RS-15-129, Westinghouse Report CCE-15-27, Revision 1, Braidwood Units 1 and 2 - Responses to NRC Request for Additional Information (Rai)Regarding Ultimate Heat Sink Temperature Increase License Amendment Request, April 2015
HNP-11-001, Request for License Amendment, Measurement Uncertainty Recapture Power Uprate
L-2011-084, Response to NRC Request for Additional Information Regarding Extended Power Uprate License Amendment Request No. 205 and Containment and Ventilation Issues
W3F1-2018-0031, Supplemental Information Supporting the License Amendment Request Regarding Use of the Tranflow Code for Determining the Pressure Drops Across the Steam Generator Secondary Side Internal Components
DCL-17-038, Diablo Canyon Power Plant, Units 1 & 2, Revised Updated Final Safety Analysis Report, Rev. 23, Chapter 6, Engineered Safety Features (Redacted)
NL-17-0534, Joseph M. Farley Nuclear Plant, Updated Final Safety Analysis Report, Revision 27, Effective Page Listing, Table of Contents, Chapter 1 Redacted, Introduction and General Description of the Plant
LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Section 15.4, Condition Iv - Limiting Faults
NL-16-2280, Vogtle Electric Generating Plant, Units 1 & 2, Updated Final Safety Analysis Report, Figures 3.7.B.2-27 Through 3D-17
NL-16-2280, Vogtle Electric Generating Plant, Units 1 & 2, Updated Final Safety Analysis Report, Chapter 1, Introduction and General Description of Plant
WM 16-0008, Wolf Creek Generating Station, Redacted Version of Revision 29 to Updated Safety Analysis Report, Chapter 6.0 - Engineered Safety Features
DCL-11-072, LAR-11-006 - Revision to Technical Specification 3.3.5. Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation
NRC 2010-0065, Response to Request for Additional Information on License Amendment Request 261, Extended Power Uprate
NRC 2016-0034, Point Beach Nuclear Plant, Units 1 and 2, Periodic Update of the Final Safety Analysis Report (Fsar), Chapter 14 - Safety Analysis
LR-N07-0030, License Change Request S06-010, Steam Generator Feedwater Pump Trip, Feedwater Isolation Valve Response Time Testing and Containment Cooling System
L-07-017, License Amendment Requests Nos. 334 and 205
L-05-006, Response to Request for Additional Information in Support of License Amendment Requests Nos. 317 and 190
ULNRC-05117, Technical Specification Revisions Associated with the Steam Generator Replacement Project
L-85-016, Application to Amend License DPR-23,revising Tech Specs to Delete Requirement Re Maint of High Borated Water Inventory in Boron Injection Tank,Per Generic Ltr 85-16.Proprietary XN-NF-85-17(P) Encl.Rept Withheld (Ref 10CFR2.790).Fee Paid
ULNRC-04928, Responses to RAIs on the Proposed Revision to Technical Specification (TS) 1.1, Definitions, TS 3.7.3, Main Feedwater Isolation Valves (Mfivs), and Steam Generator Tube Rupture with Overfill Re-Analysis
L-PI-04-017, Development and Qualification of a GOTHIC Containment Evaluation Model for the Prairie Island Nuclear Generating Plants
L-03-127, Containment Conversion Pre-application Report
NRC 2002-0004, License Amendment Request 223 Re Containment Pressure
IA-98-351, Forwards Rept Re MSLB with Superheat Release.Rept Outlines NRR Actions & Conclusions to Date & May Be Useful in Auditing 10CFR50.49 Activities.Issue Has Received Significant NRC & Industry Attention
ULNRC-03916, Forwards Util 120-day Response to NRC GL 98-04, Potential for Degradation of ECCS & CCS After LOCA Because of Const & Protective Coating Deficiencies & Foreign Matl in Containment. Supporting Info,Encl
NL-21-1015, Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specification 3.7.2 Limiting Condition for Operation to Remove One Main Steam Isolation Valve System
ST-HL-AE-1779, Forwards Annotated FSAR Revs Re MSIV Closure Logic to Enable Resolution of SER Confirmatory Item 17.Related Correspondence
ULNRC-03425, Provides Addl Info to Support Plant Amend Application & Mod That Will Replace Existing Fixed Logic Sys W/Programmable Digital Processor Equipment
NRC-85-3008, Discusses Steam Superheating During Steam Line Break for Dry Containments.Reanalysis Not Required for Mass & Energy Releases Inside Large Dry & Subatmospheric Containments. Encls Withheld (Ref 10CFR2.790)
NSD-NRC-97-4947, Submits Refs & Changes for SSAR Chapter 7.Encl 1 Will Be Included in SSAR Rev 11.NRC Requested to Review Encl 2 & Provide Westinghouse W/Nrc Status for Item,Including 1052 & 1053 for Interlocks
AEP-NRC-2022-62, 1 to Updated Final Safety Analysis Report, Chapter 14.3, Reactor Coolant System Pipe Rupture (Loss of Coolant Accident) (Unit 1)
05000272/LER-1995-011, :on 880222,inconsistency Between WCAP-11634 Analysis Used for Postulated Steam Line Breaks Outside Containment & Updated FSAR Was Discovered Due to Inadequate Design Review
ULNRC-05206, License Amendment Request OL 1258 Revision to Technical Specification 3.7.3, Main Feedwater Isolation Valves (Mfivs)
NRC 2002-0080, Response to Request for Additional Information License Amendment Request 223 Technical Specification LCO 3.6.4, Containment Pressure