IR 05000387/1999003
| ML17164B023 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 04/19/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17164B021 | List: |
| References | |
| 50-387-99-03, 50-387-99-3, 50-388-99-03, 50-388-99-3, NUDOCS 9904290226 | |
| Download: ML17164B023 (24) | |
Text
4 U.S. NUCLEAR REGULATORYCOMMISSION
REGION I
Docket Nos:
License Nos:
50-387, 50-388, 72-28 NPF-14, NPF-22 Report No.
50-387/99-03, 50-388/99-03 Licensee:
PP&L, Inc.
2 North Ninth Street Allentown, Pennsylvania 19101 Facility:
Susquehanna Steam Electric. Station Location:
P.O.'Box 35
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Berwick, PA 18603-0035 Dates:
February 2, 1999 through March 15, 1999 Inspectors:
S. Hansell, Senior Resident Inspector J. Richmond, Resident Inspector A. Blarney, Resident Inspector
'.
Gray, Senior Reactor Inspector Approved by:
Curtis J. Cowgill, Chief Reactor Projects Branch 4 Division of Reactor Projects 904290226 99'04i9 PDR ADQCK 05000387
EXECUTIVE SUMMARY Susquehanna Steam Electric Station (SSES), Units 1 &2 NRC Inspection Report 50-387/99-03, 50-388/99-03 This inspection included aspects of PP&L, Inc.'s operations, maintenance, and engineering, at SSES.
The integrated report covers a six week period of routine resident inspection activities and an inspection of your independent spent fuel storage facility.
~Oeratinne
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In a Licensee Event Report, PP&L identified that on two occasions, the Unit 2 core spray quarterly flowsurveillance test did not meet the Technical Specification (TS) acceptance criteria due to a procedure error. PP&L's proposed and completed corrective actions, including procedure and programmatic actions, were good. This Severity Level IV violation is being treated as a Non-Cited Violation, consistent with Appendix C of the NRC Enforcement Policy. This violation is in PP&L's corrective action program as condition report 98-3070.
LER 50-388/98-011 is closed.
(Section 08.1)
Maintenance
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The observation of ultrasonic testing of the residual heat removal injection valves determined that the technique applied was effective in identifying significant stem cracking. The ultrasonic testing minimized personnel radiation exposure and resulted in the timely identification of an additional cracked valve stem. (Section M1.1)
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The residual heat removal service water (RHRSW) radiation monitors do not meet requirements of General Design Criteria 64. Specifically, the radiation monitors would not be functioning following a postulated design basis accident, since the monitors can not be manually started locally in the high post accident area radiation levels.
In addition, the location of a backup grab sample would not provide a representative sample as delineated in Regulatory Guide 1.21. This is a violation of 10 CFR Part 50, Appendix B, Criterion III,"Design Control," which requires, in part, selection of suitable equipment that are essential to the safety related function of the system.
This Severity Level IVviolation is being treated as a Non-Cited Violation, consistent with Appendix C of the NRC Enforcement Policy. This violation is documented in PP8 L's corrective action program as condition report 91031.
(Section E2.1)
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The planning and construction of the Susquehanna Independent Spent Fuel Storage Facility (ISFSF) were being accomplished well. Quality Assurance involvement has been evident throughout the ISFSF project. (Section E2.2)
A TABLEOF CONTENTS I. Operations
Conduct of Operations..
01.1 Unit Operations and Operator Activities
Operational Status of Facilities and Equipment 02.1 Operational Safety System Alignment
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02.2 Failure of the Unit 1 Residual Heat Removal Injection Valve HV-151F017B
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Miscellaneous Operations Issues 08.1 Licensee Event Report Review
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II. Maintenance M1 Conduct of Maintenance
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M1.1 Surveillance and Pre-Planned Maintenance ActivityReview.
III. Engineering E2 Engineering Support of Facilities and Equipment E2.1 Residual Heat Removal Service Water Radiation Monitor..
E2.2 Independent Spent Fuel Storage Facility Activities..
E8 Miscellaneous Engineering Issues E8.1 Followup of Open Items.........
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V. Management Meetings
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X1 Exit Meeting. Summary
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INSPECTION PROCEDURES USED..
ITEMS OPENED, CLOSED, AND DISCUSSED LIST OF ACRONYMS USED
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Re ort Details Summa of Plant Status
'usquehanna Steam Electric Station (SSES) Unit 1 was at 100% power at the beginning of the inspection period.
Power was reduced to 60% on February 5, to perform repairs to the reactor water cleanup (RWCU) non regenerative heat exchanger and returned to 100% power on February 9. On February 18, power was reduced to 75% when the "B" loop of feedwater was removed from service, due to a steam leak on the feedwater flow element flange and returned to 100% power on February 21. On March 5, power was reduced to 60%, to perform main condenser in-leakage testing.
Following leakage testing and repair on March 8, power was increased to 100% for the remainder of the inspection period.
SSES Unit 2 operated at 100% power through the inspection period until March 13. Unit 2 shutdown on March 13, to start the Unit 2 ninth refueling outage.
I. 0 erations
Conduct of Operations
'1.1 Unit 0 erations and 0 erator Activities (71707)
The inspectors determined routine operator activities were adequately prescribed, communicated, and conservatively performed in accordance with SSES procedures.
Effective controls were implemented for safe plant operation.
Control room activities were well performed.
Control room logs accurately reflected plant activities. Control room operators displayed good questioning perspectives during work planning meetings and prior to releasing scheduled work activities.
Operational Status of Facilities and Equipment 02.1 0 erational Safe S stem Ali nment (71707)
During routine plant tours, the proper alignment and operability of various safety systems, engineered safety features, and on-site power sources were verified. In addition, a partial walkdown of the Unit 1 "B" loop residual heat removal (RHR) system and the Unit 1 RWCU system was conducted.
The inspectors observed operators'ctions to return both systems to service, which included a partial system filland vent, following system maintenance.
Overall equipment operability, material condition, and housekeeping conditions were good;
'Topical headings such as 01, MB, etc., are used in accordance viith the NRC standardized reactor inspection report outline.
Individual reports are not expected to address all outline topic I'
Failure of the Unit 1 Residual Heat Removal In'ection Valve HV-151F017B (71707, 93702)
On February 11, 1999, the Unit 1 "B" loop of RHR was being restored to service following planned maintenance.
During the system filland vent the plant control operator observed the system pressure decrease unexpectedly then return to normal pressure over a five minute period.
PP&L initiallyconcluded that there was a potential flow restriction in the condensate transfer system keep fillsupply to Unit 1 "B" loop of RHR. An operability determination (CR 90208) was completed which documented the flow restriction and the "B" loop of RHR was declared operable.
A work plan was developed to determine the location of the flow restriction in the keep fillline. During execution of the work plan, on February 27, 1999, PP&L discovered that the valve disk on the RHR injection valve, HV-151F017B had separated from the valve stem.
The stem was found severed at the top thread where the stem threads into the valve disk.
The Unit 1 "B" loop of RHR was declared inoperable and the valve stem and disk were replaced.
PP&L performed ultrasonic inspections on three similar RHR injection valves and identified that the Unit 2 "A" RHR injection valve had a crack that extended through 75%
of the stem.
The Unit 2 "A" RHR injection valve stem and disk were replaced.
Currently PP&L has replaced the valve stem and disc on the affected RHR injection valves and determined by ultrasonic examination that the other two valves do not exhibit similar problems.
A detailed followup inspection is scheduled in Aprilto monitor and review the ongoing PP&L root cause analysis.
Miscellaneous Operations Issues Licensee Event Re ort Review (92700)
Closed LER 50-387/98-016-00 Reactor Scram Due to Main Generator Trip On October 3, 1998, Unit 1 experienced a generator backup lockout relay trip which resulted in a main generator trip and a reactor scram. Condition report 71508, documented that the scram was due to pitted contacts on the main generator potential transformer (PT) circuitry, which resulted in a false signal to the generator ground circuitry. The inspectors monitored PP&L's field verifications and root cause analysis.
This included a review of test procedure TP-198-001, "Testing of Main Generator Sync Breaker 1R101 Circuitry", the Plant Operations Review Committee (PORC) review of this procedure, and PP&L's corrective actions.
The PORC appropriately focused on plant safety issues, including off site power and personnel safety. The verification of the trip relays circuitry was methodical and well performed The PP&L root cause analysis identified that maintenance personnel did not silver plate these contacts during the previous refueling outage.
This resulted in the main generator PT contact pitting and the generator trip. The inspectors'n field review verified that PM E1784-51, "Inspect and Clean Isophase Bus, PT Cabinets and Insulators" was revised
to clean and silver plate all primary and secondary PT disconnect contacts each outage.
No violations of NRC requirements were identified. This LER is closed.
Closed LER 50-388/98-011-00 Core Spray Quarterly Flow Surveillance Did Not Meet Acceptance Criteria PP&L identified that on at least two occasions, the Unit 2 core spray quarterly flow surveillance test did not meet the Technical Specification (TS) acceptance criteria due to a procedure error. A seven pounds per square inch gauge (psig) correction factor was added inappropriately to the core spray pump discharge pressure indication. The error resulted in a low core spray loop pressure.
The condition existed since a 1986 surveillance procedure change.
The inspectors reviewed the Unit 2 core spray quarterly surveillance test procedures, TSs, and the procedure change review process.
The inspectors'n field review verified that the deficiencies were corrected, and procedures SO-251-A02/B02, "Quarterly Core Spray Flow Verification Division I and II,"were revised to remove the 7 psig correction factor from the pump discharge pressure calculation.
Since PP&L inappropriately revised the Unit 2 core spray quarterly surveillance tests in 1986, on two occasions the TS required flowwas not achieved for the core spray system.
On both occurrences, the TS action would have required the core spray loop to be returned to an operable status within seven days, or place the Unit in at least hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and in cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The TS required actions were not known or taken on both occasions.
The safety consequence was minimal because the slightly lower core spray discharge pressure would not have resulted in exceeding any reactor fuel design limits for the postulated design bases accidents.
The inspectors determined that PP&L properly identified and reported this issue, and found PP&L's proposed and completed corrective actions to be good.
In conclusion, in a Licensee Event Report, PP&L identified that on two occasions, the Unit 2 core spray quarterly flowsurveillance test did not meet the Technical Specification (TS) acceptance criteria due to a procedure error. PP&L's proposed and completed corrective actions, including procedure and programmatic actions, were good. This Severity Level IVviolation is being treated as a Non-Cited Violation, consistent with Appendix C of the NRC Enforcement Policy. This violation is in PP&L's corrective action program as condition report 98-3070.
LER 50-388/98-011 is closed.
(NCV 50-388/99-03-01)
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II~ Maintenance Conduct of Maintenance Surveillance and Pre-Planned Maintenance Activit Review 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.
Observations and Findin s The inspectors observed portions of the following work activities and surveillances:
Work Authorizations S85377 S98604 P82538 S85145 S90486 Z81 484 S83604 S90619 S90745 S90758 V90446 S83814 S 84214 Z90803 Unit 1 "B" Feedwater Flow Element Leak Repair Unit 2 "A" Drywell Hydrogen/Oxygen Analyzer l&C Repair Unit 2 "A" Drywell Hydrogen/Oxygen Analyzer I&C Leak Test Unit 1 RWCU Non-Regenerative Heat Exchanger Leak Repair Unit 1 RWCU Non-Regenerative Heat Exchanger Leak Repair
'nit 1 HELB Door-592 Blocked Open for RWCU Leak Repairs Unit 1 RCIC Instrument Root Valve Steam Leak Repair Unit 1 "B" RHR Keepfill System Investigation &Testing RHR HV-151F017B Internal Inspection and Repair RHR HV-151F017A Ultrasonic Test Feedwater Level Control System 1/3-Element Confidence Check Replace Belts on the "A"SBGT Fan Replace Bearing Oil Seals on the "A"SBGT Fan
"D" Emergency Diesel Jacket Water Drain and Fill Surveillances SO-152-002 SO-149-B05 SO-1 49-015 OP-278-001 SO-070-001 B SE-159-041 SE-273-400 Unit 1 HPCI Full Flow Surveillance Test RHR HV-151F017B Valve Stroke Test RHR HV-151F017B Remote Position Indication Test Traversing In-Core Probe System - Placed TIP detectors in shield Monthly Standby Gas Treatment Surveillance.
LLRTof Containment Purge Supply Valve Penetration No. ¹25/201A Unit 2 "A" Drywell Hydrogen/Oxygen Analyzer Local Leak Rate Test In addition, selected portions of procedures, drawings, and vendor technical manuals, associated with the maintenance and surveillance activities, were also reviewed and determined to be acceptable.
In general, maintenance personnel were very knowledgeable of their assigned activitie High Energy, Line Break (HELB) Door Control Following repairs to the Unit 1 RWCU non-regenerative heat exchanger, the inspectors observed that a HELB door remained blocked open and unattended.
This was of concern because it would allow a steam leak, from a normally closed RWCU room, to potentially enter an open ventilation damper into a 4Kv emergency switchgear room, located directly across from the open HELB door. The inspectors discussed the observed conditions and the required compensatory actions with operations shift personnel.
A HELB door watch was promptly stationed, the blockage through the door was cleared, and the door was shut. The failure to perform the required compensatory actions, as required by SSES procedures, constitutes a violation of minor significance and is not subject to formal enforcement action.
Ultrasonic Test of RHR Injection Valves HV-151F017A and HV-251F017B The inspectors observed the ultrasonic testing of HV-151F017A and HV-251F017B and concluded that the ultrasonic equipment calibration and test technique were adequate to verify significant indications existed in the valve shaft. Although the calibration and testing was insufficient to quantify the minimum size indication that could be detected, PP&L concluded that indications greater than 0.1 inches would be identified.
The inspe'ctors observed restoration activities, following the ultrasonic testing, including a valve timing stroke surveillance test, and concluded maintenance technicians and electricians'were very knowledgeable of the work they performed, and demonstrated a
good questioning attitude for unexpected conditions.
Conclusions The observation of ultrasonic testing of the residual heat removal injection valves determined that the'technique applied was effective in identifying significant stem cracking. The ultiasonic testing minimized personnel radiation exposure and resulted in the timely identification of an additional cracked valve stem.
The NRC identified a High Energy Line Break (HELB) door that was blocked open and left unattended, following repairs to a reactor water cleanup system heat exchanger.
The open HELB door was of concern because it would allow a steam leak, from a normally closed room, to potentially enter a 4Kv emergency switchgear-roo \\
Ill. En ineerin E2 Engineering Support of Facilities and Equipment E2.1 Residual Heat Removal Service Water Radiation Monitor a.
This inspector reviewed the operation of the residual heat removal service water (RHRSW) system radiation monitor.
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Observations and Findin s The RHRSW radiation monitors are maintained in a readiness status with the monitor electronics energized and no flowthrough the detector.
Flow through the radiation monitor is obtained by manually starting the sample pump in the Reactor Building. High area radiation levels would prevent starting the radiation monitor sample pump and may place the monitor in alarm after a postulated design bases accident. The radiation monitors are required to monitor and detect primary coolant leakage into the RHRSW system during normal operations, anticipated operational occurrences, and from postulated accidents.
Monitoring this effluent pathway is required because the RHRSW system is operated at a lower pressure than the RHR system.
The pressure difference willallow any primary RHR heat exchanger leakage to leak into the RHRSW system and be discharged to the'spray pond.
Design Basis Design basis document (DBD) 009, section 2.6.1.2.2, Revision 1 requires that "A means shall be provided for monitoring effluent discharge paths for radioactivity that may be released through RHRSW system piping and components during normal operations, including anticipated operational occurrences, and from postulated accidents."
This DBD further states "This requirement ensures any potential radioactive release through the RHRSW system are monitored and controlled during all plant operating conditions consistent with 10CFR50, Appendix A, General Design Criteria (GDC) 64 requirements."
Criterion 64, "Monitoring radioactivity releases" states that "Means shall be provided for monitoring the reactor containment atmosphere, spaces containing components for recirculation of loss-of-coolant accident fluids, effluent discharge paths, and the plant environs for radioactivity that may be released from normal operations, including anticipated operational occurrences, and from postulated accidents."
The DBD also states that since radiation monitors are provided downstream of each RHR heat exchanger certain original design specification requirements are no longer applicable.
The original design specification required operating the RHRSW system pressure at a higher level than the RHR system pressure to prevent radioactive release through the RHRSW system.
PP&L has taken advantage of the relaxation of this design specifications by installing RHRSW pumps that develop lower pressure at the RHR heat exchanger than the RHR pumps.
Therefore, any leakage of radioactive
material in the RHR heat exchanger will be transferred to the RHRSW system and discharged to the spray pond.
PP&L's Final Safety Analysis Report (FSAR) section 11.5, "Process and Effluent Radiological Monitoring and Sampling Systems" states that the RHRSW radiation monitoring system is to generally conform to GDC 60, 63, and 64 and Regulatory Guide 1.21, Rev. 1, dated June 1974. The technical requirements manual (TRM), section 3.11.1, "Liquid Effluents" requires one radiation monitor per RHRSW loop to be operable at all times. In addition, the basis for this TRM states in part "The OPERABILITYand use of this instrumentation is consistent with the requirements of General Design Criteria (GDC) 60, 63, and 64 of Appendix A to 10 CFR Part 50." Therefore, the RHRSW radiation monitors are required to monitor the RHRSW system discharge piping for radioactivity that may be released from postulated accidents.
Post Accident Condition PP8L will not be able to start the RHRSW radiation monitor manually, following a postulated design basis accident, nor will PP8L be able to collect a representative RHRSW sample at the monitor. The RHRSW radiation monitors (and sample pump controls) are located at elevation 645 and 683 in the Reactor Building. PP8L has concluded that the Reactor Building will be generally inaccessible for several days after the postulated design bases accident.
The Plant Shielding Analysis (FSAR section 18.1.2) calculated dose rates of 100 rem/hour in the general area of these radiation monitors and on elevation 645 these dose rates could be much greater.
The radiation exposure guidelines would not permit an operator to locally start these radiation monitors during an accident.
Ifthe radiation monitor was inoperable, TRM section 3.11.1, "Liquid Effluents" requires grab samples to be obtained and analyzed at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
PP8 L normally obtains grab samples locally at the radiation monitor. During post accident condition this sampling point willnot be available due to high dose rates.
PP&L stated these samples would be, obtained from the spray pond or the effluent stream to the spray pond.
Neither of these sample locations will be representative of the RHRSW effluent stream as required by Regulatory Guide 1.21. The spray pond as well as the RHRSW return to the spray pond have numerous effluent streams combined that prevent it from being a representative sample.
Therefore, during postulated accidents conditions RHRSW system. discharge piping willnot be able to be monitored as required by plant design.
Conclusions J
The residual heat removal service water (RHRSW) radiation monitors do not meet requirements, of General Design Criteria 64. Specifically, the radiation monitors would not be functioning following a postulated design basis accident, since the monitors can not be manually started locally in the high post accident area radiation levels.
In addition, the location of a backup grab sample would not provide a representative sample as delineated in Regulatory Guide 1.21. This is a violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," which requires, in part, selection of suitable equipment that
are essential to the safety related function of the system.
This Severity Level IVviolation is being treated as a Non-Cited Violation, consistent with Appendix C of the NRC Enforcement Policy. This violation is documented in PP&L's corrective action program as condition report 91031.
(NCV 50-388/99-03-02)
E2.2 Inde endentS ent Fuel Stora e Facili Activities a0 b.
PP&L is currently establishing an Independent Spent Fuel Storage Facility (ISFSF) to provide dry storage of spent fuel at SSES.
The scope of this inspection reviewed the available major project components, tasks and areas of the plant affected by the project.
Observations and Findin s The ISFSF project includes an outside storage pad, horizontal storage modules (HSM),
fuel canisters, fuel transfer cask, lifting and rigging equipment, cranes and load carrying vehicles, welding equipment, a vacuum drying system, helium leak testing equipment, a.
detailed operating procedure, work documentation packages and Quality Assurance overview. Prior to transfer of spent fuel from the fuel pool to the ISFSF dry fuel storage area, the operating procedure and system components willbe verified by conducting a fuel transfer demonstration, without spent fuel.
The inspectors toured the fuel transfer route both inside and outside the plant and reviewed portions of the applicable procedures.
The planned project activities and steps were discussed with the ISFSF Project Manager and Quality Assurance Analyst. As reported separately in NRC Inspection Report No. 72-1004/99-201, the casting of concrete HSM modules and roof slabs were observed at the concrete fabricator during the week of February 16, 1999.
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From the observations and review of project plans, the inspector noted that the ISFSF project planning is nearing completion. The construction of the storage pad and concrete aspects of the HSM,modules has been essentially completed.
The fabrication of storage canisters is in progress and willbe inspected by the NRC at a later time. The near term items to be completed by PP&L include finalizing the steps of the plan for the fuel transfer demonstration, without spent fuel to show that the equipment and procedures are workable.
No items of concern were identified by the inspector.
Conclusion The planning and construction of the Susquehanna Independent Spent Fuel Storage Facility (ISFSF) were being accomplished well and no significant concerns were identified. Quality Assurance involvement has been evident throughout the ISFSF projec W Uf
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Miscellaneous Engineering Issues Followu of 0 en Items Closed URI 50-387 388/98-07-06 (37551,92700,92903,40500)
Operation with Ultimate Heat Sink Above Maximum Allowed Temperature The NRC questioned the adequacy of the Technical Specification (TS) surveillance acceptance criteria for several monitored key parameters of the ultimate heat sink (UHS),
including UHS average water temperature, UHS level, and specific flow rate values for various emergency core cooling system (ECCS) components cooled from the UHS.
PP&L had not uniquely accounted for instrument uncertainties in the measurement of the key parameters used to satisfy TS acceptance criteria, and the margin for the associated instrument uncertainties was not apparent in the PP&L calculations for the analytical maximum temperature limitof the UHS. Unresolved Item (URI) 50-387,388/98-07-06 was opened to review a PP&L assessment of the margins available in the UHS analysis and an assessment of measurement uncertainty as applied to the surveillance procedures.
PP&L submitted a TS change request, in June 1998, to revise the UHS average water temperature surveillance requirement.
In conjunction with that request, PP&L responded to an NRC Request for Additional Information (RAI) on the effects of the UHS temperature instrument uncertainties on systems and components that interface with the UHS. The NRC is reviewing PP&L's response to the RAI, and willaddress any outstanding issues which arise as a result of the RAI review, as part of the requested TS amendment.
The inspectors reviewed PP&L calculation EC-016-1025, dated March 15, 1999, which verified that the TS surveillance limitwas adequate, including instrument uncertainties and assumed heat loss. The inspectors also reviewed PP&L calculation EC-01 6-1032, dated December 22, 1998, which quantified UHS temperature measurement uncertainties, to be applied to the analytical limitfor UHS average water temperature.
The inspectors concluded that the calculation methodology currently utilized by PP&L was appropriate to ensure the adequacy of the TS surveillance acceptance criteria.
In conclusion, the NRC reviewed the adequacy of the Technical Specification (TS)
surveillance acceptance criteria for several monitored key parameters of the Ultimate Heat Sink (UHS) and concluded PP&L had not uniquely accounted for instrument uncertainties.
PP&L took adequate corrective actions which included performance of several calculations to verify adequate margin existed in the analytical limitfor the UHS average water temperature limit, and submitted a TS amendment request.
No violations of NRC requirements were identified. This unresolved item is close V. Mana ement Meetin s X1 Exit Meeting Summary I
The inspectors presented the inspection results to members of PP&L management at the conclusion of the inspection period, on March 22, 1999.
PP&L acknowledged the findings presented.
The inspectors asked PP&L whether any materials examined during the inspection should be considered proprietary.
No proprietary information was identifie IP 37551 IP 40500 IP 60853 IP 60854 IP 61726 IP 62707 IP 71707 IP 71750 IP 92700 IP 92901 IP 92902 IP 92903 IP 92904 IP 93702 INSPECTION PROCEDURES USED Onsite Engineering Observations Effectiveness of Licensee Controls in Identifying, Resolving, and Preventing Problems On-site Fabrication of Components and Construction of an ISFSI Preoperational Testing of an ISFSI Surveillance Observations Maintenance Observations Plant Operations Plant Support Activities On Site Followup of Reports Followup Plant Operations Followup Maintenance Followup Engineering Followup Plant Support Prompt Onsite Response to Events at Operating Power Reactors
~Oened ITEMS OPENED, CLOSED, AND DISCUSSED NCV 50-388/99-03-01 NCV 50-388/99-03-02 Closed Closed Core Spray Quarterly Flow Surveillance Did Not Meet Acceptance Criteria (section 08.1)
Residual Heat Removal Service Water (RHRSW)
Radiation Monitor (section E2.1)
~Udated Closed LER 50-387/98-016-00 Closed Reactor Scram Due to Main Generator Trip (Section 08.1)
LER 50-388/98-011-00 Closed Core Spray Flow Surveillance Failed (Section 08.1)
URI 50-387,388/98-07-06 Closed Operation with Ultimate Heat Sink Above Maximum Allowed Temperature (Section E8.2)
CFR CR DBD ECCS FSAR HELB HPCI IS.C ISFSI IR Kv LCO LER NCV NDAP NRC OD RCIC RHR RHRSW RWCU SSES TS UHS URI UT LIST OF ACRONYMS USED Code of Federal Regulations Condition Report Design Basis Document Emergency Core Cooling System Final Safety Analysis Report High Energy Line Break High Pressure Coolant Injection Instrument and Controls Independent Spent Fuel Storage Facility
[NRC] Inspection Report Kilovolts Limiting Condition for Operation Licensee Event Report Non-Cited Violation Nuclear Department Administrative Procedure Nuclear Regulatory Commission Operability Determination Reactor Core Isolation Cooling Residual Heat Removal Residual Heat Removal Service Water Reactor Water Clean Up Susquehanna Steam Electric Station Technical Requirements Manual Technical Specification Ultimate Heat Sink
[NRC] Unresolved Item Ultrasonic Test