications were referenced and followed as required.

The inspectors observed that following the Unit Supervisor's (US) direction to start the "1A"Turbine Building Filter Exhaust Fan (TBFE), the fan dampers would not realign.

Neither the US or Equipment Operator (EO) were aware that blocking permit (1-98-1816) was in place on the fan dampers air supply.

The pre-job brief observed by the inspectors indicated that the fan was bypassed with no mention of a blocking permit. The inspectors also noted that the blocking permit was not mentioned on any shift turnover documents.

PPRL has written condition report (CR) 85665 to address the proble Conclusions The observed activities were conducted using approved procedures, plant operator building rounds were thorough, and building log readings were complete and properly maintained.

Additionally, communications between control room and field operators were good, operations personnel were knowledgeable of their responsibilities, control board awareness was good, and TS were satisfied.

However, operations personnel were unsuccessful in starting the "A"TBFE fan because they were unaware that a blocking permit was still in effect on the fan dampers'ir supply.

Operational Status of Facilities and Equipment 02.1 Unit 1 Plant Startu followin Reactor Scram Ins ection Sco e 71707 Startup reactivity control activities performed by PCOs and USs were observed during the October 9, 1998, startup.

The observations included control rod manipulations to criticality through heat up conditions and'the investigation of the main generator synchronization breaker opening.

Observations and Findin s

PPSL's reactivity control was improved through changes to GO-100-002, "Plant Startup, heat up and Power Operation".

Opening the Main Steam Isolation Valves (MSIVs) prior to startup prevented delays during heat up and allowed the operators, to focus more clearly on reactivity management.

Ranging Intermediate Range Monitors was performed by one PCO; however, the PCO continued to utilize the computer display during heat up. The PCOs are not typically trained on the use of the computer display during startup.

The use of the computer display was identified by PPSL as a contributing factor to a previous reactor trip during startup.

During this startup the PCOs were challenged by several minor equipment issues.

The inspectors observed twenty-one control rod withdrawals, eight required additional operator actions, such as cycling the direction control valves, to withdraw the control rod from the fullyinserted position.

During generator synchronization the main synchronization breaker re-opened unexpectedly.

During investigation of the breaker problem the inspectors observed the in-field automatic synchronization breaker control relay testing, generator metering checks and Plant Operations Review Committee (PORC) test procedure review. These activities were well conducted by qualified personnel.

Ce Conclusions The observed operator performance during the October 9, 1998, startup was good.

Changes made to plant procedures in response to a previous event were effective in minimizing operator distractions and resulted in improved control of core reactivit However, operators were challenged by minor equipment problems and discrepancies between training and actual operation of nuclear instrumentation.

Operations Procedures and Documentation 03.1 Technical S ecification Problems (71707,40500)

Several procedural problems, observed during this inspection period, were related to inadequate incorporation of the Improved Technical Specifications into the Technical Specifications, including:

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Unit 1 5 2 drywell floor drain sump level instrumentation was declared inoperable, when operations determined the surveillance test frequency had not been performed monthly as required by TS. Prior to the change in TS this test was performed quarterly.

Surveillance tests were successfully performed prior to exceeding the TS allowed outage time. The surveillance procedure frequency was revised to monthly.

No violation of NRC requirements was identified.

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Unit 1 & 2 source range monitors (SRMs) were declared inoperable, when Instrument and Control technicians determined the existing surveillance test did not include the individual SRM channel indications in the test acceptance criteria, as required by TS (CR 76261).

Surveillance tests were revised to include the TS requirements, and successfully performed, PPS.L will issue a

Licensee Event Report (LER). This issue will be reviewed for any potential violations of NRC requirements during the LER review.

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The NRC identified PPSL had inappropriately entered a 30 day Limiting Condition for Operations (LCO) for planned maintenance on the "1A" RHRSW pump in accordance with TS section 3.7.1. TS allows a 30 day LCO with one RHRSW subsystem inoperable.

Technical Specification Interpretation (TSI) 1-97-007, reduces the LCO to 7 days because TS do not account for potential failures within the RHRSW subsystem flow paths.

PPRL exited the LCO within 7 days and improved references between TSI and TS.

No violation of NRC requirements was identified.

In conclusion, PPSL identified several problems associated with the implementation of TS, including a missed surveillance test (drywell floor drain sump level instrumentation),

a PPSL determination that prior surveillances for the source range monitors were not adequately performed, and an instance where a more conservative TS Interpretation was not recognized.

In each case, the inspectors concluded PPSL took prompt and effective initial corrective action P Miscellaneous Operations Issues 08.1 Licensee Event Re ort Review (71707,92700)

Closed LER 50-387 98-010-00 Feedwater/Main Turbine Trip Surveillance Inadequate PP&L identified, on four past occasions, that the required TS surveillance for the eighteen month logic system functional test of the feedwater/main turbine trip system actuation instrumentation including the. actuating device was not completed.

The closure of the reactor feedwater pump turbines steam emission valves and closure of the main turbine stop valves were not tested.

Based on in-field reviews of SSES TS, testing procedures and electrical schematics drawings, the inspectors determined that PP&L properly identified, corrected and reported the missed TS surveillances.

The Unit 1 procedures were modified to perform this testing and the Unit 2 procedures are scheduled to be modified prior to the next test.

Additionally, there were no identified occurrences of reactor feedwater pump turbines or main turbine high water level trip logic actuation failures. The inspectors concluded this condition had no significant impact on the system's ability to respond to a reactor vessel high water level during the four missed surveillances.

PP&L's proposed and completed corrective actions were good.

In conclusion, PP&L identified that four eighteen month logic system functional feedwater/main turbine trip system actuation instrumentation tests, required by TS, were missed.

PP&L's corrective actions, including procedure and programmatic actions, were good.

This did not represent a repetitive condition because it resulted from a single failure to establish an adequate surveillance procedure.

Therefore, this non-repetitive, licensee identified violation is being treated as a non-cited violation, consistent with Section VII.B.1 of the NRC Enforcement Policy. This LER is closed.

(NCV 50-387,388/98-11-01)

Closed LER 50-387 98-012-00 Suppression Pool Temperature Surveillance Testing Did Not Meet TS Suppression pool temperature surveillance testing did not meet the definition of a channel functional test.

On June 3, 1998, PP&L determined that the monthly TS surveillance requirements for the suppression pool temperature monitoring system (SPOTMOS), TS 4.6.2.1.c.2., functions 3a, b, c, and d had not been met since 1982, the initial issuance of the TS surveillance requirement.

PP&L determined that a SPOTMOS micro processor system integral self-test failed to check actuation of four alarm relays in the system as required by TS. The same relays were tested successfully during other 18 month channel calibrations.

PP&L concluded that successful completion of the 18 month test provided reasonable assurance that the relays would have always functioned as require The inspectors performed an onsite document review of the'licensee evaluation of the event and corrective actions.

The inspectors verified the successful completion of the previous 18 month channel calibration tests and PP&,L corrective actions which revised monthly procedure SI-1(2)59-203, SPOTMOS channel functional test.

Additionally, successful performance of the revised test and licensee review for similar computer self-testing issues were verified. The inspectors determined that the PP&.L evaluation of the event and corrective actions were appropriate.

The LER met the requirements of 10 CFR 50.73..

The failure to perform a monthly channel functional test on four SPOTMOS alarm relays was a violation of TS 4.6.2.1.c.2, functions 3a, b, c, and d surveillance requirements.

This licensee-identified and corrected violation is being treated as a Non-Cited Violation, consistent with Section VII.B.1 of the NRC Enforcement Policy.

(NCV 50-387,388/98-1 1-02)

Closed LER 50-387 98-013-00 Continuous Fire Watch Not Established within TS Time Limit On the evening of June 30, 1998, a door to the rod position indication system (RPIS) panel 1C615 was removed to cool components in the panel.

Removal of the panel door made the associated halon fire suppression system inoperable.

TS 3.7.6A action "a" stated that with one or more halon systems inoperable, within one hour establish a continuous fire watch. A continuous fire watch was not established until 4:35 pm on July 1, 1998, which was greater than one hour after the panel door was opened.

The inspectors performed an onsite document review of PP&L's evaluation of the event and verified subsequent corrective actions.

PP&L found that system engineering had incorrectly determined the operability of the halon system.

System engineering did not reference procedure NDAP-QA-0441

- Fire Protection System Status, which would have specified the requirements for halon system operability and fire watch requirements.

PP&L's evaluation "and corrective actions were appropriate and included issuing a revised fire protection system status change form, establishing a continuous fire watch, and reemphasizing the importance of following and checking procedures.

The failure to comply with TS had minimal safety impact in that the halon system remained functional and a CO2 backup system and associated fire detection were operable at that time. The LER met the requirements of 10 CFR 50.73.

The failure to establish a continuous fire watch within one hour of halon system inoperability was a violation of TS 3.7.6.4 action "a" requirements.

This non-repetitive licensee-identified and corrected violation is. being treated as a Non-Cited Violation, consistent with Section VII.B.1 of the NRC Enforcement Policy. This LER is closed.

(NCV 50-387,388/98-11-03)

Closed LER 50-387 98-015-00 Continuous Fire Watch Not Per Requirements of TS

PP&L identified that a one hour fire watch had been established, instead of a continuous fire watch, as required by TS 3.7.7.

Based on an in-field review of the issues reported in this LER, including TS, operating and surveillance procedures, and associated corrective actions, the inspectors found PP&L's proposed or completed corrective actions to be good.

The root cause of this event was determined to be an inadequate drawing change notice.

Notations to a drawing specified dampers and doors as TS related fire barriers, but did not remove, or clearly supersede, an existing note from a prior drawing change notice which conflicted with the new drawing change.

In conclusion, PP&L identified, in a Licensee Event Report, that a one hour fire watch was established, instead of a continuous fire watch, as required by TS.

PP&L's corrective actions, including procedure and programmatic actions, were good.

This non-repetitive, licensee identified and corrected violation is being treated as a non-cited violation, consistent with Section VII.B.1 of the NRC Enforcement Policy. This LER is closed.

(NCV 50-387/98-11-04)

08.2 Followu of 0 en Items (71707,92901)

Closed Unresolved Item URI 50-388 97-10-01 TS 3.0.3 Entry to Support Surveillance Activities This event was previously discussed in NRC Inspection Report (IR) 50-387, 388/97-10 and updated in NRC IR 50-387, 388/98-01.

PP&L actions were determined to be acceptable to permit completion of operability testing on equipment required for operation.

This action was needed to avoid a transient and was not merely for operational convenience.

No violation was identified, this URI is closed.

Closed IFI 50-387 388 98-01-01 Operator Re-Qualification Training Program The weaknesses in the operator license requalification program were previously discussed in NRC IR 50-387, 388/98-01. The inspectors reviewed PP&L corrective actions which included a revision of exam bank questions, static exam questions, and job performance measures (JPM). The inspectors verified that the actions were complete by reviewing a sample of 40 new questions and 20 JPMs.

The questions reviewed were at the higher order comprehensive/analysis level and the JPMs, including time critical and alternate path, had critical tasks properly notated.

Additionally, the inspectors interviewed an operations training supervisor who stated that alternate path JPMs will be incorporated into the upcoming (January 1999) requal exam JPM sets. The inspectors concluded that the corrective actions were timely and acceptable.

No violation was identified. The IFI is close II. Maintenance M1 Conduct of Maintenance M1.1 Pre-Planned Maintenance and Surveillance Activit Review a.

Ins ection Sco e 61726 62707 The inspectors observed and reviewed selected portions of pre-planned maintenance and surveillance activities, to determine whether the activities were conducted in accordance with NRC requirements and SSES procedures.

b.

Observations and Findin s Based on the indicated sample of safety related work authorizations and surveillances, the inspectors found pre-planned maintenance and surveillance activities were appropriately conducted and controlled.

The sample included:

Work Authorizations V82281

"B" Control Rod Drive Pump Seal Water Line Leak Repair P81669 2E230B Unit Cooler Inspect, Clean, and Eddy Current Test A81'227

"B" EDG 20 Year Overhaul Surveillances SO-024-001 SO-024-004 SO-030-001 SM-102-001 SO-256-001 SE-024-A01

"B" EDG Monthly Performance Test EDG Intercooler Emergency Service Water (ESW) Valve Exercise

"A" CREOASS Performance Test 125 VDC Station Batteries Weekly Electrical Parameter Check U1 5 Generator "E" Weekly Control Rod Exercising Diesel Generator "A" Integrated Surveillance Test In addition, selected portions of equipment permits (e.g., tagouts), procedures, drawings, and vendor technical manuals, associated with the maintenance activities, were also reviewed and determined to be acceptable.

In general, maintenance personnel were knowledgeable of their assigned activities.

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Conclusions Based on a sample of pre-planned maintenance and surveillance activities, the activities were found to be properly conducted and controlled by qualified worker M1.2

On-Line Emer ent Work - Risk Assessment Ins ection Sco e 62706 A weakness in PP&L's program for risk assessment, when taking equipment out of service for on-line maintenance, was described in NRC IR 50-387, 388/98-04.

The inspectors assessed PP&L's ongoing corrective actions in addressing this issue through interviews with two work week managers, a unit supervisor, and a review of the revised work management document.

Observations and Findin s The inspectors found that the Susquehanna Team Manual, which described the work management process, was revised on June 6, 1998, and September 24, 1998, and contained an expanded section which further addressed equipment out of service requirements.

The requirements, as stated, were based on the FSAR, TS, Technical Requirement Manual, and Maintenance Rule (MR) risk significance.

On-line maintenance and emergent work is now performed following a review in accordance with the Susquehanna Team Manual and a safety assessment in accordance with quality assurance procedures.

The work week manager in conjunction with operations shift personnel are also part of the review process.

Additionally, actions by PP&L to implements an online computer system'(ORAM/

Sentinel) risk monitor to support a rolling 13 week work schedule are ongoing and are expected to be completed in early 1999.,

Conclusions The inspectors concluded that the PP&L method to assess plant risk for on-line/emergent work, by reviewing the work in accordance with the current revision of the Susquehanna Team Manual and quality assurance procedure NEPD-QA-0900, meets the intent of the maintenance rule.

- N)2 Maintenance and Material Condition of Facilities and Equipment M2.1 Material Condition of the Facilit Ins ection Sco e 62707 The inspectors reviewed numerous equipment problems and assessed PP&L response.

Observations and Findin s PP&L continues to experience equipment problems that result in challenges to the operators and maintenance technicians.

On October 18 and October 19, 1998, PP&L experienced Unit 1 reactor coolant conductivity increases due to failures of the "B" RWCU purge pump diaphragm (CR 75903 & 75921).

On October 31,

1998, a third failure of a Unit 1 RWCU purge pump diaphragm occurred; however, operations isolated RWCU prior to a reactor coolant conductivity increase.

Several other degraded material conditions observed during this inspection period included:

Unit 1 rod drive control system failure resulted in an inability to move any control rod, except for a scram, due to an improper termination (an alligator clip) at an HCU directional control valve (CR 76078).

Both Unit 2 CRD pumps experienced repeated seal water leaks, requiring CRD pumps to be swapped and removed from service, for leak repairs, multiple times.

RWCU PCIV actuated due to demineralizer inlet high temperature, following a loss of RBCCW to the NRHX, during maintenance activities (Unit 1, CR 75919, 10-19-98).

"C" EDG failed to start during a monthly surveillance test.

After repairs, the EDG unexpectedly tripped during the cooldown run (CRs 75808, 75809).

Unit 2 experienced repeated containment instrument gas header pressure drops, following system maintenance, due to a stuck open check valve, resulting in an automatic swap over to the backup Automatic Depressurization System (ADS) bottle header (CRs 76264, 76485).

On-line leak seal repair was performed to HV-12710C "C" reactor feedwater pump turbine steam supply valve for a second time, the valve still has a minor steam leak.

Main generator'ync breaker failed to close during startup; reactor remained at 12% power on Bypass Valve (BPV) pressure control for 5 days (Unit 1, CR 98-3154, 10-11-98), see Section 02.1 for details.

The inspectors continued to review PP&L response to the "2B" Residual Heat Removal Service Water (RHRSW) pump shaft failure, which occurred during July 1998.

PPBcL responded by accelerating the preventative maintenance (PM)

overhauls for the remaining RHRSW and Emergency Service Water (ESW) pumps.

Three of the four RHRSW pumps shafts were replaced with new shafts less susceptible to stress corrosion cracking.

The last RHRSW pump is scheduled to be completed in December 1999. The ESW pumps are scheduled to be overhauled during the summer of 1999.

PPSL management's proactive response to the pump shaft failure aggressively resolved this potential common cause failure.

Conclusion Operators and maintenance technicians responded properly to numerous equipment problems.

Previous analysis and completed corrective actions have not been fully effective at preventing recurrence of some equipment problems.

Examples included repetitive instances of Unit 2 containment instrument gas header pressure loss due to a stuck o'en check valve, "C" emergency diesel failure to start, and a repetitive steam leak on reactor feedwater pump turbine steam supply valve HV-12710C requiring a second on-line leak seal repair.

However, PP5L management's proactive

response to the RHRSW pump shaft failure aggressively resolved this potential common cause failure.

III. En ineerin E8 Miscellaneous Engineering Issues E8.1 Followu of 0 en Items (37551,92903)

Closed URI 50-387 388 96-06-01 Feedwater Loop Seal PPS.L's offsite dose analysis limited Secondary Containment Bypass Leakage (SCBL)

to 5 scfh.

PPSL assumed that there would be no SCBL through the feedwater (FW)

penetration because a loop seal would exist in the FW penetration.

On January 15, 1996, PP&L identified that the FW containment boundary loop seal as described in the Final Safety Analysis Report (FSAR) was not achievable.

However, based on Local Leak Rate Test (LLRT) results, PPSL concluded that the total SCBL values for both units were below 5 scfh.

The inspectors determined and documented in IR 50-387,388/96-01that PPRL initial corrective actions were adequate.

Unresolved Item (URI) 96-06-01 was opened to track additional issues identified during investigation of the SCBL. These additional issues (IR 50-387,388/96-06) were subsequently found to be bounded by PPSL's licensing and design basis and are discussed below.

(a)

The post accident offsite dose calculations in the FSAR Section 15.6.5 assumed a 5 gpm leak into the secondary containment from engineering safety feature system pumps, seals and valves.

PPSL identified in CR 96'-

0504 that this analysis did not include leakage from other sources including the non-seismic CRD header.

PPSL evaluated this additional leakage not previously included in the FSAR (including CRD leakage) and determined that the total leakage was less than 5 gpm on each unit, Additionally, the licensee concluded that the CRD lines were seismically rugged and would not rupture.

Therefore, the plant was within the design basis.

(b)

CR 96-0522identified that the HPCI and Reactor Core Isolation Cooling (RCIC) turbine exhaust spargers will be uncovered in less than 30 days with a leak of 3.3 gpm and no suppression pool make up. Actual plant data indicated that leakage from the valves listed in the TS and FSAR tables was less than 3.3 gpm and would not uncover the HPCI and the RCIC turbine exhaust spargers.

In addition, PPSL received approval, through license amendment 149/ 119 to allow for makeup to the suppression pool.

Based on this amendment and the small amount of make up that would be required, the 30 day water seal credited in the FSAR is vali NRC IR 50-387,388/96-10documented that the inspectors reviewed the PPtkL plan to resolve these issues and concluded that the plan was thorough.

The inspectors documented in the report that they were unable to determine if the increase from 5 to 9 SCFH of leakage was appropriate for SCBL without further NRC review. The NRC has completed review of the proposed increase in SCBL and issued license amendments 151 /178 which specified a new SCBL limitof 9 scfh.

PPSL's corrective actions as delineated through PLA-4831 include modification of the FW penetration valves FW 7 A/B valve seat and replaced the FW 39 A/B valve to improve their leakage performance and incorporated these valves into TS as PCIV. This was completed during the Unit 1 tenth refueling outage in the spring of 1998. The same corrective actions are scheduled for Unit 2 during the ninth refueling outage in the spring of 1999.

The inspectors concluded that lack of an adequate feedwater loop seal existed since initial plant startup.

Once the lack of an adequate loop seal and other related issues were documented in condition reports, PPSL performed thorough operability determinations that met the requirements of Generic Letter (GL) 91-18, Evaluation of Degraded and Non-Conforming Conditions.

PPSL's completed and planned corrective actions were well planned and thorough.

No violation was identified, this URI is closed.

Closed URI 50-387 388 97-09-03 Leakage Rate Test for FW 7A/B Containment Isolation Valves The FW'containment isolation valve arrangement credited in the SSES design basis consists of three isolation valves for each FW header.

Short term containment isolation is provided by the FW check valves inside containment (10 A/B) and the FW 7A/B check valves outside containment.- Manually operated stop-check valves (32A/B) are located farther upstream to provide long term isolation capability.

In July 1996, PPRL issued CR 96-1407 which identified inconsistencies between the FSAR and TS for the FW containment boundary penetrations.

Specifically, the FSAR considered the 7 A/B valves containment isolation valves, but the TS did not consider the 7 A/B valves as containment isolation valves.

The absence of the 7 A/B valves in TS resulted in the valves not being leak tested or PPSL not requesting a specific test exemption as required by 10 CFR 50 Appendix J.

PP&L has corrected the inconsistencies in the FSAR and TS on Unit 1 by modifying, testing and correcting TS to accurately reflect the FW 7 A/B valves as containment isolation valves.

This was completed during the Unit 1 tenth refueling outage in the spring of 1998. The same corrective actions are scheduled during the Unit 2 ninth refueling outage in the spring of 1999. Additionally, the inspectors reviewed PPRL's operability determination and concluded that although these conditions represent a degradation in the defense in depth, they appear to be consistent with the guidance in GL 91-1 In conclusion, PP&L identified inconsistencies within the FSAR and TS for the feedwater containment boundary penetrations.

These inconsistencies resulted in PP&L not requesting an exemption from 10 CFR 50 Appendix J testing, PP&L's completed corrective actions and scheduled corrective actions were thorough and complete.

This non-repetitive, PP&L identified and corrected violation is being treated as a Non-Cited Violation, consistent with Section VII.B.1 of the NRC Enforcement Policy. (NCV 50-387,388/98-11-05)

Closed URI 50-387 388 97-09-04 RWCU Isolation Valves and Requirements of GDC 55 The isolation valves (FW 7 A/B and 39 A/B) for RWCU branch lines are part of the FW penetration isolation arrangement but, do not meet the containment isolation requirements of General Design Criteria (GDC) 55. The RWCU branch line isolation arrangement was not discussed in the FSAR. Although the RWCU isolation valves 82A/B can provide long term positive closure of the line this deviation from GDC 55 does not appear to have been previously reviewed by the NRC.

The inspectors reviewed PP&L correspondence (PLA-4831), FSAR, TS, leak rate testing data and corrective action plans.

This inspection identified that PP&L had modified the Unit 1 FW 7 A/B valve and replaced the FW 39 A/B valve to improve their leakage performance; incorporated these valves into the ITS as part of the modifications.

This was completed during the Unit 1 tenth refueling outage in the spring of 1998. The same corrective actions are scheduled for Unit 2 during the Unit 2 ninth refueling outage in the spring of 1999. Additionally, the inspectors reviewed PP&L's operability determination and concluded that although these conditions represent a degradation'in the design basis defense in depth, they appear to be consistent with the guidance in GL 91-18. PP&L's completed corrective actions and scheduled corrective actions were adequate.

No violation was identified, this URI is closed.

Closed URI 50-387 388 97-09-05 Consequential Failure of FW 10A/B Check Valves PP&L's initial evaluation of a FW line break inside containment assumes the inboard check valve (10 A/B) associated with the broken line is disabled.

This issue is of concern since failure of the FW 10 A/B valve in conjunction with a the worst case single failure could lead to failure of the containment to isolate on a FW line break in containment.

The inspectors performed in-field reviews of PP&L correspondence (PLA-4831) and revised operability determination (CR 96-0046, revision 14). This review identified recent analysis of the FW piping that demonstrated the stresses imposed on the 10 A/B valve were within the valve design stress allowables and accelerations due to pipe whip loads on the 10 A/B valve internals were acceptable by the manufacture.

Therefore, the inspectors concluded that the 10 A/B valves should be capable of performing their isolation function.

No violation was identified, this URI is close Closed URI 50-387 388 98-01-09 Primary Containment Penetration Leak Rate Test - NOED (EA98-188)

PPRL identified two Unit 1 primary containment penetrations (15 pressure instruments) and five Unit 2 primary containment penetrations (18 pressure instruments) which had not been leak rate tested in accordance with TS requirements.

PPSL requested and received a notice of enforcement discretion (NOED) for the associated surveillance requirements, and issued LER 50-387/98-002 for this issue.

NRC inspection report 50-378,388/98-10reviewed and closed the LER, based on an in-field review of the PPRL corrective actions, and issued Non-Cited Violation (NCV) 50-387/98-10-03for the identified TS violation.

No further followup or corrective action review is necessary for this issue.

Therefore, this unresolved item is closed.

Closed VIO 50-387 388 97-06-13 Failure to Perform 50.59 Evaluation for Mod to RHRSW and ESW Pump Motors The violation was issued because protective screens had been added.to the RHRSW and ESW pumps without the documented safety evaluation that is required by 10 CFR 50.59.

The inspectors performed an in-field walkdown of the RHRSW and ESW pumps and reviewed PPtkL's evaluation on the protective screens.

In CR 97-3595 PPSL dispositioned the installation of the screens as "Use-as-is" with a 10 CFR 50.59 determination.

The inspectors concluded the CR was acceptable and NDAP-QA-1202, "Nuclear Department Modification Program", contained adequate controls to prevent similar field changes from circumventing the design change process.

This violation is closed.

IV. Plant Su ort R1 Radiological Protection and Chemistry Controls R1.1 Radiolo ical Controls - External and Internal Ex osure a.

Ins ection Sco e 83750-03 The inspector evaluated the effectiveness of selected aspects of the applied radiological controls program.

The evaluation included a selective review of the adequacy and implementation of the following radiological controls program elements:

access controls to radiologically controlled areas use and adequacy of personnel occupational exposure monitoring devices maintenance of personnel occupational radiation exposures (external and internal) within applicable regulatory limits and As-Low-As-Reasonably-Achievable (ALARA)

'

implementation of the radiation work permit program including the effectiveness of work planning The inspector evaluated the licensee's performance in the above selected areas via observation of activities, tours of the radiologically controlled area (RCA),

discussions with cognizant personnel, review of historical documentation, and review an'd evaluation of applicable station procedures.

b.

Observations and Findin s PPRL implemented effective access controls to the radiological controlled areas of the station.

The access controls included use of Radiation Work Permits (RWPs)

and use of bar code readers to control personnel access in conjunction with computerized log-in and access turnstiles activated by electronic dosimeters.

No access control deficiencies were identified.

PPKL supplied and required use of appropriate personnel monitoring devices for access to the RCA. Thermoluminescent dosimeters and personnel alarming dosimeters were observed to be properly worn to measure external dose.

Access controls for high radiation areas were effective and radiological postings and labels throughout the areas toured provided additional administrative controls and information to the wo'rker. Radiological surveys were also available in-plant.

PPS.L maintained personnel occupational radiation exposures (external and internal)

within applicable regulatory limits and ALARA. A review of historical personnel exposure data for 1998 (as of mid-October), identified that individual exposure results'for total effective dose equivalent, lens of the eye dose equivalent, and shallow-dose equivalent were well below regulatory requirements.

Further, the maximum individual committed effective dose equivalent for any one individual was well within applicable limits.

A selective review of the planning and implementation of RWP No.98-089, used for repair of a steam leak in the main steam pipeway at 40% reactor power, found the permit to be detailed and effective.

Cumulative occupational exposure for this task was within reasonable established exposure goals.

c.

Conclusions PP&L implemented effective radiological controls at SSES.

Access controls to radiologically controlled areas were effective, appropriate occupational exposure monitoring devices were provided and used, personnel occupational exposure was

,maintained within applicable regulatory limits and As-Low-As-Reasonably-Achievable (ALARA),and the radiation work permit program was implemented properly for control of radiological wor R1.2 Radioactive Materials Contamination Surve s

and Monitorin a.

Ins ection Sco e 83750-03 The inspector evaluated the effectiveness of PP&L's surveys, monitoring and control of radioactive materials and contamination.

The evaluation included a selective review of the adequacy and effectiveness of the following radioactive material and contamination control program elements:

surveys and monitoring of radioactive material and contamination calibration status of survey and monitoring equipment proper use of personal contamination monitors and friskers adequacy of surveys during work that involved changing exposure conditions tracking of personnel contamination events and goals, percentage of the station that was identified as contaminated.

The inspector evaluated the licensee's performance in the above selected areas via observation of activities, tours of the RCA, discussions with cognizant personnel, review of historical documentation, and review and evaluation of applicable station procedures.

b.

Observations and Findin s PPSL implemented an effective radioactive material and contamination control program.

Personnel contamination monitors (friskers) and radiation survey meters exhibited current calibration stickers and were appropriately used by personnel exiting areas (e.g., whole body contamination monitors at the RCA exit).

A health physics technician was observed to conduct a routine periodic area survey for radiation, contamination, and airborne radioactivity, in a capable and effective manner.

Further, survey records contained appropriate radiological information for the radiation worker.

Goals to assist in monitoring and tracking the effectiveness of person'nel and area contamination continued to be maintained and used to gauge the overall effectiveness of the stations programs.

c.

Conclusions PPSL implemented overall effective surveys, monitoring, and control of radioactive materials and contamination, The surveys, monitoring, and controls were performed with calibrated and properly used devices.

Personnel and area contamination rates were properly tracked and trende ALARAPro ram Ins ection Sco e 83750-03 The inspector evaluated the effectiveness of PPSL's program to maintain occupation radiation exposure as low as is reasonably achievable. The evaluation include a selective review of the adequacy and effectiveness of the following ALARAprogram elements:

conduct of Station ALARACommittee (SAC) including SAC meeting minutes pre-job ALARAreview and post-job ALARAmeeting for RWP 98-089 ALARAreport for the last outage preparations and dose projections for hydrogen water chemistry.

The inspector evaluated PPSL's performance in the above selected areas via observation of activities, tours of the RCA, discussions with cognizant personnel, review of historical documentation, and review and evaluation of applicable station procedures.

Observations and Findin s PPSL implemented an overall effective program to maintain occupation exposure ALARA". The SAC provided active oversight of the stations exposure control programs including evaluation of cumulative dose versus projected doses, Employee ALARAConcerns, and the proposed hydrogen injection rate reduction strategy.

The SAC meeting minutes documented ALARAissues and goal status.

The pre-job ALARAreviews were of good quality. The pre-job review for RWP 98-089 (repair of a steam leak in the main steam pipeway at 40% reactor power) was detailed and thorough and the post-job ALARAmeeting for this RWP was well-attended and resulted in a summary of lessons learned with good participation on the part of those present.

The ALARAreport for the last outage (Unit 1 10RIO) provided an overall summary of radiological protection during the outage.

Lessons learned and potential improvements were described.

The station met its outage ALARAgoal by completing all outage work for 205 person-rem compared to a goal of 242.

Preparations and dose projections for hydrogen water chemistry were extensive and on-going for a number of years, including visits to other plants already operating with hydrogen water chemistry in order to gather data on operational experience.

Conclusions PPRL implemented an overall effective program to maintain occupation radiation exposure as low as is reasonably achievable. The Station ALARACommittee provided good oversight of program areas.

The pre-job ALARAreview and the post-job ALARAmeeting for RWP 98-089 (steam tunnel work) were detailed and

effective.

The ALARAreport for the last outage was of good quality, and extensive preparations were on-going for hydrogen water chemistry.

R1.4 Radiation Protection Pro ram Chan es a.

Ins ection Sco e 83750-03 The inspector evaluated selected changes in the radiation protection program organization, staffing, programs and procedures.

The inspector evaluated PPSL's performance in the above areas via observation of activities, tours of the RCA, discussions with cognizant personnel, review of historical documentation, and review and evaluation of applicable station procedures.

b.

Observations and Findin s PPSL recently selected a new Radiological Operations Supervisor. The supervisor met applicable TS qualification requirements.

PPSL was upgrading its radiological control access facilities including the instrument calibration facilities and the health physics counting facilities. Planned improvements included additional whole-body friskers at the exit point. Shielding in the roof was being provided to maintain acceptable background levels for the calibration and counting facilities in anticipation of the initiation of hydrogen water chemistry.

The use of hydrogen water chemistry was expected to increase ambient background radiation levels.

c.

Conclusions PPSL properly implemented its TS qualification requirements when selecting a new Radiological Controls Operations Supervisor.

PPSL was making proactive improvements to its radiological controls facilities to maintain survey and monitoring program capabilities in anticipation of initiation of hydrogen water chemistry.

R7 Quality Assurance in Radiological Protection Activities a0 Ins ection Sco e 83750-03 The inspector evaluated the effectiveness of PPKL's self-identification and'orrective action processes.

The evaluation include a selective review of the adequacy and effectiveness of the following program elements and documents:

QA surveillance reports corporate assessments self-assessments Condition Reports

20" The inspector evaluated PP5L's performance in the above area via observation of activities, tours of the RCA, discussions with cognizant personnel, review of applicable documentation, and review and evaluation of applicable station procedures.

b.

Observations and Findin s PPSL implemented an overall effective self-identification and corrective action program in the area of radiological controls. Quality assurance surveillances were focused in scope and highly detailed and resulted in a number of findings.

Corporate assessments included use of independent outside consultation, For example, a detailed evaluation was performed of the 1997 ALARAperformance which concluded that the program was good and well implement and identified areas for improvement.

The SAC developed an action plan which addressed all the recommendations for improvement.

Quarterly self-assessments were performed by the Health Physics Group that were performance-based and focused on in-field radiological controls.

A low threshold was established and implemented for initiation of CRs.

The CRs were evaluated for trends.

PPSL issued CR 98-2968, on September 17, 1998, which identified and documented an adverse trend in the number of events involving posting discrepancies of high radiation areas in 1998 at SSES.

Station management assigned a high significance level (level 2) to this CR and required a timely cause determination (by October 26, 1998). The CR documentation included an analysis for cause and identified corrective actions. PPSL's actions on this matter demonstrated an appropriate level of station management's sensitivity to radiological controls issues.

C.

Conclusions PPSL's self-identification and corrective action processes in the area of radiation protection were effective.

Quality Assurance surveillances, corporate assessments, and self-assessments continued to be effective in identifying, at a low threshold, deficiencies and improvement opportunities.

Corrective actions were implemented for findings.

S2 Status of Security Facilities and Equipment (71750,92904)"

Security computer replacement, in the Security Control Center, is currently in progress and was observed to be well controlled.

The inspector's'bserved operations of the Security Control Center and the Alternate Security Control Center and verified that the alarm stations were well staffed and included an extra operator, as needed, to support system testing.

The inspectors also verified,

.through observations and interviews, that the computer replacement modification did not interfere with operational activities, or the execution of the detection, assessment and response function V. IVlana ement Meetin s X1 Exit Meeting Summary The inspectors presented the inspection results to members of licensee management at the conclusion of the inspection report period on November 30, 1998. The licensee acknowledged the findings presented.

The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary.

No proprietary information was identifie ATTACHMENT1 INSPECTION PROCEDURES USED IP 37550 IP 37551 IP 40500 IP 61726 IP 62707 IP 71707 IP 83750 IP 92700 IP 92901 IP 92902 IP 92903 IP 92904 Engineering Onsite Engineering Observations Effectiveness of Licensee Controls in Identifying, Resolving, and Preventing Problems Surveillance Observations Maintenance Observations Plant Operations Occupational Radiation Exposure On Site Followup of Reports Followup Plant Operations Followup Maintenance Followup Engineering Followup Plant Support

~Oened None.

ITEMS OPENED, CLOSED, AND DISCUSSED 50-387,388/98-1 1-01 50-387/98-1 1-02 50-387/98-1 1-03 50-387/98-1 1-04 NCV Feedwater/Main Turbine Trip Surveillance Inadequate (section 08.1)

NCV Suppression Pool Temperature Surveillance Testing Did Not Meet TS (section 08.1)

NCV Continuous Fire Watch Not Established within TS Time Limit (section 08.1)

NCV Continuous Fire Watch Not per Requirements of TS (section 08.1)

50-387, 388/98-11-05 NCV Leak Rate Test for FW 7A/B Containment Isolation Valves (section E8.1)

~Udeted None.

Closed 50-387/98-010-00 50-387/98-01 2-00 50-387/98-01 3-00 LER Feedwater/Main Turbine Trip Surveillance Inadequate (section 08.1)

LER Suppression Pool Temperature Surveillance Testing Did Not Meet TS (section 08.1)

LER Continuous Fire Watch Not Established within TS Time Limit (section 08.1)

50-387/98-01 5-00 50-388/97-10-01 50-387, 388/98-01-01 50-387,388/96-06-01 LER Continuous Fire Watch Not per Requirements of TS (section 08.1)

URI TS 3.0.3 Entry to Support Surveillance Activities (section 08.2)

IFI Operator Re-Qualification Training Program (section 08.2)

URI Feedwater Loop Seal (section E8.1)

Attachment

50-387,388/97-09-03 50-387,388/97-09-04 50-387,388/97-09-05 URI Leak Rate Test for FW 7A/B Containment Isolation Valves (section E8.1)

URI RWCU Isolation Valves and Requirements of GDC 55 (section E8.1)

URI Consequential Failure of FW 10 A/B Check Valve (section E8.1)

50-387,388/98-01-09 URI Primary Containment Penetration Leak Rate Test-NOED (section E8.1)

50-387,388/97-06-13 URI Failure to Perform 50.59 Evaluation for Mod to RHRSW and ESW Pump Motors (section E8.1)

Attachment

ADS ALARA BPV CFR CR CRD CREOASS ECCS EDG E.l.

ESF ESW FSAR FW GDC GL gpm HCU HPCI IR ITS LCO, LER MSIV NCV NDAP.

NOED NOV NPO NRC NRHX OD PCIV PCO PDR PM PORC PPtkL QA QC RCIC RBCCW RCA RHR RHRSW RWCU LIST OF ACRONYMS USED Automatic Depressurization System As-Low-As-Reasonably-Achievable Turbine Bypass Valve Code of Federal Regulations Condition Report Control Rod Drive Control Room Emergency Outside Air Supply System Emergency Core Cooling System Emergency Diesel Generator Escalated Enforcement Item Engineered Safety Feature Emergency Service Water Final Safety Analysis Report Feedwater General Design Criteria Generic Letter Gallons Per Minute Hydraulic Control Unit High Pressure Coolant Injection System

[NRC] Inspection Report Improved Technical Specification Limiting Condition for Operation Licensee Event Report Main Steam Isolation Valve Non-Cited Violation Nuclear Department Administrative Procedure Notice of Enforcement Discretion Notice of Violation Nuclear Plant Operator Nuclear Regulatory Commission RWCU Non-Regenerative Heat Exchanger Operability Determination Primary Containment Isolation Valve Plant Control Operator Public Document Room Preventative Maintenance Plant Operations Review Committee Pennsylvania Power and Light Company Quality Assurance Quality Control Reactor Core Isolation Cooling System Reactor Building Closed Cooling Water System Radiologically Controlled Area Residual Heat Removal System Residual Heat Removal Service Water System Reactor Water Clean Up

Attachment

awe SAC SCBL SCFH SER SPOTMOS SRM SRV SS SSES STA TS TSI URI US VDC VIO WA Radiation Work Permit Station ALARACommittee Secondary Containment Bypass Leakage Standard Cubic Feet per Hour Safety Evaluation Report Suppression Pool Temperature Monitoring System Source Range Monitor Safety Relief Valve Shift Supervisor

.

Susquehanna Steam Electric Station Shift Technical Advisor Technical Specification Technical Specification Interpretation

[NRC] Unresolved Item Unit Supervisor Voltage Direct Current Violation Work Authorization