IR 05000382/2023010
| ML24029A254 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 01/31/2024 |
| From: | Vincent Gaddy NRC/RGN-IV/DORS/EB1 |
| To: | Sullivan J Entergy Operations |
| Cullum C | |
| References | |
| IR 2023010 | |
| Download: ML24029A254 (20) | |
Text
January 31, 2024
SUBJECT:
WATERFORD STEAM ELECTRIC STATION - COMPREHENSIVE ENGINEERING TEAM INSPECTION REPORT 05000382/2023010
Dear Joseph Sullivan:
On January 22, 2024, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Waterford Steam Electric Station and discussed the results of this inspection with you 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 involved a violation of NRC requirements and was determined to be Severity Level IV. We are treating this violation as a non-cited violation (NCV) consistent with Section 2.3.2 of the Enforcement Policy.
If you contest the violation or the significance or severity of the violation documented in this inspection report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN:
Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region IV; the Director, Office of Enforcement; and the NRC Resident Inspector at Waterford Steam Electric 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, Vincent G. Gaddy, Chief Engineering Branch 1 Division of Operating Reactor Safety Signed by Gaddy, Vincent on 01/31/24 Docket No. 05000382 License No. NPF-38
Enclosure:
As stated
Inspection Report
Docket No.
05000382
License No.
Report No.
Enterprise Identifier:
I-2023-010-0020
Licensee:
Entergy Operations, Inc.
Facility:
Waterford Steam Electric Station
Location:
Killona, LA
Inspection Dates:
November 27, 2023 to December 15, 2023
Inspectors:
J. Braisted, Senior Reactor Inspector
D. Bryen, Reactor Inspector
W. Cullum, Senior Reactor Inspector
C. Franklin, Reactor Inspector
N. Mentzer, Reactor Inspector
E. Rosario, Reactor Inspector
F. Thomas, Reactor Inspector
C. Young, Senior Reactor Analyst
Approved By:
Vincent G. Gaddy, Chief
Engineering Branch 1
Division of Operating Reactor Safety
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting a Comprehensive Engineering Team Inspection at Waterford Steam Electric Station, 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
Failure to Verify the Adequacy of Design of the Alternate Source Term Dose Consequence Analysis and Provide Complete and Accurate Information to the Commission Cornerstone Significance/Severity Cross-Cutting Aspect Report Section Barrier Integrity Green Severity Level IV NCV 05000382/2023010-01 Open/Closed None (NPP)71111.21M The inspectors identified a Green finding and associated Severity Level IV Non-cited Violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion III, Design Control, and 10 CFR 50.9, Completeness and Accuracy of Information, when the licensee failed to verify the adequacy of design of the loss of coolant accident alternative source term radiological dose consequence analysis and provide related information to the Commission that was complete and accurate in all material respects.
Additional Tracking Items
None.
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 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
===71111.21M - Comprehensive Engineering Team Inspection The inspectors evaluated the following components and listed applicable attributes, permanent modifications, and operating experience:
Structures, Systems, and Components (SSCs) (IP section 03.01)===
- Updated final safety analysis report (UFSAR), technical specifications (TS),and TS bases document
- Motor-operated valve (MOV) program documents, in-service testing (IST)plan, and SI system design basis document
- Design bases review calculation for SI-602A, including evaluations of torque switch settings and weakest link
- Valve and actuator vendor manuals for installation, operation, and maintenance
- Inservice testing records including procedures and results for position indication, leakage, opening, and closing
- Diagnostic testing records including procedures and results
- Maintenance work order history records and visual inspections
- Visual inspection of the as-built configuration and material condition
- (2) Refueling Water Storage Pool
- Updated final safety analysis report (UFSAR), technical specifications (TS),and TS bases document
- Safety injection (SI) system design basis document
- Calculations for refueling water storage pool (RWSP) volume design requirements and required submergence to prevent vortexing
- Records of chemistry samples for temperature, boron concentration, and pH
- SI system piping and instrumentation diagram
- Maintenance work order history records of visual inspections
- Visual inspection of the as-built configuration and material condition
- Updated final safety analysis report (UFSAR), technical specifications (TS),and TS bases document
- Air-operated valve (AOV) program documents, in-service testing (IST) plan, and CCW system design basis document
- Design bases review calculation for CC-200B
- Valve and actuator vendor manuals for installation, operation, and maintenance
- Inservice testing records including procedures and results for position indication, opening, and closing
- Diagnostic testing records including procedures and results
- Maintenance work order history records and visual inspections
- Visual inspection of the as-built configuration and material condition
- (4) Component Cooling Water Surge Tank
- Updated final safety analysis report (UFSAR), technical specifications (TS),and TS bases document
- Component Cooling Water (CCW) system design basis document
- Calculations for CCW surge tank volume design requirements and required submergence to prevent vortexing
- CCW Surge tank level & Makeup circuit design
- Records of internal coating inspection, and CCW Chemistry controls including biocide,
- CCW system piping and instrumentation diagram
- Maintenance work order history records of visual inspections
- Visual inspection of the as-built configuration and material condition
- (5) Emergency Diesel Generator (EDG) A Room Exhaust Fan
- Updated final safety analysis report (UFSAR), technical specifications (TS),and TS bases document
- EDG room ventilation system design basis document
- Calculations for EDG heat load and required ventilation air flow
- Records of EDG ventilation seismic qualification and system testing
- EDG room ventilation system diagram
- Maintenance work order history records and visual inspections
- Visual inspection of the as-built configuration and material condition (6)4kV-ESWGR-3B, 4.16kV Switchgear 3B3-S
- Updated final safety analysis report (UFSAR), technical specifications (TS),and TS bases document
- Design calculations
- Design drawings
- Diagnostic testing records including procedures and results
- Maintenance work order history records
- Visual inspection of the as-built configuration and material condition
- (7) High Pressure Safety Injection Header Cold Leg Flow Control Isolation Valve SI-226A
- Updated final safety analysis report (UFSAR), technical specifications (TS),and TS bases document
- Motor-operated valve (MOV) program documents, in-service testing (IST)plan, and SI system design basis document
- Design bases review calculation for SI-226A, including evaluations of torque switch settings and weakest link
- SI system piping and instrumentation diagram
- Valve and actuator vendor manuals for installation, operation, and maintenance
- Inservice testing records including procedures and results for position indication, opening, and closing
- Diagnostic testing records including procedures and results
- Maintenance work order history records and visual inspections
- Visual inspection of the as-built configuration and material condition
- Updated final safety analysis report (UFSAR), technical specifications (TS),and TS bases document
- EFW System design basis document
- Calculations for pump head curve and NPSH
- EFW system piping and instrumentation diagram
- Pump vendor manuals for installation, operation, and maintenance
- Inservice testing records including procedures and results
- Maintenance work order history records and visual inspections
- Visual inspection of the as-built configuration and material condition (9)480 V Switchgear Bus 3B31-S (SSD-ESWGR-31B)
- Updated final safety analysis report, technical specifications, and technical specifications bases document.
- Visual inspection of switchgear for as-built configuration and material condition
- Environmental conditions for equipment
- Voltage analysis and short circuit calculations
- Protection settings and coordination study
- Vendor manual for conformance with manufacturer instructions for installation and maintenance
- Preventive maintenance and periodic verification testing
- Component maintenance history
- Schematic and wiring drawings
- System health report
- System health report
- Maintenance, inspection, and testing procedures
- Component maintenance history
- Schematic and wiring drawings
- Vendor operating and maintenance requirements
- Short circuit calculations, coordination studies, load flow calculations, and maintenance activities to ensure they were appropriate for the design of the Class 1E DC 3B-DC-S Power Distribution Panel
- Corrective action documents
Modifications (IP section 03.02) (5 Samples)
- (1) EC-77544, Broad Range Gas Monitor (BRGM) System Replacement
- (2) EC-83354, Replace SI-512A with a Swing Check Valve
- (4) EC-83913, Main Steam Isolation Valve (MSIV) Solenoid Valve Upgrade
- (5) EC-84942, AMTEK Solid State Controls Component Substitution Justification
10 CFR 50.59 Evaluations/Screening (IP section 03.03) (18 Samples)
- (1) EC-75220, Evaluation of Impacted Safety Related Calculations
- (2) EC-77544, Broad Range Gas Monitor System Replacement
- (4) EC-88373, Stroke Time Evaluation of Various SI, CVR, SBV Valves
- (5) EC-90132, Clarification of Seismic Design Requirements for Piping and Valves Upstream of EBA Tank Isolation Valves (EBA-203, EBA-204, 4BA1/2-5)
===90302 for Essential Chillers RFRMCHL0001 A, B, C
- (8) EC-92876, Evaluate Use of Alloy 690 Thermowells for Cold Leg RTDS
- (9) EC-94658, FHD-232A Leak Repair Evaluation
- (12) PAD-OP-902-005 Screen, Station Blackout Recovery Procedure Change
- (13) PAD-OP-901-310 Screen, Loss of Train A Safety Bus Procedure Change
- (16) EC-94858, ECE91-053 12IAV55C Acceptance Range Change
- (17) EC-95070, Temporary Load Bank for Integrated Diesel Test A Train Turbine Control System
- (18) EC-90044, Temp Indication Switch Standard EN-IC-S-025-W
Operating Experience Samples (IP section 03.04)===
- (1) OE-NOE-2022-00067, INPO IER L3 22-3 Power Excursion and Delayed Scram During Zero-Power Physics Testing
- (2) OE-NOE-2020-00166, IRIS 475373 Manual Scram due to Electro-Hydraulic Control System Levels
- (3) OE-NOE-2022-00192, IRIS 535302 Loss of Power to Emergency Class Buses
INSPECTION RESULTS
Failure to Verify the Adequacy of Design of the Alternate Source Term Dose Consequence Analysis and Provide Complete and Accurate Information to the Commission Cornerstone Significance/Severity Cross-Cutting Aspect Report Section Barrier Integrity Green Severity Level IV NCV 05000382/2023010-01 Open/Closed None (NPP)71111.21M The inspectors identified a Green finding and associated Severity Level IV Non-cited Violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion III, Design Control, and 10 CFR 50.9, Completeness and Accuracy of Information, when the licensee failed to verify the adequacy of design of the loss of coolant accident alternative source term radiological dose consequence analysis and provide related information to the Commission that was complete and accurate in all material respects.
Description:
Waterford submitted license amendment request (LAR) 198 (ML042020294) to the NRC for full implementation of an alternative radiological source term (AST) for calculating offsite radiological doses and doses to control room personnel during design basis accidents (DBAs). The NRC approved the request on March 29, 2005, in a letter to Waterford (ML050890248) which also contained its safety evaluation of the change. NRC Regulatory Guide (RG) 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, dated July 2000, provides guidance on methods acceptable to the NRC for implementing an AST, and Waterford described use of this guidance in their original LAR submittal in addition to other relevant correspondence submitted to the NRC (ML042470194, ML042890193, and ML043010129). As stated in RG 1.183, full implementation of the AST revises the plant licensing basis to specify the AST in place of the previous accident source term, and the DBA loss of coolant accident (LOCA)must be re-analyzed using the guidance in Appendix A. However, licensees may use methods other than those described in RGs (i.e., exceptions) for compliance with regulations if they provide an acceptable basis.
In the NRCs safety evaluation of the dose consequences of Waterfords DBA LOCA, the NRC stated, based upon information provided in correspondence, that Waterford performed an analysis of the radiological consequences of a (large break) LBLOCA for the (extended power uprate) EPU conditions using the guidance on source term and activity release in RG 1.183. Waterford evaluated three release pathways in the LBLOCA dose analysis: (1)leakage to the reactor auxiliary building serviced by the controlled ventilation area system (CVAS);
- (2) leakage to the secondary containment that is serviced by the shield building ventilation system (SBVS); and
- (3) leakage from the containment directly to the environment.
Additionally, emergency core cooling system (ECCS) components outside containment were assumed to leak 0.5 gallons per minute (gpm) of containment sump water into areas serviced by the CVAS, starting when the ECCS goes into recirculation mode. The CVAS and SBVS are safety-related systems that may be credited to limit the radiological dose consequences offsite and in the control room in accordance with RG 1.183.
Appendix A to RG 1.183 describes the assumptions that are acceptable to the NRC for evaluating the radiological consequences of a LOCA at a light water reactor. Section 5 of Appendix A provides the specific details for evaluating engineered safety feature (ESF)system leakage. These are systems that recirculate sump water outside of the primary containment, such as the ECCS at Waterford. The release source includes leakage through valve packing glands, pump shaft seals, flanged connections, and other similar components. The source may also include leakage through valves isolating interfacing systems. The guidance goes on to state that consideration should be given to leakage through valves isolating ESF recirculation systems from tanks vented to atmosphere, e.g.,
ECCS pump miniflow return to the refueling water storage tank. At Waterford, the equivalent tank is called the refueling water storage pool (RWSP). Finally, leakage should be taken as two times the sum of the simultaneous leakage from all components in the ESF recirculation systems.
The inspectors reviewed Procedure OP-903-110, Reactor Auxiliary Building (RAB) Systems Fluid Leak Test, which Waterford cited as the basis for the ESF system leakage criteria of 0.5 gpm. The purpose of OP-903-110 is to provide leak rate tests on those portions of ESF systems outside containment that could contain highly radioactive fluid during a serious transient or accident. The inspectors observed that the ECCS pumps take suction from the RWSP and return it to the RWSP through the miniflow return lines during the test. In other words, certain valves that would normally be closed and isolate the RWSP from the ECCS during the recirculation phase of a LOCA are open during the test, so the test does not verify that the system leakage criteria is met. Additionally, the inspectors observed that the RWSP is not vented to the CVAS but to the nonsafety-related RAB normal ventilation system whose design does not allow it to be credited to limit radiological dose consequences in accordance with RG 1.183. Finally, the inspectors reviewed Calculation EC-S92-001, Revision 1, which was Waterfords evaluation of NRC Information Notice 91-56, Potential Radioactive Leakage to Tank Vented to Atmosphere, but had since been retired. The inspectors observed that valves SI-106A(B), SI-107A(B), SI-120A(B), SI-121A(B), SI-412A(B), SI-417A(B), and CS-118A(B) were all listed as potential leak paths from the containment sump to the RWSP. Since RG 1.183 specifically stated to consider unfiltered leakage through isolation valves into tanks, such as the RWSP, and the Waterford analysis did not consider this source, nor state any exceptions, the inspectors concluded that the leakage through the valves was an unanalyzed and unfiltered release pathway and, consequently, Waterford had provided incomplete and inaccurate information to the NRC.
The inspectors also reviewed the NRCs safety evaluation of LAR 186 (ML030760620)regarding realignment of RWSP boundary isolation valves to the nonsafety-related RWSP purification system. This LAR authorized Waterford to implement a design change where two normally closed manual valves (FS-423 and FS-404) were altered to normally open. In the safety evaluation, the NRC concluded that Waterford would have adequate time to isolate the RWSP either from a pipe crack in the RSWP purification system concurrent with a LOCA or from draw-down by the RWSP purification system aligned to the spent fuel pool (SFP) during a LOCA before reaching the analytical limit of the RWSP. The inspectors also observed that the need to isolate the RWSP from the RWSP purification system is time-critical and the operation is performed locally. Though the RWSP purification system is not part of the ECCS, it is hydraulically connected to the RWSP at the onset of a LOCA, and the inspectors observed that the operator action to close the isolation valves may not occur until after Waterford transitioned to recirculation from the containment sump. Additionally, portions of the purification system are in areas not serviced by the CVAS, and Waterford does not implement a purification system leak test like OP-903-110. The inspectors also observed that EC-S92-001 identified that SI-347 is another normally open valve with potentially similar consequences. Since RG 1.183 specifically stated to consider leakage through valve packing, pump seals, flanged connections, and valves isolating other systems, such as the RWSP purification system, and the Waterford analysis did not consider this source, nor state any exceptions, the inspectors concluded that the leakage through the valves was another unanalyzed and unfiltered release pathway and, consequently, Waterford had provided incomplete and inaccurate information to the NRC.
Corrective Actions: The licensee performed an initial evaluation of the unanalyzed release pathways and determined the additional dose offsite and to the control room would remain below the acceptance criteria in 10 CFR 50.67, "Accident Source Term," for allowable dose. Additionally, the unanalyzed release pathways are serviced by the nonsafety-related RAB normal ventilation system which is filtered but not to the degree that RG 1.183 would allow crediting it for filtration.
Corrective Action References: CR-WF3-2023-18180
Performance Assessment:
Performance Deficiency: The inspectors determined that the failure to verify the adequacy of design of the LOCA AST offsite and control room radiological dose consequence analysis in accordance with 10 CFR Part 50, Appendix B, Criterion III, was a performance deficiency. This performance deficiency resulted in the licensee providing incomplete and inaccurate information to the NRC in accordance with 10 CFR 50.9.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Design Control attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the licensee failed to verify the adequacy of design of the radiological dose consequence analysis when it failed to account for leakage of contaminated water through isolation valves to the refueling water storage pool which vents to an area of the reactor auxiliary building that represents an unfiltered release pathway and additional dose offsite and to the control room.
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 determined the finding was of very low safety significance (Green) because the finding only represented a degradation of the radiological barrier function provided for the control room, auxiliary building, or spent fuel pool.
Cross-Cutting Aspect: Not Present Performance. No cross-cutting aspect was assigned to this finding because the inspectors determined the finding did not reflect present licensee performance.
Enforcement:
The ROPs significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it is necessary to address this violation which impedes the NRCs ability to regulate using traditional enforcement to adequately deter non-compliance.
Severity: This violation was less serious than the Severity Level III violation example described by Section 6.9.c.1 of the NRC Enforcement Policy, which stated that Inaccurate or incomplete information is provided or maintained. If this information had been completely and accurately provided or maintained, it would likely have caused the NRC to reconsider a regulatory position or undertake a substantial further inquiry. Specifically, the inspectors determined, based upon a review of multiple, approved AST LAR submittals for other licensees, which did include leakage through RSWP isolation valves in their LOCA analyses, that the NRC would not likely have caused the NRC to reconsider a regulatory position or undertake a substantial further inquiry had the information been complete and accurate. The Policy did not contain an applicable Severity Level IV violation example. However, the inspectors determined the violation met the Policy definition of Severity Level IV in that it created the potential of a more than minor safety consequence because it resulted in an increase in radiological dose offsite and to the control room.
Violation: Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, states, in part, that measures shall be established for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program.
Contrary to the above, between July 15, 2004, and December 11, 2023, the licensee failed to verify or check the adequacy of design of its LOCA AST offsite and control room radiological dose analysis. Specifically, licensee did not include in its source term, two unfiltered release pathways through isolation valves to the RWSP and through the RWSP purification system when implementing the guidance described in RG 1.183.
10 CFR 50.9, Completeness and Accuracy of Information, states, in part, that information provided to the Commission by a licensee shall be complete and accurate in all material respects.
Contrary to the above, between July 15, 2004, and October 19, 2004, the licensee failed to provide information to the Commission that was complete and accurate in all material respects. Specifically, in the licensing correspondence listed below, the licensee documented that they had performed the radiological dose consequence analysis using the guidance of RG 1.183, they had taken some exceptions to RG 1.183, and that ECCS was assumed to leak 0.5 gpm outside containment but to areas of the RAB serviced by the CVAS. However, the dose consequence analysis did not conform to the guidance in RG 1.183 because it did not include release pathways (i.e., ECCS leakage through RSWP isolation valves and leakage through the RWSP purification system) that should have been analyzed. Though Waterford did take exceptions to RG 1.183, the exceptions were not related to leakage through isolation valves. Additionally, these release pathways involved releases to areas of the RAB not serviced by the CVAS and were, therefore, unfiltered. Because the licensee did not follow the guidance of RG 1.183 by failing include the unfiltered release pathways in their dose analysis or provide an acceptable alternative to excluding the unfiltered release pathways, the information provided to the NRC was incomplete and inaccurate. Finally, this information was material to the NRC because full implementation of the AST revised the licensing basis for Waterfords accident source term and the addition of these unfiltered release pathways increases the analyzed radiological dose offsite and to the control room.
- Letter W3F1-2004-0053, License Amendment Request NPF-38-256, Alternate Source Term, dated July 15, 2004 (ML042020294)
- Letter W3F1-2004-0076, Supplement 2 to Amendment Request NPF-38-256, Alternate Source Term, dated September 1, 2004 (ML042470194)
- Letter W3F1-2004-0095, Supplement 3 to Amendment Request NPF-38-256, Alternate Source Term, dated October 13, 2004 (ML042890193)
- Letter W3F1-2004-0101, Supplement 4 to Amendment Request NPF-38-256, Alternate Source Term, dated October 19, 2004 (ML043010129)
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
- On January 22, 2024, the inspectors presented the Comprehensive Engineering Team Inspection results to Joseph Sullivan and other members of the licensee staff.
- On December 14, 2023, the inspectors presented the exit meeting for the Comprehensive Engineering Team Inspection results to Joseph Sullivan, Site Vice President, and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
3-H
Diesel Generator Room Ventilation
Emergency Feedwater System Head Curves
EC-E91-055
AC Short Circuit Calculations
EC-E91-251
Short Circuit Study For PDP 3B-DC-S & PDP 3B1-DC-S
EC-I95-019
Plant Protection System Indication and Recording Loop
Uncertainty Calculation
EC-M84-001
Tank Volume vs. Level Tables
EC-M92-027
MOV Design Basis Review Calculation No. SI.003
EC-M92-037
MOV Design Basis Review Calculation No. SI.002
EC-M97-026
Required Submergence to Prevent Vortexing in the RWSP
EC-M97-044
Design Basis Review for CCW Isolation Valves CC-200A(B),
CC-501, CC-562, CC-563, and CC-727
EC-M98-008
RWSP Volume Design Requirements
EC-S92-001
NRC IN 91-56: Post-LOCA Releases Through RWSP
EC-S97-020
Toxic Chemical Analysis to Assess Control Room
Habitability
EC90-050
Degraded Voltage Relay Setpoint & Plant Load Study
EC91-056
Relay Settings and Coordination Curves for 6.9kV 4.16kV
and 480 buses
ECE13-001
480 VAC Masterpact NT/NW Micrologic 5.0A Device
Settings and Coordination Study
ECE17-001
EF# Arc Flash Risk Assessment
ECE91-050
Degraded Voltage Relay Setpoint & Plant Load Study
ECE91-194
Load Study For PDP-3B-DC-S and 3B1-DC-S
ECM21-002
Minimum Required Wall Thickness for 132 CW lines and
Selected Tanks
ECS09-005
Air-Operated Valves - Design Basis Accident Times
EE2-11-05
Calculation of Fault Clearing Time of Protective Devices
Used as Back-up Protection of Electrical Penetration
MNQ-10-12
Calculations
MNQ9-2
Component Cooling Water System
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
MNQ9-46
Corrective Action
Documents
Condition Report
(CR-WF3-)
2014-03711, 2015-01904, 2021-00147, 2021-00244, 2023-
15701, 2023-17135, 2022-01335, 2023-15663, 2019-01534,
20-00258, 2021-05371, 2021-05624, 2023-15971, 2022-
03722, 2023-13732, 2023-15260, 2020-06272, 2019-08585,
23-17364, 2022-02962, 2023-00351, 2023-00867, 2023-
290, 2023-15683, 2023-16984,
Corrective Action
Documents
Resulting from
Inspection
Condition Report
(CR-WF3-)
23-17870, 2023-17880, 2023-17881, 2023-17892, 2023-
17910, 2023-17911, 2023-17912, 2023-18092, 2023-18180,
23-18189, 2023-18230, 2023-18042, 2023-17930, 2023-
17926
1564-1630
Containment Spray Pumps Design Curve
74470-772-001
Instrument Nozzles, Waterford III Piping, Component Code
- 91-0118-0001
B289, Sheet 109
Power Distribution & Motor Data, 125 V D-C Distribution
Panel 3B-DC-S
B289, Sheet 21
Power Distribution and Motor Data 480V SWGR
[Switchgear] 3B31-S One Line Diagram
B289, Sheet 21-1
Power Distribution and Motor Data, 480V SWGR
[Switchgear] 3B31-S One Line Diagram
B289, Sheet
21A1
Power Distribution & Motor Data, 480V Switchgear 3B31-S
Programmer Settings
B424
Control Wire Diagram: Component Cooling Water Isolation
and Makeup
ESSE-BA-IC144
Main Control Room Emergency Breathing Air Piping
F42932
16" Type 9220 Valve Assembly w/Bettis 732C-SR80-M3
Actuator
D
G160
Flow Diagram: Component Closed Cooling Water System
G163
Flow Diagram Containment Spray & Refueling Water
Storage Pool
G167
Flow Diagram Safety Injection System
G195
Piping Diagram Safety Injection System
G287, Sheet 1
25 VDC and 120 VAC One Line Diagram
Drawings
H33760-1201
Sensor, Temperature
C
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Component Cooling Water Surge Tank
09/23/1977
MOV Design Basis Review Calculation NO SI.003
Evaluation of Impacted Safety Related Calculations
Broad Range Gas Monitor (BRGM) System Replacement
Replace SI-512A with a Swing Check Valve
MSIV Solenoid Valve Upgrades
Planned FCR for EC-83913 - MSIV Solenoid Valve Upgrade
- Incorporate Vendor Supplied Documentation
AMTEK Solid State Controls Component Substitution
Justification
000
Update to Consider Normal Shutdown with LOOP as
Bounding CCW Pump Runout Case
Stroke Time Evaluation of Various SI, CVR, and SBV Valves
Temperature Indication Switch Standard EN-IC-S-025-W
000
Clarification of Seismic Design Requirements for Piping and
Valves Upstream of EBA Tank Isolation Valves (EBA-203,
EBA-204, 4BA1/2-5)
MOV Design Basis Review Calculation No. SI.003
Evaluation of EC-M89-002 to Increase Supply Pressure for
N2 Accumulators III and IV
Minimum Required Wall Thickness for 132" CW Lines and
Selected Tanks
Evaluate Use of Alloy 690 Thermowells for Cold Leg RTDs
Generate Air Operated Valve Design Basis Calculation for
PMU-144
FHD-232A Leak Repair Evaluation
ECE91-053 12IAV55C Acceptance Range Change
Temporary Load Bank for Integrated Diesel Test A Train
Turbine Control System
RCS Foreign Material Evaluation - RCITE0112CD
Thermowell and RTD CR-WF3-2022-3722 - Evaluation for
Continued Operation
Engineering
Changes
ER-W3-99-0184-
01-02
Weld Repair of Inconel Instrument Nozzles on the Hot Legs
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Engineering
Evaluations
22-01
Substitute CPC D RTD Input RC ITE0112CD1 with RC
ITE0115-1
457002467
Paul Munroe Enertech MSIV Pump Valve
LOU-1564.108
EBASCO Purchase Specification for Main Steam Isolation
Valves
08/02/1984
LOU-1564.724
EBASCO Purchase Specification for Miscellaneous Shop
Fabricated Tanks
09/14/1984
SEP-WF3-IST-1
WF3 Inservice Testing Bases Document
SEP-WF3-IST-2
WF3 Inservice Testing Plan
SEP-WF3-IST-3
WF3 Inservice Testing Cross Reference Document
TCABP
Time Critical Action Program Bases
TD-B015.0045
Babcock and Wilcox CCW and ACCW Pump Model
2X14X16 DE Manual
TD-B237.0195
GH Bettis Service Instructions, Disassembly, & Reassembly
for Spring Return Series Pneumatic Actuators
TD-C173.0015
C & D AutoReg Charger Handbook, 3 Phase, 6 Pulse SCR
Chargers
TD-C173.0025
C & D Standby Battery Vented Cell Installation & Operating
Instructions, RS-1476, Section 12-800
TD-F130.0015
Fisher Controls Valve Bodies
TD-F130.0015
Fisher Controls Valves Bodies
TD-G080.0155
General Electric Low Voltage Power Circuit Breakers,
Various Instructions
TD-L200.0045
Limitorque Type SMB Instruction & Maintenance Manual
Bulletin SMBI-82D
TD-S250.0025
Solid State Controls Ferroresonant UPS Technical Manual
Number 95-006000-90 Rev. B 9/91
TD-W120.3095
Westinghouse Molded Case Circuit Breakers Series C, K-
Frame, For Type DK, KDB, KD, HKD, KDC, KW, HKW,
KWC
W3-DBD-001
Safety Injection System
305
W3-DBD-003
W3-DBD-008
Electrical Distribution (DC Portion) Design Basis Document
301
Miscellaneous
W3-DBD-011
Electrical Distribution (AC Portion) Design Basis Document
2
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
W3-DBD-14
Safety Related, Air Operated Valves
304
W3-DBD-4
Component Cooling Water, Auxiliary Component Cooling
Water
307
EN-DC-115-01
Industry Standard Design Process (IP-ENG-001)
Motor Operated Valve (MOV) Test Data Review
Process Applicability Determination
Reactivity Management Program
Conduct of Operations
Operator Fundamentals Program
Time Critical Action Program Standard
ME-003-330
480 Volt G.E. Switchgear Breakers
318
OP-002-003
Component Cooling Water
2
OP-002-006
Fuel Pool Cooling and Purification
331
OP-010-003
Plant Startup
365
OP-010-005
Plant Shutdown
347
OP-0101-007
Plant Cooldown
007
OP-901-212
Rapid Plant Power Reduction
24
OP-901-310
Loss of Train A Safety Bus
314
OP-901-311
Loss of Train B Safety Bus
2
OP-901-510
Component Cooling Water System Malfunction
306
OP-901-511
Instrument Air Malfunction
OP-902-000
Standard Post Trip Actions
OP-902-005
Station Blackout Recovery Procedure
OP-902-008
Functional Recovery Procedure
OP-902-009
Emergency Operating Procedure Standard Appendices
23
OP-903-001
Technical Specification Surveillance Logs
105
OP-903-014
Emergency Feedwater Flow Verification
318
OP-903-046
Emergency Feed Pump Operability Check
25
OP-903-118
Primary Auxiliaries Quarterly IST Valve Tests
OP-903-121
Safety System Quarterly IST Valve Tests
OP-903-129
Component Cooling Water Makeup Pump Operability Check
Procedures
OP-904-004
Turbine Miscellaneous Test and Checks
2
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
UNT-006-033
Technical Specifications Surveillance Frequency List
Work Orders
Work Order (WO-
)
00019593, 00019634, 00019649, 52873452, 52944019,
2960918, 52960918, 53034122, 00495839, 00167498,
00514233, 00565026, 00026241, 00026243, 00298246,
00050249, 00563831, 00563832, 51794244, 52984649,
2333000, 52364836, 52487650, 52504349, 52544989,
2575173, 52581739, 52609098, 52645886, 52667983,
2675538, 52689470, 52691659, 52778435, 52881422,
2832648, 52944180, 52962382, 52962657, 54008005,
00504476, 00504477, 00527322, 53033253, 52958193,
2963037, 53002842, 00536432, 54032998, 53036041,
2947469, 52968126, 54003050