IR 05000220/2024002
ML24218A057 | |
Person / Time | |
---|---|
Site: | Nine Mile Point ![]() |
Issue date: | 08/05/2024 |
From: | Erin Carfang NRC/RGN-I/DORS |
To: | Rhoades D Constellation Energy Generation, Constellation Nuclear |
References | |
IR 2024002 | |
Download: ML24218A057 (1) | |
Text
August 5, 2024
SUBJECT:
NINE MILE POINT NUCLEAR STATION, UNITS 1 AND 2 - INTEGRATED INSPECTION REPORT 05000220/2024002 AND 05000410/2024002
Dear David Rhoades:
On June 30, 2024, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Nine Mile Point Nuclear Station, Units 1 and 2. On July 30, 2024, the NRC inspectors discussed the results of this inspection with Carl Crawford, Plant Manager and other members of your staff. The results of this inspection are documented in the enclosed report.
One finding of very low safety significance (Green) is documented in this report. This finding did not involve a violation of NRC requirements.
If you disagree with a cross-cutting aspect assignment or a finding not associated with a regulatory requirement in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region I; and the NRC Resident Inspector at Nine Mile Point Nuclear Station. This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely, Erin E. Carfang, Chief Projects Branch 1 Division of Operating Reactor Safety
Docket Nos. 05000220 and 05000410 License Nos. DPR-63 and NPF-69
Enclosure:
As stated
Inspection Report
Docket Numbers:
05000220 and 05000410
License Numbers:
Report Numbers:
05000220/2024002 and 05000410/2024002
Enterprise Identifier: I-2024-002-0041
Licensee:
Constellation Energy Generation, LLC
Facility:
Nine Mile Point Nuclear Station, Units 1 and 2
Location:
Oswego, NY
Inspection Dates:
April 1, 2024 to June 30, 2024
Inspectors:
C. Kline, Senior Resident Inspector
B. Sienel, Resident Inspector
E. Eve, Senior Project Engineer
Approved By:
Erin E. Carfang, Chief
Projects Branch 1
Division of Operating Reactor Safety
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Nine Mile Point Nuclear Station, Units 1 and 2, in accordance with the Reactor Oversight Process. The Reactor Oversight Process is the NRCs program for overseeing the safe operation of commercial nuclear power reactors. Refer to https://www.nrc.gov/reactors/operating/oversight.html for more information.
List of Findings and Violations
Incorrect Application of Maintenance Strategy Process Leads to Loss of Feedwater Flow and Automatic Reactor Scram Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green FIN 05000410/2024002-01 Open/Closed
[P.3] -
Resolution 71153 A self-revealed Green finding was identified for Constellations failure to adequately evaluate the impact of increased feed water level control valve cycling on balance seal longevity in accordance with ER-AA-200, "Preventive Maintenance Program." Specifically, following installation of the Unit 2 digital feedwater level control (DFWLC) system, Constellation identified that level control valve 2FWS-LV10 cycle frequency increased approximately five-fold. Constellation recognized the increased cycling could accelerate wear of the Teflon balance seal and valve packing. The site did not appropriately assess the increased wear and adjust the maintenance interval for internal valve software inspection and replacement. The increased cycle rate led to premature balance seal failure and internal valve flow instability, which resulted in stem/disc connection failure. The separation of the disc from the valve stem resulted in partial loss of feedwater flow, and an automatic reactor scram on September 2, 2023.
Additional Tracking Items
Type Issue Number Title Report Section Status LER 05000410/2023-001-00 Automatic Reactor Scram on Low Level Due to Partial Loss of Feedwater 71153 Closed LER 05000410/2023-001-01 Supplement to LER 2023-001-00, Automatic Reactor Scram on Low Level Due to Partial Loss of Feedwater 71153 Closed
PLANT STATUS
Unit 1 began the inspection period at rated thermal power. On April 12, 2024, the unit reduced power to 65 percent for rod sequence exchange, scram time testing, reheater steam leak repairs, and turbine valve testing. The unit returned to rated thermal power on April 13, 2024.
On April 14, 2024, the unit reduced power to 90 percent for rod line adjustment and returned to rated thermal power the same day. On May 23, 2024, the unit reduced power to 95 percent to remove reactor recirculating pump (RRP) 12 from service for corrective maintenance and returned to rated thermal power the same day. On May 23, 2024, the unit reduced power to 85 percent to return RRP 12 to service and returned to rated thermal power the same day. On May 25, 2024, the unit reduced power to 95 percent to remove RRP from service due to high vibrations and returned to rated thermal power the same day. On May 26, 2024, the unit reduced power to 85 percent to return RRP 12 to service and returned to rated thermal power the same day.
Unit 2 began the inspection period at rated thermal power. On April 4, 2024, the unit reduced power to 94 percent in response to a feedwater heater transient. The unit returned to rated thermal power April 5, 2024. On May 17, 2024, the unit reduced power to 74 percent for rod pattern adjustment and sequence exchange. On May 18, 2024, the unit returned to rated thermal power.
INSPECTION SCOPES
Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors performed activities described in IMC 2515, Appendix D, Plant Status, observed risk significant activities, and completed on-site portions of IPs. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.
REACTOR SAFETY
71111.01 - Adverse Weather Protection
Seasonal Extreme Weather Sample (IP Section 03.01) (1 Sample)
The inspectors evaluated readiness for seasonal extreme weather conditions prior to the onset of seasonal hot temperatures for the following systems:
- (1) Unit 1 emergency diesel generators 102 and 103 and Unit 2 high pressure core spray on May 22, 2024
71111.04 - Equipment Alignment
Partial Walkdown Sample (IP Section 03.01) (4 Samples)
The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:
- (1) Unit 2 reactor core isolation cooling system on April 19, 2024
- (2) Unit 2 high pressure core spray system on April 25, 2024
- (3) Unit 2 standby liquid control system on May 13, 2024
- (4) Unit 1 containment spray system on June 12, 2024
Complete Walkdown Sample (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated system configurations during a complete walkdown of the Unit 1 core spray system on June 24, 2024.
71111.05 - Fire Protection
Fire Area Walkdown and Inspection Sample (IP Section 03.01) (7 Samples)
The inspectors evaluated the implementation of the fire protection program by conducting a walkdown and performing a review to verify program compliance, equipment functionality, material condition, and operational readiness of the following fire areas:
- (1) Unit 2 turbine building 277', west switchgear room, fire area 84, on April 11, 2024
- (2) Unit 2 turbine building 250' west, fire area 50, on April 11, 2024
- (3) Unit 1 turbine building 369' storage tank area, fire area 5, on May 22, 2024
- (4) Unit 1 turbine building 289' auxiliary extension, fire area 5, on May 22, 2024
- (5) Unit 1 reactor building east 198' to 237' core spray 12 areas, fire area 1, on June 24, 2024
- (6) Unit 1 reactor building west 198' to 237' core spray 11 areas, fire area 2, on June 24, 2024
- (7) Unit 2 control building 261' Division I diesel room, fire area 28, on June 25, 2024
Fire Brigade Drill Performance Sample (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated the onsite fire brigade training and performance during an unannounced fire drill on May 31, 2024.
71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance
Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01) (2 Samples)
- (1) The inspectors observed and evaluated Unit 2 licensed operator performance in the control room during a power reduction to 74 percent and rod pattern adjustment on May 17, 2024.
- (2) The inspectors observed and evaluated Unit 1 licensed operator performance in the control room during reactor protection system and fire pump surveillances on June 17, 2024.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (3 Samples)
- (1) The inspectors observed a Unit 2 simulator evaluation that included rapidly lowering intake bay levels, loss of a low pressure reheater string, 'B' feed pump trip and recirculating pump runback, and unisolable leak from the reactor core isolation cooling system on May 7, 2024.
- (2) The inspectors observed a Unit 1 simulator evaluation that included a seismic event, recirculation pump seal failure, and large break loss of coolant accident on May 14, 2024.
- (3) The inspectors observed a Unit 2 simulator evaluation that included an instrument air compressor trip, control rod drive pump trip, stuck open primary relief valve, reactor scram, and loss of coolant casualty on June 18, 2024.
71111.12 - Maintenance Effectiveness
Maintenance Effectiveness (IP Section 03.01) (1 Sample)
The inspectors evaluated the effectiveness of maintenance to ensure the following structures, systems, and components (SSCs) remain capable of performing their intended functions:
- (1) Review of periodic evaluation in accordance with Title 10 of the Code of Federal Regulations (10 CFR) 50.65(a)(3) on April 9, 2024
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Risk Assessment and Management Sample (IP Section 03.01) (5 Samples)
The inspectors evaluated the accuracy and completeness of risk assessments for the following planned and emergent work activities to ensure configuration changes and appropriate work controls were addressed:
- (1) Unit 2 elevated risk during Division II emergency diesel generator system outage window on April 29, 2024
- (2) Unit 1 elevated risk during 122 core spray system and 121 containment spray system planned maintenance on May 7, 2024
- (3) Unit 1 elevated risk due to failure of the Unit 1 diesel fire pump monthly surveillance on May 14, 2024
- (4) Unit 2 elevated risk due to high pressure core spray unit cooler cleaning and capacity test on May 21, 2024
- (5) Unit 1 elevated risk due to battery charger 161B out of service on May 29, 2024
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (6 Samples)
The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments:
- (1) Unit 2 reactor core isolation cooling turbine bearing oil pressure greater than alert range on April 18, 2024
- (2) Unit 2 automatic depressurization system safety relief valve 127 stem nut unable to be adjusted on May 13, 2024
- (3) Unit 2 nitrogen leak on supply line to automatic depressurization system accumulator on June 3, 2024
- (4) Unit 2 service water spring can blocked on June 3, 2024
- (5) Unit 2 starting air system malfunction alarm for Division I diesel on June 17, 2024
- (6) Unit 2 service water pump 1F strainer backflush valve blockage on June 26, 2024
71111.24 - Testing and Maintenance of Equipment Important to Risk
The inspectors evaluated the following testing and maintenance activities to verify system operability and/or functionality:
Post-Maintenance Testing (PMT) (IP Section 03.01) (9 Samples)
- (1) N2-OSP-EGS-R001, Diesel Generator ECCS Start and Load Reject Division I and II, following diesel generator slow start modification, on April 10, 2024
- (2) N2-MFT-327B, Digital Feedwater Level Control System Power Ascension Test, following actuator modification during refueling outage N2R19, on April 11, 2024
- (3) N2-OSP-RPV-R@003, Reactor Pressure Vessel and Class I Systems Leakage Test, with the reactor pressure vessel solid, following refueling outage N2R19, on April 11, 2024
- (4) N2-OSP-EGS-M@001 Diesel Generator and Diesel Air Start Valve Operability Test-Division I and II, following Division II emergency diesel generator system outage window, on April 30, 2024
- (5) N1-ST-Q8B Liquid Poison Pump 12 and Check Valve Operability Test, following test switch replacement and breaker testing, on April 30, 2024
- (6) N2-OSP-SLS-Q001 Standby Liquid Control Pump Check Valve Relief Valve Operability Test and ASME XI Pressure Test, following pump valve seat inspection, on May 2, 2024
- (7) N1-ST-Q1D, Core Spray 122 Pump and Valve Operability Test, following core spray topping pump 122 seal replacement, on May 10, 2024
- (8) N1-ST-Q1C, Core Spray 112 Pump and Valve Operability Test, following valve preventive maintenance, on June 11, 2024
- (9) N1-PM-C3, Diesel Fire Pump, Combined Auto Start Orifice Flow and Low Air Pressure Operability Test, following diesel fire pump replacement, on June 13, 2024
Surveillance Testing (IP Section 03.01) (3 Samples)
- (1) N2-OSP-EGS-R007, Operating Cycle Diesel Generator Simulated Loss of Offsite Power Division III, on April 23, 2024
- (2) N1-ST-M4B, Emergency Diesel Generator 103 and PB 103 Operability Test, on May 6, 2024
- (3) N2-OSP-EGS-R008, Operating Cycle Diesel Generator Simulated Loss of Offsite Power with an Emergency Core Cooling System Division III Initiation, on May 23, 2024
71114.06 - Drill Evaluation
Required Emergency Preparedness Drill (1 Sample)
- (1) On June 11, 2024, the inspectors evaluated a Unit 2 emergency preparedness drill which included Alert, Site Area Emergency, and General Emergency event declarations.
Additional Drill and/or Training Evolution (1 Sample)
- (1) The inspectors evaluated a Unit 2 simulator scenario that included a stuck open primary relief valve, reactor scram, and loss of coolant casualty that required an Alert declaration on June 18,
OTHER ACTIVITIES - BASELINE
===71151 - Performance Indicator Verification
The inspectors verified licensee performance indicators submittals listed below:
BI01: Reactor Coolant System (RCS) Specific Activity Sample (IP Section 02.10)===
- (1) Unit 1 (for the period of April 1, 2023 through March 31, 2024)
- (2) Unit 2 (for the period of April 1, 2023 through March 31, 2024)
BI02: RCS Leak Rate Sample (IP Section 02.11) (2 Samples)
- (1) Unit 1 (for the period of April 1, 2023 through March 31, 2024)
- (2) Unit 2 (for the period of April 1, 2023 through March 31, 2024)
===71152S - Semiannual Trend Problem Identification and Resolution
Semiannual Trend Review (Section 03.02)===
- (1) The inspectors reviewed Constellations corrective action program to identify potential trends that might be indicative of a more significant safety issue.
71153 - Follow Up of Events and Notices of Enforcement Discretion Event Report (IP Section 03.02)
The inspectors evaluated the following licensee event reporting determination to ensure it complied with reporting requirements.
- (1) LERs 05000410/2023-001-00 and 05000410/2023-001-01, Automatic Reactor Scram on Low Level Due to Partial Loss of Feedwater, and Supplement to the Automatic Reactor Scram on Low Level Due to Partial Loss of Feedwater (ADAMS Accession Nos. ML23306A048 and ML24060A053, respectively). The inspectors reviewed the original and updated LER submittals. The inspection conclusions associated with these LERs are documented in this report under Inspection Results. These LERs are closed.
INSPECTION RESULTS
Observation: Semiannual Trend 71152S The inspectors evaluated a sample of issues and events that occurred from January 2024 through June 2024 to determine whether issues were appropriately considered as emerging or adverse trends. The inspectors verified issues were appropriately evaluated by Constellation staff for potential trends and addressed within the scope of the corrective action program or through department review.
The inspectors did not identify any new trends that could indicate a more significant safety issue. The inspectors determined that Constellation personnel identified trend issues at a low threshold and entered them into the corrective action program for resolution. The inspectors noted that issues were identified with the Unit 1 diesel fire pump engine reliability even after it was replaced in late 2023. The engine was again replaced this quarter. The inspectors will continue to monitor the reliability of this equipment through routine inspection.
Incorrect Application of Maintenance Strategy Process Leads to Loss of Feedwater Flow and Automatic Reactor Scram Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events
Green FIN 05000410/2024002-01 Open/Closed
[P.3] -
Resolution 71153 A self-revealed Green Finding was identified for Constellations failure to adequately evaluate the impact of increased feedwater level control valve cycling on balance seal longevity in accordance with ER-AA-200, "Preventive Maintenance Program." Specifically, following installation of the Unit 2 digital feedwater level control (DFWLC) system, Constellation identified that level control valve 2FWS-LV10 cycle frequency increased approximately five-fold. Constellation recognized the increased cycling could accelerate wear of the Teflon balance seal and valve packing. The site did not appropriately assess the increased wear and adjust the maintenance interval for internal valve software inspection and replacement. The increased cycle rate led to premature balance seal failure and internal valve flow instability, which resulted in stem/disc connection failure. The separation of the disc from the valve stem resulted in partial loss of feedwater flow, and an automatic reactor scram on September 2, 2023.
Description:
On September 2, 2023, at approximately 0632, Unit 2 experienced an automatic reactor scram on low reactor water level due to a sudden loss of feedwater flow, downstream of the B feedwater pump 2FWS-P1B. Operations personnel received multiple alarms and identified that reactor water level was rapidly lowering. Shortly after receiving initial alarm indications, the reactor scrammed on low reactor water level and operators entered N2-EOP-RPV, "Reactor Pressure Valve Control for Modes 1-3."
The DFWLC system controls reactor water level during normal power operations using two of three electrically-driven, constant speed feed pumps which discharge to pneumatically controlled level control valves 2FWS-LV10 A, B, or C (LV10s). The flow rate is controlled by a feedwater controller which controls level control valve position based on inputs from reactor vessel water level, steam flow rate, and feedwater flow rate transmitters. In March 2020, Nine Mile Point implemented a Unit 2 DFWLC system, per approved engineering change ECP-17-00688. The new system modified valve actuators but retained the feed flow throttle valves from the legacy design. The legacy design feed flow throttle valves regulate flow by positioning a cylindrical plug within a perforated disc stack. The plug is positioned vertically within the disc stack by the valve stem, covering or uncovering flow ports as necessary to raise or lower flow. The disc stack is ported to promote uniform circumferential flow distribution that is required for high pressure, high flow applications. A balance seal keeps the plug centered within the disc stack and ensures leak tightness. The legacy design required the replacement of the stem and plug assembly and soft goods at 6-year intervals. The inspectors noted that at the time of modification, the 6-year overhaul preventive maintenance for LV-10B was due in 2024 and no internal valve inspection or repairs were conducted.
Post-modification testing in 2020 identified several issues with valve response, including that the valves cycled roughly five times more frequently than the legacy control system. A work group evaluation was conducted to address DFWLC design deficiencies and performance issues, and a digital feedwater team was established to evaluate and correct the unexpected valve response.
In May 2020, the digital feedwater team recognized and documented that the increased cycling of the LV10s would accelerate wear of the Teflon balance seal and valve packing.
The inspectors noted that, in July 2020, Constellation initiated maintenance action to advance valve packing replacement by two years. However, contrary to ER-AA-200, "Preventive Maintenance Program," Section 4.6, "Maintenance Strategy Analysis and Optimization,"
Step 4 which states, If trend or predictive maintenance predicts significant equipment degradation or equipment failure before the next scheduled PM then Engineering or Maintenance should initiate an issue report (IR) per the Corrective Actions Process and ensure a work request is created for corrective maintenance, Constellation failed to create work orders to verify the condition of the balance seals before the expiration of the legacy 6-year maintenance interval.
Following the scram on September 2, 2023, 2FWS-LV10B was disassembled for inspection.
Inspections revealed that the plug was disconnected from the stem. Inspections also noted that an anti-rotation pin was absent from a tab washer that was welded to the valve stem.
Remnants of the Teflon balance seal assembly were located on the top surface of the plug.
Subsequent failure analysis determined that the failed balance seal allowed high pressure fluid to pass between the plug and plug cage to the region above the plug causing hydraulic instability, side loading, and vibration on the stem/plug assembly. Over time the anti-rotation pin welded connection experienced fatigue failure and separated. Combined with the vibration, this resulted in loss of preload between the stem and plug, which allowed the stem to unthread, fall into the flow stream, and stop feed flow through the 'B' feed header, causing an automatic reactor scram on low reactor water level.
Corrective Actions: Following the reactor scram, Constellation replaced the stem and disc assembly and installed an engineering change to improve the stem to plug anti-rotation connection on 2FWS-LV10B (additional welds between lock washer and plug). Additionally, Constellation increased maintenance frequency to replace internal valve software.
Corrective Action References: IRs 04700148, 04334474, and 04332838
Performance Assessment:
Performance Deficiency: Constellation's failure to initiate an IR and create a work order request for maintenance on the level control valve balance seals per ER-AA-200, Preventive Maintenance Program, when performance monitoring indicated accelerated wear that represented a significant risk to component reliability, was a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to modify maintenance periodicity for replacing the balance seal, given the significantly increased valve cycle rate, resulted in ring failure which caused a feedwater transient resulting in an unplanned automatic reactor scram.
Significance: The inspectors assessed the significance of the finding using IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors assessed significance using Question B, Transient Initiators, of Exhibit 1, Initiating Events Screening Questions, and determined the finding did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. Therefore, the inspectors determined the finding to be of very low safety significance (Green).
Cross-Cutting Aspect: P.3 - Resolution: The organization takes effective corrective actions to address issues in a timely manner commensurate with their safety significance. Specifically, corrective actions did not resolve and correct the impact of increased level control valve cycling on the balance seal.
Enforcement:
Inspectors did not identify a violation of regulatory requirements associated with this finding.
EXIT MEETINGS AND DEBRIEFS
The inspectors confirmed that proprietary information was controlled to protect from public disclosure.
- On July 30, 2024, the inspectors presented the integrated inspection results to Carl Crawford, Plant Manager and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Procedures
N1-OP-64
Meteorological Monitoring
2100
N2-OP-102
Meteorological Monitoring
2800
Seasonal Readiness
Work Orders
C93920534
C93938972
C93976602
Corrective Action
Documents
04703406
04777820
Drawings
C-18007-C
Reactor Core Spray Piping and Instrumentation Diagram
064
C-18012-C
Reactor Containment Spray System
047
Procedures
N1-OP-14
Containment Spray System
048
N1-OP-2
Core Spray System
03900
N2-OP-33-
Lineups
High Pressure Core Spray System - Lineups
2
N2-OP-36A-
Lineups
Standby Liquid Control System - Lineups
000
Fire Plans
N1-PFP-0101
Unit 1 Pre-Fire Plans
00700
N2-FPI-PFP-0201 Unit 2 Pre-Fire Plans
007
Miscellaneous
DCD-805
Nine Mile Point Unit 1 NFPA 805 Design Criteria
Procedures
N1-FPM-FPE-
A001
Annual Inspection of Portable Fire Extinguishers
Fire Drill Performance
Control of Transient Combustible Material
71111.07A Work Orders
C93852040
Corrective Action
Documents
04774076
Corrective Action
Documents
Resulting from
Inspection
04772681
Procedures
N1-PM-M9
Operation of Fire Pumps
019
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Work Orders
C93953647
Corrective Action
Documents
04760604
04773482
Procedures
N2-OSP-ADS-
R106
Automatic Depressurization System Accumulator and
Pneumatic Supply Leak Rate Test
001T2
N2-OSP-ICS-
Q@002
RCIC Pump and Valve Operability Test and System Integrity
Test and ASME XI Functional Test and Analysis
018
Work Orders
C93849925
C93976604
Procedures
N1-NMP-081-113 Overhaul of Core Spray Topping Pump and Accessories
Inspection
00600
N1-PM-C3
N1-PM-C3 Diesel Fire Pump, Combined Auto Start Orifice
Flow and Low Air Pressure Operability Test
019
N1-ST-M4B
Emergency Diesel Generator 103 and PB 103 Operability
Test
28
N1-ST-Q1C
Core Spray 112 Pump and Valve Operability Test
23
N1-ST-Q1D
Core Spray 122 Pump and Valve Operability Test
26
N1-ST-Q8B
Liquid Poison Pump 12 and Check Valve Operability Test
016
N2-OSP-EGS-
M@001
Diesel Generator and Diesel Air Start Valve Operability Test-
Divisions I and II
26
N2-OSP-EGS-
R007
Operating Cycle Diesel Generator Simulated Loss of Offsite
Power Division III
010
N2-OSP-EGS-
R008
Operating Cycle Diesel Generator Simulated Loss of Offsite
Power with an ECCS Division III Initiation
013
N2-OSP-SLS-
Q001
Standby Liquid Control Pump Check Valve Relief Valve
Operability Test and ASME XI Pressure Test
24
Work Orders
C93766938
C93793759
C93958210
C93975470
71151
Miscellaneous
Regulatory Assessment Performance Indicator Guideline
71152S
Corrective Action
Documents
04758100
04774076
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
04774366