IR 05000387/2003003
| ML031920524 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 07/11/2003 |
| From: | Shanbaky M Reactor Projects Region 1 Branch 4 |
| To: | Shriver B Susquehanna |
| References | |
| IR-03-003 | |
| Download: ML031920524 (35) | |
Text
July 11, 2003
SUBJECT:
SUSQUEHANNA STEAM ELECTRIC STATION - NRC INTEGRATED INSPECTION REPORT 05000387/2003003 AND 05000388/2003003
Dear Mr. Shriver:
On June 28, 2003, the US Nuclear Regulatory Commission (NRC) completed an inspection at your Susquehanna Steam Electric Station Units 1 and 2. The enclosed integrated inspection report presents the results of that inspection, which was discussed with Richard L. Anderson, Vice President - Nuclear Operations, and other members of your staff on July 3, 2003.
This inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.
This report documents one self-revealing finding of very low safety significance (Green). This finding was determined to involve a violation of NRC requirements. However, because of the very low safety significance and because the issue was entered into your corrective action program, the NRC is treating this finding as a non-cited violation (NCV), consistent with Section VI.A of the NRC Enforcement Policy. If you contest the NCV in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C.
20555-0001; with copies to the Regional Administrator Region I; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, D.C. 20555-0001; and the NRC Resident Inspector at the Susquehanna Steam Electric Station.
Since the terrorist attacks on September 11, 2001, the NRC has issued five Orders (dated February 25, 2002, January 7, 2003 and April 29, 2003) and several threat advisories to licensees of commercial power reactors to strengthen licensee capabilities, improve security force readiness, and enhance access authorization. The NRC also issued Temporary Instruction 2515/148 on August 28, 2002, that provided guidance to inspectors to audit and inspect licensee implementation of the interim compensatory measures (ICMs) required by the February 25th Order. Phase 1 of TI 2515/148 was completed at all commercial nuclear power plants during calendar year (CY) 02, and the remaining inspections are scheduled for completion in CY 03. Additionally, table-top security drills were conducted at several licensees to evaluate the impact of expanded adversary characteristics and the ICMs on licensee protection and mitigative strategies. Information gained and discrepancies identified during the
Bryce audits and drills were reviewed and dispositioned by the Office of Nuclear Security and Incident Response. For CY 03, the NRC will continue to monitor overall safeguards and security controls, conduct inspections, and resume force-on-force exercises at selected power plants.
Should threat conditions change, the NRC may issue additional Orders, advisories, and temporary instructions to ensure adequate safety is being maintained at all commercial power reactors.
In accordance with 10CFR2.790 of the NRCs "Rules of Practice," a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publically Available Records (PARS) component of the NRCs document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
If you have any questions please contact me at 610-337-5209.
Sincerely,
/RA/
Mohamed Shanbaky, Chief Projects Branch 4 Division of Reactor Projects Docket Nos.
50-387; 50-388 License Nos. NPF-14, NPF-22
Enclosure:
Inspection Report 05000387/2003003, 05000388/2003003 w/Attachment: Supplemental Information
Bryce
REGION I==
Docket Nos.: 05000387, 05000388 License Nos.: NPF-14, NPF-22 Report No.:
05000387/2003003, 05000388/2003003 Licensee:
PPL Susquehanna, LLC Facility:
Susquehanna Steam Electric Station Location:
769 Salem Boulevard Berwick, PA 18603 Dates:
March 30, 2003 to June 28, 2003 Inspectors:
S. Hansell, Senior Resident Inspector J. Richmond, Resident Inspector P. Kaufman, Senior Reactor Inspector R. Kuntz, Reactor Engineer P. Frechette, Physical Security Inspector D. Silk, Senior Emergency Preparedness Specialist Approved by: Mohamed M. Shanbaky, Chief Projects Branch 4 Division of Reactor Projects
Enclosure ii SUMMARY OF FINDINGS IR 05000387/2003003, 05000388/2003003; 03/30/2003 - 06/28/2003; Susquehanna Steam Electric Station, Units 1 and 2. Post Maintenance Testing.
The report covered a 3 month period of inspection by resident inspectors, and announced inspections by reactor engineers, a physical security specialist, and an emergency preparedness specialist. One Green non-cited violation (NCV) was identified. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter 0609 "Significance Determination Process" (SDP). Findings for which the SDP does not apply may be Green or assigned a severity level after NRC management review. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 3, dated July 2000.
A.
NRC-Identified and Self-Revealing Findings Cornerstone: Barrier Integrity
Green. A self-revealing non-cited violation of very low safety significance of Technical Specification 5.4.1 was identified, because PPL did not adequately implement their written procedures for post maintenance testing of a standby gas treatment system (SGTS) damper. On November 19, 2002, maintenance was performed on the damper and the damper was returned to an operable status without performing an adequate post maintenance or operational test. The inadequate test did not verify that the damper could perform its safety function after completion of maintenance activities.
Four months later, PPL discovered that the damper could not perform its safety function.
PPL corrected the condition and restored the damper to an operable condition.
This finding is more than minor because it is similar to examples 1.a and 5.b in NRC Inspection Manual 0612 Appendix E, "Examples of Minor Issues." This violation is of very low safety significance because the finding only represented a degradation of the radiological barrier function provided by the SGTS. During the 4 month period, there were no events that required a SGTS actuation.
A contributing cause of this finding was related to the Human Performance cross-cutting area, in that maintenance technicians and operators did not follow procedures to perform an adequate post maintenance test. As a result, the component was returned to service while in a degraded condition and was unable to perform its safety function.
(Section 1R19.2)
B.
Licensee Identified Violations None.
Enclosure iii TABLE OF CONTENTS SUMMARY OF FINDINGS.................................................... ii Summary of Plant Status..................................................... 1 1.
REACTOR SAFETY................................................... 1 1R04 Equipment Alignments............................................ 1
.1 Partial System Walkdowns................................... 1 1R05 Fire Protection.................................................. 2
.1 Routine Plant Area Inspections
............................... 2 1R07 Heat Sink Performance........................................... 3 1R11 Licensed Operator Requalification................................... 4 1R12 Maintenance Implementation....................................... 4 1R13 Maintenance Risk Assessment and Emergent Work..................... 5 1R14 Non-Routine Plant Evolutions
...................................... 6
.1 Unit 2 Feedwater Heater Train Isolation......................... 6
.2 Units 1 and 2 Technical Specification Required Shutdown due to Inoperable Control Room Emergency Ventilation System........... 7 1R15 Operability Evaluations
........................................... 7 1R16 Operator Work-Around Cumulative Review............................ 8 1R17 Permanent Plant Modifications...................................... 9
.1 Unit 1 4kV Breaker Replacement.............................. 9 1R19 Post Maintenance Testing........................................ 10
.1 Routine Post Maintenance Testing Observations................. 10
.2 Standby Gas Treatment System Damper Failure................. 11 1R20 Unit 2 Refueling Outage Activities.................................. 14
.1 Control of Outage Activities................................. 14
.2 Reactor Plant Startup Activities.............................. 15 1R22 Surveillance Testing............................................. 15 1R23 Temporary Plant Modification
..................................... 16 1EP6 Drill Evaluation................................................. 17 3.
SAFEGUARDS...................................................... 17 3PP2 Access Control................................................. 17 3PP3 Response to Contingency Events.................................. 18 4.
OTHER ACTIVITIES.................................................. 18 4OA1 Performance Indicator Verification.................................. 18 4OA2 Problem Identification and Resolution............................... 20
.1 Routine PI&R Review...................................... 20 4OA3 Event Follow-up................................................ 20
.1 (Closed) LER 05000388/2000005-01.......................... 20
.2 (Closed) LER 05000388/2003002-00.......................... 20 4OA4 Cross Cutting Aspects of Findings.................................. 21 4OA6 Meetings..................................................... 21
.1 Exit Meeting Summary..................................... 21 ATTACHMENT: SUPPLEMENTAL INFORMATION
Table of Contents (contd)
Enclosure iv KEY POINT OF CONTACT.................................................. A-1 LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED........................... A-1 LIST OF DOCUMENT REVIEWED............................................ A-2 LIST OF ACRONYMS...................................................... A-3
Enclosure Report Details Summary of Plant Status Susquehanna Steam Electric Station (SSES) Unit 1 began the inspection period at full power.
On May 2, reactor power was reduced to approximately 18% for planned maintenance to the main transformers. Reactor power was returned to 100% on May 5. On June 6, reactor power was reduced to approximately 30% for planned maintenance to repair a main turbine extraction steam line leak. Reactor power was returned to 100% on June 8. The unit operated at or near full power for the remainder of the inspection period, with exceptions for control rod pattern adjustments, main turbine control valve testing, and main condenser waterbox cleaning.
Unit 2 began the inspection period shutdown, in a maintenance and refueling outage. The unit was restarted on April 20, and achieved 100% reactor power on April 24. On June 11, a Technical Specification required shutdown was commenced, due to an inoperable control room emergency outside air supply system (section 1R14.2). The reactor shutdown was stopped at 98% power when the problem was corrected. Reactor power was returned to 100% on June 11, and operated at or near full power for the remainder of the report period, with exceptions for control rod pattern adjustments and main turbine control valve testing.
1.
REACTOR SAFETY Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, and Emergency Preparedness 1R04 Equipment Alignments
.1 Partial System Walkdowns (71111.04Q)
a.
Inspection Scope The inspectors performed partial system walkdowns to verify system and component alignment and to note any discrepancies that would impact system operability. The inspectors verified selected portions of redundant or backup systems or trains were available while certain system components were out of service. The inspectors reviewed selected valve positions, electrical power availability, and the general condition of major system components. The walkdowns included the following systems:
"A, "B, "D, and "E" emergency diesel generator (EDG) governor, fuel racks, and air start subsystems, while "C" EDG inoperable - unable to pick up load
Unit 2 reactor core isolation cooling (RCIC) system, while the high pressure coolant injection (HPCI) system was out of service for planned maintenance
"A" EDG with one of two starting air compressor inoperable b.
Findings No findings of significance were identified.
1R05 Fire Protection
Enclosure
.1 Routine Plant Area Inspections (71111.05Q)
a.
Inspection Scope The inspectors reviewed PPLs fire protection program to determine the required fire protection design features, fire area boundaries, and combustible loading requirements for selected areas. The inspectors walked down those areas to assess PPLs control of transient combustible material and ignition sources, fire detection and suppression capabilities, fire barriers, and any related compensatory measures to assess PPL's fire protection program in those areas. The areas and documents reviewed included:
Plant Areas and Fire Zones
Unit 2 reactor feed pump turbines
Unit 2 reactor building main steam line tunnel
Unit 2 condenser bay during outage
Emergency service water pump house, security control center, and diesel &
motor driven fire pump areas
Unit 1 "B" residual heat removal pump room, during suppression pool cooling alignment
"C" EDG, following EDG overhaul
Unit 1 and Unit 2 lower relay rooms
Unit 1 and Unit 2 battery rooms and DC distribution panel areas Pre-fire Plans Procedures and Documents
FP-213-278, "RFP Turbine A, B, C Rooms, TB Elevation 670"
FP-213-253, "Main Steam Pipeway"
FP-013-139, "Unit 1 Lower Relay Room"
FP-013-142, "Unit 2 Lower Relay Room"
FP-013-204, "Diesel Fire Pump Room"
FP-013-205, "Fire and Service Water Pump Area"
FP-013-360, "Security Control Center"
FP-013-200/201, "ESSW Pump House"
FP-113-105, "Unit 1 "B" RHR Pump Room"
FP-013-195, "C" Diesel Generator Bay"
FP-013-168 and 169, "Unit 1 Equipment and Battery Rooms"
FP-013-170 and 171, "Unit 2 Equipment and Battery Rooms"
NDAP-QA-0440, "Control of Transient Combustible & Hazardous Material"
NDAP-QA-0441, "Fire Protection System Station Control"
NDAP-QA-0449, "Fire Protection System Program"
Enclosure b.
Findings No findings of significance were identified.
1R07 Heat Sink Performance (71111.07B)
a.
Inspection Scope The inspector reviewed PPLs methods (inspection, cleaning, maintenance, and performance monitoring) used to ensure adequate heat removal capability of the Unit 2 residual heat removal (RHR) service water heat exchanger (2E205B), and the emergency diesel generator (EDG) A-E jacket water coolers (OE507A-E), lube oil coolers (OE506A-E), and inter-coolers (OE505A-D). Included in this review was a comparison to the commitments made in response to Generic Letter 89-13, "Service Water System Problems Affecting Safety-Related Equipment." In particular, the inspector reviewed various maintenance procedures to verify the cleaning and inspection activities, and their frequencies, were reasonable for ensuring that the heat exchangers were maintained capable of performing as designed.
The inspector reviewed the flow balance testing results of the emergency service water (ESW) system conducted on September 19, 2001, to verify that the surveillance test results were recorded and performance data adequately trended in-order to monitor for potential macro fouling conditions. The inspector compared the testing results to the acceptance criteria in procedure TP-054-076, "ESW Flow Balance."
The inspector reviewed the design fouling factor assumptions for the selected heat exchangers and the engineering analyses of minimum calculated flowrates. This review was performed to verify that the minimum calculated flowrates, in conjunction with the heat transfer capability of the heat exchangers, supported the minimum heat transfer rates assumed for the heat exchangers during normal and emergency shutdown conditions. The inspector also reviewed the eddy current test records for these heat exchangers, and verified that the number of plugged tubes in the heat exchangers was bounded by assumptions contained in the engineering analyses. The inspector reviewed these criteria to ensure that the minimum design bases assumptions were technically justified.
The chemical treatment program for the emergency service water and RHR service water systems was reviewed to verify that potential bio-fouling mechanisms had been adequately identified, corrective measures implemented when necessary, and results monitored for effectiveness. To assess the capability of these systems to support their normal and emergency functions, the inspector reviewed system health reports, monthly clam team reports, and the SSES Chemistry Manual, and held discussions with members of the chemistry department and the emergency service water system engineer. Additionally, the inspector performed a walkdown of the spray pond to assess the condition of the water.
b.
Findings
Enclosure No findings of significance were identified.
1R11 Licensed Operator Requalification (71111.11Q)
a.
Inspection Scope On June 3, the inspectors observed licensed operator performance in the simulator during an emergency preparedness exercise. The inspectors assessed the operators adherence to Technical Specifications (TSs), emergency plan implementation, and the use of emergency operating procedures. The inspectors evaluation focused on the operating crews satisfactory implementation of the emergency plan and emergency action level (EAL) classifications for the simulated equipment malfunctions. The inspectors reviewed the ability of the simulator to model the actual plant performance.
In addition, the inspectors observed PPLs critique of the operators performance. The observed emergency preparedness exercise included:
Unusual Event declaration per EAL 16.1, "Security Event," due to a credible site-specific threat
Alert declaration per EAL 16.2, "Security Event," due to an imminent credible site-specific threat
Site Area Emergency declaration per EAL 4.3 "General," due to the Loss of Offsite Power in conjunction with a Loss of Coolant Accident b.
Findings No findings of significance were identified.
1R12 Maintenance Implementation (71111.12Q)
a.
Inspection Scope The inspectors evaluated PPLs work practices and follow-up corrective actions for selected system, structure, or component (SSC) issues to assess the effectiveness of PPL's maintenance activities. The inspectors reviewed the performance history of those SSCs and assessed PPLs extent of condition determinations for these issues with potential common cause or generic implications to evaluate the adequacy of PPLs corrective actions. The inspectors reviewed PPL's problem identification and resolution actions for these issues to evaluate whether PPL had appropriately monitored, evaluated, and dispositioned the issues in accordance with PPL procedures and the requirements of 10 CFR 50.65, "Requirements for Monitoring the Effectiveness of Maintenance." In addition, the inspectors reviewed selected SSC classification, performance criteria and goals, and PPL's corrective actions that were taken or planned, to verify whether the actions were reasonable and appropriate. The following issues were reviewed:
Equipment Issues
Enclosure
Unit Common "C" EDG failed to pick up load (CR 474234), during SE-024-C01,
"Integrated DG Test"
Unit 2 main steam system maintenance rule functional failures on automatic depressurization system (ADS) (CR 463478), during SE-283-005, "ADS Logic System Functional 24-month Test"; and safety relief valve (SRV) acoustic monitor system (CR 457812), during SI-283-326, "SRV Position Indication Channel Calibration 24-month Test" Procedures and Documents
Maintenance Rule Basis Documents for EDG and main steam systems
System Health Reports for EDG and main steam systems
NDAP-QA-0413, "SSES Maintenance Rule Program"
Maintenance Rule Expert Panel meeting minutes, PLI-92576, dated June 12
Condition Reports 474234, 457812, and 463478
Work Orders 474276, 474275 and 474300 b.
Findings No significant observations or findings were identified.
1R13 Maintenance Risk Assessment and Emergent Work (71111.13)
a.
Inspection Scope The inspectors reviewed the assessment and management of selected maintenance activities to evaluate the effectiveness of PPLs risk management for planned and emergent work. The inspectors compared the risk assessments and risk management actions to the requirements of 10 CFR 50.65(a)(4) and the recommendations of NUMARC 93-01 Section 11, "Assessment of Risk Resulting from Performance of Maintenance Activities." The inspectors evaluated the selected activities to determine whether risk assessments were performed when required and appropriate risk management actions were identified.
The inspectors reviewed scheduled and emergent work activities with licensed operators and work-coordination personnel to verify whether risk management action threshold levels were correctly identified. In addition, the inspectors compared the assessed risk configuration to the actual plant conditions and any in-progress evolutions or external events to evaluate whether the assessment was accurate, complete, and appropriate for the issue. The inspectors performed control room and field walkdowns to verify whether the compensatory measures identified by the risk assessments were appropriately performed. The selected maintenance activities included:
Unit 2 "C" RFPT control system repairs, WO 471880
Unit Common "A" EDG tripped while at full load, CR 475852
Unit 2 HPCI planned on-line maintenance; Generic Safety Assessment GSA-052-005
Enclosure
Unit 2 main turbine steam leak repair, CR 475398
Unit Common "A" EDG air compressor inoperable and "E" EDG substituted for the "C" EDG due to a planned overhaul
Unit Common "A" control structure (CS) chiller trip while "B" CS chiller was out of service for planned maintenance b.
Findings No findings of significance were identified.
1R14 Non-Routine Plant Evolutions (71111.14)
.1 Unit 2 Feedwater Heater Train Isolation a.
Inspection Scope On April 22, the Unit 2 "C" feedwater heater string was isolated, due to a suspected steam leak. As a result, feedwater temperature decreased to less than the minimum feedwater temperature used in the analysis for the minimum critical power ratio (MCPR).
Operations entered TS 3.2.2 for MCPR, until the feedwater heater string isolation was completed. PPL entered this into their corrective action program as condition report 468816.
The inspectors reviewed operating logs, core thermal power limits, plant procedures, and interviewed plant personnel for this issue to independently determine what occurred and evaluate the initiating cause. The inspectors assessed personnel performance during this event to evaluate whether the operator response was appropriate and in accordance with procedures and training.
Procedures and Documents
TS 3.2.2, "MCPR Core Thermal Limits"
ON-247-001, "Loss of Feedwater Heating Extraction Steam"
ON-247-002, "Loss of Feedwater Heater String"
Condition Reports 468816, 468817, and 468820
Enclosure b.
Findings No findings of significance were identified.
.2 Units 1 and 2 Technical Specification Required Shutdown due to Inoperable Control Room Emergency Ventilation System a.
Inspection Scope On June 11, both divisions of the control structure (CS) chillers were inoperable when the "A" CS chiller tripped while the "B" CS chiller was out of service for planned maintenance. The control structure chillers are a safety related support system for the control room emergency outside air supply system, a safety system required by TS 3.7.3. At 12:17 p.m., both Unit 1 and Unit 2 entered TS 3.0.3, "Limiting Condition for Operations Not Met." PPL initiated a Unit 2 reactor shutdown at 1:15 p.m., as required by TS 3.0.3. The Unit 2 shutdown was stopped at 1:20 p.m., at approximately 98%
reactor power, when the "B" CS chiller was returned to an operable condition. PPL entered this issue into their corrective action program as condition report 479166.
The inspectors reviewed operating logs, plant procedures, and interviewed plant personnel for this issue to independently determine what occurred and evaluate the initiating cause. The inspectors assessed personnel performance during this event to evaluate whether the operator response was appropriate and in accordance with procedures and training.
Procedures and Documents
TS 3.7.3 and Basis, "Control Room Emergency Outside Air Supply System"
TS 3.7.4 and Basis, "Control Room Floor Cooling System"
Technical Requirements Manual 3.7.9 and Basis, "Control Structure HVAC"
ON-030-001, "Loss of Control Structure HVAC"
OP-030-001, "Control Structure Chilled Water System"
Condition Report 479166 b.
Findings No findings of significance were identified.
1R15 Operability Evaluations (71111.15)
a.
Inspection Scope The inspectors reviewed operability determinations that were selected based on risk insights, to assess the adequacy of the evaluations, the use and control of compensatory measures, and compliance with the Technical Specifications. In addition, the inspectors reviewed the selected operability determinations to verify whether the
Enclosure determinations were performed in accordance with NDAP-QA-0703, "Operability Assessments." The inspectors used the Technical Specifications, Technical Requirements Manual, Final Safety Analysis Report (FSAR), and associated Design Basis Documents as references during these reviews. The issues reviewed included:
Unit 2 white substance identified on reactor vessel internals, CR 460592 and EWR 461695
Unit 1 "A" core spray full flow test valve, HV-F015A, failed to close during quarterly surveillance, CR 466056
Unit 1 and Unit 2 RHR operation in shutdown cooling mode, during a loss of offsite power, CR 473770
Unit 1 and Unit 2 Fuel pool skimmer surge tank high make-up rate, CR 470365
Unit 2 2D630 battery in equalize for extended period
Unit Common "A" EDG tripped from full load during monthly surveillance test, CR 475852 b.
Findings No findings of significance were identified.
1R16 Operator Work-Around Cumulative Review (71111.16)
a.
Inspection Scope The inspectors reviewed the most significant control room deficiencies, status control tags, and selected corrective action reports to determine whether the functional capability of a system or a human reliability response during an event would be affected.
The inspectors evaluated the operators ability to implement abnormal and emergency operating procedures during postulated plant transients with the existing equipment deficiencies. The review included an evaluation of the cumulative and synergistic effects of the identified operator work-arounds. The following documents were included in the review:
Procedures and Documents
OI-AD-096, revision 4, "Operator Work-Arounds"
Equipment Performance and Material Condition (EPMC) List, "All Open Items Report"
EPMC List, "Operator Work-arounds Report"
EC-049-1051, "Peak Pressure in RHR Piping without Fill & Vent"
OP-149-005, "RHR Suppression Pool Cooling"
Condition Reports 468877 and 478776 More Significant Operator Work-arounds
Enclosure
480VAC load center breaker age related lubrication problem, could lead to loss of main turbine electro-hydraulic control system, main turbine trip with loss of bypass valves and reactor scram with loss of heat sink
Control room telephone system does not support multiple users during a declared emergency, could result in delays in emergency communications
Unit 2 reactor recirculation pump automatic speed run-back could place reactor core flow in Stability Region-2 of the power to flow map
Reactor protection system transfer between the motor generator and the alternate supply requires standby safety systems to be manually started/stopped and safety features for primary containment isolation valves to be bypassed, to avoid automatic actuations during the power supply switching evolution b.
Findings No findings of significance were identified.
1R17 Permanent Plant Modifications (71111.17A)
.1 Unit 1 4kV Breaker Replacement a.
Inspection Scope The inspectors reviewed the 4kV switchgear breaker modification which replaced the Westinghouse air magnetic circuit breakers (original plant equipment) with a Cutler Hammer vacuum circuit breaker.
The inspectors reviewed the modification work instructions, post modification test procedures, and test acceptance criteria to assess whether the testing would verify that affected breaker interlocks and system functions satisfied regulatory and design requirements. The inspectors observed portions of testing activities to verify whether the activities were properly performed in accordance with approved procedures. The inspectors reviewed the test data to evaluate whether the test acceptance criteria were satisfied and whether any unintended system interactions had been identified.
The inspectors reviewed the affected procedures and design basis documents to verify that the affected documents were appropriately updated. The following documents were included in the review:
Procedures and Documents
DCP 375651, "4kV Switchgear Breaker Replacement"
OP-000-001, revision6, section 4.1 "Breakers - Operability Policy"
MT-GE-048, revision 0, "Cutler Hammer Type DHP-VR 4.16 KV Circuit Breaker Inspection and Maintenance"
MT-GE-005, PCAF 2003-5120, "Cutler Hammer Circuit Breaker Seismic Restraints"
Enclosure
TP-104-021, revision 0, "Initial Installation of Unit 1 ESS Bus Incoming Feeder Vacuum Circuit Breakers"
Work order RLWO 478835
Condition reports 481168 and 482609 b.
Findings No findings of significance were identified.
1R19 Post Maintenance Testing (71111.19)
.1 Routine Post Maintenance Testing Observations a.
Inspection Scope The inspectors observed portions of post maintenance testing (PMT) activities in the field to determine whether the tests were performed in accordance with the approved procedures. The inspectors assessed the tests adequacy by comparing the test methodology to the scope of maintenance work performed. In addition, the inspectors evaluated the test acceptance criteria to verify whether the test demonstrated that the tested components satisfied the applicable design and licensing bases and the Technical Specification requirements. The inspectors reviewed the recorded test data to determine whether the acceptance criteria were satisfied. The post maintenance testing activities reviewed included:
Unit 2 Main Turbine Over-speed testing, OP-293-002, after replacing the General Electric turbines with Siemens turbines
Unit 2 "C" RFPT control signal failure and PDU replacement, WO 471880
Unit Common "A" EDG retest after trip logic relay replacements, WO 475858 and SO-024-A01
Unit Common "A" control structure chiller PMT after service water strainer cleaning, WO 300746 b.
Findings No findings of significance were identified.
Enclosure
.2 Standby Gas Treatment System Damper Failure a.
Inspection Scope The inspectors observed portions of post maintenance testing (PMT) activities in the field, on April 17, 2003, for standby gas treatment system (SGTS) damper PDDM-075-54B, and reviewed the completed PMT records (WO 467651 and SE-170-011). The inspectors also reviewed the dampers previous PMT record (WO 400346) for a test conducted on November 19, 2002.
The inspectors assessed the tests' adequacy by comparing the test methodology to the scope of the maintenance work performed. In addition, the inspectors evaluated the test acceptance criteria to verify whether the criteria adequately demonstrated that the damper could perform its intended safety function. The inspectors also verified whether the tests were performed in accordance with the approved procedures, and reviewed the recorded test data to determine whether the acceptance criteria were satisfied. The following documents were included in the review:
Procedures and Documents
MT-GE-030, "ITT Damper Hydramotor NH91/NH95 Overhaul"
V-475 sheet-2, "SGTS Outdoor/Zone Pressure Differential Control Schematic"
M334-46 sheet 2, "Wiring Diagram, Local Control Panel OC883B"
SE-170-011, RTSV 464948, dated 04-17-03, "Secondary Containment Drawdown and In-leakage Test, Zones I & III"
Work Orders 400346, 425144, and 467651
Condition Reports 96-0336, 467613, 467829, 467830, and 468337 b.
Findings Introduction A self-revealing non-cited violation of very low safety significance (Green) of Technical Specification 5.4.1 was identified, because PPL did not adequately implement their written procedures for the PMT of a SGTS damper (i.e., MT-GE-030 and NDAP-QA-0302). On November 19, 2002, maintenance was performed on the damper, and the damper was returned to an operable status without performing an adequate PMT or operational test to verify that the SGTS damper could perform its safety function after completion of maintenance activities. Four months later, PPL discovered that the damper could not perform its safety function.
Enclosure Description On April 16, 2003, damper PDDM-075-54B did not open when the "B" train of SGTS was actuated. This damper is in the flow path between the SGTS and secondary containment. The dampers safety function is to modulate the air flow from the secondary containment into the SGTS, to control reactor building pressure while SGTS is aligned to the secondary containment.
PPL determined that the control wiring of the damper actuator had been reversed when the actuator was last rebuilt. As a result of the wiring error, a control signal to open the damper caused the actuator to move the damper to the closed position. This actuator had been installed on the damper on November 19, 2002, four months earlier. As a result, the "B" train of SGTS was unavailable for 4 months, and both the "A" and "B" trains of SGTS were unavailable on four separate occasions (e.g., a safety system functional failure), when the "A" train of SGTS was removed from service for scheduled maintenance, during February 10 to 12, February 24 to 25, April 11, and April 12, 2003.
The incorrect wiring condition was corrected and the damper restored to an operable status on April 16, 2003.
PPL concluded that it missed an opportunity to prevent the event when it did not perform an adequate PMT to verify damper functionality following the actuator replacement in 2002. Maintenance procedure MT-GE-030, section 8.9, "Actuator Installation and Adjustment," contained a specific installation sequence, including adjustment and wire re-connection steps. The last procedure step, in the installation sequence, required verification of proper damper operation, after all maintenance activities had been completed. However, PPL Maintenance performed the procedure steps out of sequence. PPL performed an in-place damper functional check, by locally stroking the installed damper with portable test equipment, then re-connected the field control wires.
In addition, NDAP-QA-0302, "System Status and Equipment Control," section 6.3.6, required Operations to identify and perform operational testing needed to verify Technical Specification operability, prior to equipment restoration. However, no additional operational testing (e.g., no damper stroke test) was identified or performed after all field wires were re-connected and all maintenance activities were completed to verify that the SGTS damper could perform its safety function. As a result of not performing the installation activities in the sequence specified by the maintenance procedure and not performing the required operational system testing after all field control wires were re-connected, the actuator control wire reversal problem was not identified and the damper was returned to service in an inoperable condition.
Analysis This finding is a performance deficiency because PPL did not implement written procedures to verify that a safety related damper was functional following maintenance activities, and the damper subsequently failed to perform its safety function 4 months later. This finding is more than minor because it is similar to examples 1.a and 5.b in NRC Inspection Manual 0612 Appendix E, "Examples of Minor Issues." This finding
Enclosure affected the Barrier Integrity cornerstone objective to provide reasonable assurance that physical design barriers (i.e., secondary containment) provide protection against a radiological release.
This finding was considered to have very low safety significance, and screened out as Green, using the NRC Significance Determination Process (SDP) Phase-1 Initial Screening for Reactor Inspection Findings for At-Power Situations because the finding only represented a degradation of the radiological barrier function provided by the SGTS. In addition, during the 4 month period that the condition existed, there were no events that required a SGTS actuation.
A contributing cause of this finding involved human performance errors, which are an aspect of the Human Performance cross-cutting area. The first human performance error was that maintenance technicians performed procedure steps out of sequence.
The second human performance error was that operators did not identify operational testing, as required by procedures, to perform an adequate PMT. As a result, the component was returned to service while in a degraded condition, and was unable to perform its safety function.
Enforcement Technical Specification 5.4.1 requires, in part, that written procedures shall be established and implemented as recommended in NRC Regulatory Guide (RG) 1.33 Appendix A. RG 1.33 Appendix A, section 9.a, "Procedures for Performing Maintenance," requires pre-planned maintenance activities be performed in accordance with written procedures for maintenance that can affect the performance of safety related equipment. Contrary to the above, on November 19, 2002, PPL did not implement written procedures NDAP-QA-0302 and MT-GE-030 to verify that SGTS damper PDDM-075-54B could perform its safety function after completion of maintenance activities. Specifically:
(1) NDAP-QA-0302, "System Status and Equipment Control," section 6.3.6, required, in part, that Operations Supervision (i.e., an SRO) identify and perform all operational testing needed to verify Technical Specification operability, prior to equipment restoration. However, no operational testing was identified or performed to verify that the SGTS damper could perform its safety function after the maintenance activity had been completed.
(2) Maintenance procedure MT-GE-030, "ITT Damper Hydramotor NH91/NH95 Overhaul," section 8.9.7, required PPL to verify that the SGTS damper operated properly, after maintenance activities were completed. However, PPL did not appropriately verify that the SGTS damper stroked properly after damper restoration.
Because this violation is of very low safety significance and PPL entered this finding into their corrective action program (CR 467829), this violation is being treated as a non-cited violation (NCV), consistent with Section VI.A of the NRC Enforcement Policy.
(NCV 05000387,388/2003003-01)
Enclosure 1R20 Unit 2 Refueling Outage Activities (71111.20)
.1 Control of Outage Activities a.
Inspection Scope The inspectors observed selected maintenance, testing, and equipment restoration activities to verify whether component configuration management, test control, and post maintenance checks were performed in accordance with NRC requirements and PPL procedures. The inspectors reviewed unexpected plant conditions, emergent work, and system configuration control during testing and maintenance activities to evaluate whether the activities were performed in accordance with NRC requirements and PPL procedures.
The inspectors reviewed the ASME In-service inspection data and the surveillance test data, from the reactor coolant pressure boundary operational leakage test, to evaluate whether the test acceptance criteria were satisfied. In addition, the inspectors evaluated whether the activities were performed in accordance with NRC requirements and PPL approved procedures.
Specific Activities
White substance identified on reactor vessel internals
Fuel channel bowing evaluation
Hydrostatic test, SE-200-002 Procedures and Documents
PL-NF-02-007, revision 4, "Channel Management Action Plan"
Control rod - fuel channel bowing, General Electric 10 CFR 50.21 notification (ENS 39806)
b.
Findings No findings of significance were identified.
Enclosure
.2 Reactor Plant Startup Activities a.
Inspection Scope The inspectors observed selected portions of the reactor startup from the control room to verify that Technical Specifications, license conditions, and administrative requirements were satisfied. The inspectors verified that reactor criticality occurred with the control rod positions within the allowed band predicted by the core design. The following activities and documents were reviewed or observed:
Plant Startup Activities
Reactor operational mode change review by the Plant Operation Review Committee
Startup preparations for mode change
Primary and secondary containment integrity
Reactor startup, control rod withdrawals and reactor criticality
Reactor coolant system heat up activities
Reactivity manipulations with the reactor recirculation system
HPCI and RCIC surveillance testing at low reactor pressure
HPCI surveillance test at 920 psig reactor pressure
Thermal limits verification prior to exceeding 25% reactor power
Main turbine over-speed testing and generator sync to grid, after Siemens Turbine replacement
Feedwater heater level control Procedures and Documents
GO-200-010, "ECCS and Decay Heat Removal in Modes 4 and 5"
GO-200-002, "Plant Startup, Heat up, and Power Operations" b.
Findings No findings of significance were identified.
1R22 Surveillance Testing (71111.22)
a.
Inspection Scope The inspectors observed portions of selected surveillance test activities in the control room and in the field and reviewed the test data results. The inspectors compared the test result to the established acceptance criteria and the applicable Technical Specification or Technical Requirements Manual operability and surveillance requirements to evaluate whether the systems were capable of performing their intended safety functions. The observed or reviewed surveillance tests included:
Enclosure
Unit 1 SE-170-011, "Secondary Containment Drawdown and Inleakage 24-month Test"
Unit 2 SR-255-004, "Scram Time Measurement of Control Rods," performed at 35% reactor power
Unit 1 SO-152-002, "Quarterly RCIC Flow Verification"
Unit Common SO-024-001, "A" EDG Monthly Surveillance Test," observed from the control room
Unit 1 SO-152-004, "Quarterly HPCI Valve Exercise Test," observed from the control room b.
Findings No findings of significance were identified.
1R23 Temporary Plant Modification (71111.23)
a.
Inspection Scope The inspectors reviewed temporary plant modifications to determine whether the temporary changes adversely affected system or support system availability, or adversely affected a function important to plant safety. The inspectors reviewed the associated system design bases, including the Final Safety Analysis Report (FSAR),
Technical Specifications, and assessed the adequacy of the safety determination screenings and evaluations. The inspectors also assessed configuration control of the temporary changes by reviewing selected drawings and procedures to verify whether appropriate updates had been made. The inspectors compared the actual installations to the temporary modification documents to determine whether the implemented changes were consistent with the approved documents. The inspectors reviewed selected post installation test results to verify whether the actual impact of the temporary changes had been adequately demonstrated by the test. The following temporary modifications and documents were included in the review:
Unit 2 temporary power supplied to the safety parameter display system (SPDS),
OT-290-001, "De-energizing SPDS UPS for Maintenance"
Unit 2 temp instrumentation connected to main turbine electro-hydraulic control system for 2 months, per NDAP-QA-0510 trouble shooting plan (WO 469857),
CRs 469838 and 481168 b.
Findings No findings of significance were identified.
Enclosure 1EP6 Drill Evaluation (71114.06)
a.
Inspection Scope On June 3, 2003, the inspectors observed PPLs nuclear emergency response organization (NERO) during an announced emergency preparedness training exercise to evaluate PPLs NERO performance. The simulated emergency included the activation of the operations support center, technical support center, and emergency operations facility. The control room simulator was used for the exercise.
The inspectors observed the conduct of the exercise in the control room simulator. The inspectors assessed licenced operator and NERO adherence to emergency plan implementation procedures, and their response to simulated degraded plant conditions to identify weaknesses and deficiencies in classification, notification, and protective actions recommendations. The inspectors observed PPL's critique of the simulator control room participants when the exercise ended. In addition, on June 10, the inspectors observed PPLs facility critiques to evaluate PPLs identification of weaknesses and deficiencies. The inspectors compared PPLs identified findings against the inspectors observations to determine whether PPL adequately identified failures. The inspectors review included the following documents and procedures:
Susquehanna Emergency Plan, revision 41
EP-PS-126, "Control Room Communicator" b.
Findings No findings of significance were identified.
3.
SAFEGUARDS Cornerstone: Physical Protection 3PP2 Access Control (71130.02)
a.
Inspection Scope The following activities were conducted during the inspection period to verify that PPL has effective site access controls, and equipment in place designed to detect and prevent the introduction of contraband (firearms, explosives, incendiary devices) into the protected area as measured against 10 CFR 73.55(d), the Physical Security Plan, and SSES Security Procedures.
Site access control activities were observed, including personnel and package processing through the search equipment during peak ingress periods on April 22 and 23, 2003. On April 23, 2003, observation of vehicle search activities was also conducted. On April 22, 2003, testing of all access control equipment, including metal detectors, explosive material detectors, and X-ray examination equipment, was observed.
Enclosure b.
Findings No findings of significance were identified.
3PP3 Response to Contingency Events (71130.03)
a.
Inspection Scope The following activities were conducted to determine the effectiveness of PPLs Response to Contingency Events, as measured against the requirements of 10 CFR 73.55 and the SSES Safeguards Contingency Plan:
On April 23, 2003, a review of documentation associated with PPLs force-on-force exercise program was conducted. The review included documentation and critiques for exercises conducted since the first quarter of 2002, when the exercises were resumed post 9/11/01.
On April 22, 2003, performance testing of the SSES intrusion detection and alarm assessment systems was conducted. This testing was accomplished by one inspector who toured the entire perimeter and selected areas of potential vulnerability in the intrusion detection system. Concurrently, a second inspector observed the alarm assessment capabilities from the Central Alarm Station. During the walkdown of the intrusion detection system, thirty specific locations were selected for testing.
b.
Findings No findings of significance were identified.
4.
OTHER ACTIVITIES 4OA1 Performance Indicator Verification (71151)
a.
Inspection Scope The inspectors reviewed PPLs performance indicator (PI) data to verify whether the PI data was accurate and complete. The inspectors compared the PI data against the guidance contained in NEI 99-02. The following NRC PIs and PPL documents were included in this review:
Enclosure Procedures and Documents
Nuclear Energy Institute (NEI) 99-02, revision 2, "Regulatory Assessment Performance Indicator Guideline"
NDAP-QA-0737, "Regulatory Performance Assessment"
LI-00-018, "Preparation of Performance Indicator Data, NRC Submittals, and Cornerstone Assessment Reports"
Susquehanna Licensee Event Reports for 2002 and 2003
EP-AD-022, revision 2, "Emergency Planning Performance Indicators" Mitigating Systems Cornerstone PIs
Safety System Functional Failure For the period from April 2002 to March 2003, the inspectors examined the PI data, PPL PI data summary reports, and plant records, which included selected Technical Specification limiting condition for operation logs, licensee event reports, and condition reports.
Emergency Preparedness (EP) Cornerstone PIs
Drill and Exercise Performance
Emergency Response Organization Participation
Alert Notification System Reliability For the period from October 2002 to March 2003 (since the last EP PI verification inspection), the inspector assessed the PI data submitted to the NRC. The inspector reviewed PPLs process for identifying the data that is utilized to determine the values for these three PIs. Classification, notification and protective action opportunities were reviewed from licensed operator simulator sessions and site emergency response organization drills and exercises. Attendance records for drill and exercise participation was reviewed for completeness and accuracy. Test results of the alert notification system testing were reviewed.
Physical Protection Cornerstone PIs
Fitness-for-Duty / Personnel Reliability Program
Personnel Screening Program
Protected Area Security Equipment For the period from April 2002 to March 2003, the inspector reviewed PPLs programs for gathering, processing, evaluating, and submitting data for these 3 PIs. The review included PPLs tracking and trending reports, personnel interviews, safeguards events log, and security event reports during the review period.
b.
Findings
Enclosure No findings of significance were identified.
4OA2 Problem Identification and Resolution (71152)
.1 Routine PI&R Review a.
Inspection Scope The inspectors reviewed selected condition reports (CRs), as part of the routine baseline inspection documented in this report. The CRs were assessed to verify whether the full extend of the various issues were adequately identified, appropriate evaluations were performed, and reasonable corrective actions were identified. The inspectors evaluated the CRs against the requirements of NDAP-QA-0702, "Action Request and Condition Report Process," and 10 CFR 50, Appendix B.
b.
Findings No findings of significance were identified.
4OA3 Event Follow-up (71153)
.1 (Closed) LER 05000388/2000005-01 Engineered Safety Feature Actuations due to Reactor Protection System Electrical Protection Assembly Breaker Trip
On December 5, 2000, the Unit 2 "B" reactor protection system (RPS) power was lost due to an electrical protection assembly (EPA) breaker trip. The failure resulted in a RPS "B" half scram and corresponding containment isolations. This event was initially reviewed in NRC Inspection Report 05000387,388/2001002, section 4OA3.1. PPL revised the apparent cause for the EPA failure, and provided additional corrective actions to prevent future occurrence. No new issues or additional findings were identified in this review. No violations of NRC requirements were identified. This LER is closed.
.2 (Closed) LER 05000388/2003002-00 Unusual Event Declared for a Contaminated Individual Transported Offsite
On March 24, 2003, an Unusual event was declared for a potentially contaminated individual being transported offsite to a local hospital. The individual was given medical treatment at the hospital and no contamination was found on the individual. No new issues or additional findings were identified in this review. No violations of NRC requirements were identified. This LER is closed.
Enclosure 4OA4 Cross Cutting Aspects of Findings Cross-References to Human Performance Findings Documented Elsewhere Section 1R19.2 describes a finding where a safety related damper was returned to service without verifying that the damper could perform its safety function. Four months later, PPL discovered that the damper could not perform its safety function. The dampers condition could have been reasonable have been identified, prior to its return to service, if maintenance personnel and operators had adequately implemented written procedures for a post maintenance test.
4OA6 Meetings
.1 Exit Meeting Summary On July 3, 2003, the resident inspectors presented the inspection results to R.
Anderson, Vice President - Nuclear Operations, and other members of your staff, who acknowledged the findings.
The inspectors asked PPL whether any material examined during the inspection should be considered proprietary. No proprietary information was identified.
ATTACHMENT: SUPPLEMENTAL INFORMATION
A-1 Attachment SUPPLEMENTAL INFORMATION KEY POINT OF CONTACT PPL Personnel B. Shriver, Senior Vice President and Chief Nuclear Officer R. Anderson, Vice President, Nuclear Operations T. Kirwin, Manager, Maintenance G. Ruppert, Manager, Operations D. Glassic, Outage Manager R. Ferentz, Manager, SSES Security J. Grisewood, Supervisor, Emergency Planning S. Kuhn, Supervisor, Maintenance M. Peal, Supervisor, Operations R. Lengel, Emergency Planning Jim Wolfer, Chemistry Bill Basta, Chemistry John Lines, ISI Dean Leimbach, ISI Frank Wurst, Station Engineering Jim Van Horn, Maintenance Jeff Jeanguenat, ESW System Engineer Rich Centenaro, Design Engineering LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED Opened None Opened and Closed 05000387,388/2003003-01 NCV Standby Gas Treatment System Damper Failure (Section 1R19.2)
Closed 05000388/2000005-01 LER Engineered Safety Feature Actuations due to Reactor Protection System Electrical Protection Assembly Breaker Trip (Section 4OA3.1)05000387/2003002-00 LER Unusual Event Declared for a Contaminated Individual Transported Offsite (Section 4OA3.2)
Discussed
A-2 Attachment None LIST OF DOCUMENT REVIEWED (Not Referenced in the Report)
Section 1R07: Heat Sink Performance Procedures
NDAP-QA-504, revision 4, "Heat Exchanger Program"
M-1453, revision 4, "Heat Exchanger Tube Plugging"
M-1548, revision 0, "Heat Exchanger Performance Monitoring Program"
MT-GM-025, revision 11, "Heat Exchanger - Cleaning and Inspection"
MT-216-002, revision 7, "RHR Heat Exchanger Cleaning, Inspection and Repair"
NEIM-00-1156, revision 1, "Eddy Current Database Management"
NEPM-QA-1159, revision 2, "Heat Exchanger Inspection" Engineering Specifications
H-1001, Revision 5, "Heat Exchanger/Condenser Tube Cleaning"
H-1002, Revision 2, "Protective Epoxy Coating/Lining System for Condenser Tube Sheets, Water Boxes and Large Diameter Piping"
H-1004, Revision 6, "Heat Exchanger/Condenser Inspection and Condition Assessment"
H-1005, Revision 6, "Eddy Current Examination on Plant Heat Exchangers" Calculations
EC-024-0556, Revision 2, "Evaluate Impact of 97F Emergency Service Water Temperature on Diesel Generator A B C D & E Heat Exchanger"
EC-049-1001, Revision 2, "Residual Heat Removal Heat Exchanger Performance at 7580 & 8000 GPM Residual Heat Removal Service Water Flow Rate"
EC-CHEM-1018, Revision 2, "Justification for the Assurance of Adequate Heat Removal Capabilities Using the SSES Heat Exchanger Preventative Maintenance Program"
EC-HXPM-1001, Revision 0, "Pilot Heat Exchanger Selection Evaluation Study Heat Exchanger Performance Monitoring Program"
EC-HXPM-1003, Revision 0, "Thermal Performance Test Data Evaluation & Uncertainty Analysis for 2E205B RHR Heat Exchanger"
EC-HXPM-1016, Revision 0, "Thermal Performance Test Data Evaluation & Uncertainty Analysis for 1E205A RHR Heat Exchanger"
EC-HXPM-1024, Revision 0, "Thermal Performance Test Data Evaluation & Uncertainty Analysis for Initial E Jacket Water Cooler Performance Test Prior to Cleaning Tubes"
EC-HXPM-1025, Revision 0, "Thermal Performance Test Data Evaluation & Uncertainty Analysis, E Jacket Water Cooler Performance After Cleaning Heat Exchanger Tubes" Work Orders
A-3 Attachment
ERPM 358486, "M1181-52 Clean and Inspect the RHR SW Heat Exchanger 2E205B" Condition Reports
306291, 306299, 341568, 345243, 350322, 352157, 355395, 364381, 404015, 404126, 405250, 405450, 406054, 406062, 423968, 423983, 425956 Miscellaneous
PLA-3349, "Response to Generic Letter 89-13"
PLA-3377, "Supplemental Response to Generic Letter 89-13"
PLI-61650, "Status of Generic Letter 89-13"
PLIS-45086, "Generic Letter 89-13 Flow Balancing Commitment" Sections 3PP2 & 3PP3: Access Control and Response to Contingency Events
Security Plan and Procedure Audit Number 2002-051
Safeguards Event Log
Susquehanna Steam Electric Station Physical Security Plan LIST OF ACRONYMS ADS Automatic Depressurization System CFR Code of Federal Regulations CR Condition Report CS Control Structure EAL Emergency Action Level EDG Emergency Diesel Generator EP Emergency Preparedness EPA Electrical Protection Assembly EPMC Equipment Performance and Material Condition ESW Emergency Service Water FSAR
[SSES] Final Safety Analysis Report HPCI High Pressure Coolant Injection LER Licensee Event Report MCPR Minimum Critical Power Ratio NCV Non-cited Violation NEI Nuclear Energy Institute NERO emergency response organization NRC Nuclear Regulatory Commission PI
[NRC] Performance Indicator PMT Post Maintenance Test PPL PPL Susquehanna, LLC QA Quality Assurance RCIC Reactor Core Isolation Cooling RG
[NRC] Regulatory Guide
A-4 Attachment RHR Residual Heat Removal RPS Reactor Protection System SDP Significant Determination Process SGTS Standby Gas Treatment System SPDS Safety Parameter Display System SSC Structure, System, or Component SSES Susquehanna Steam Electric Station SRV Safety Relief Valve TS Technical Specifications WO Work Order