Control of Preheat Temperature for Welding of Low-Alloy Steel
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Regulatory Guide 1.50# QUERY63d5ea34e6aa3a2e0bea4e8bf23559a9
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RS-17-126, Quad Cities Nuclear Power Station, Units 1 & 2, Revision 14 to Updated Final Safety Analysis Report, Chapter 5, Reactor Coolant System and Connected Systems
LR-N17-0088, Hope Creek Generating Station Updated Final Safety Analysis Report, Rev. 22, Chapter 5, Reactor Coolant System and Connected Systems
LR-N17-0088, Hope Creek Generating Station Updated Final Safety Analysis Report, Rev. 22, Chapter 10, Steam and Power Conversion System
LR-N17-0088, Hope Creek Generating Station Updated Final Safety Analysis Report, Rev. 22, Chapter 1, Introduction and General Description of Plant
NL-17-0534, Joseph M. Farley Nuclear Plant, Updated Final Safety Analysis Report, Revision 27, Chapter 3, Part 1
LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Appendix 3A, Pse&G Positions on Us NRC Regulatory Guides
NL-16-2280, Vogtle Electric Generating Plant, Units 1 & 2, Updated Final Safety Analysis Report, Chapter 1, Introduction and General Description of Plant
RS-16-196, Clinton - Updated Safety Analysis Report (Usar), Revision 18, Chapter 5 - Reactor Coolant System and Connected Systems
WM 16-0008, Wolf Creek Generating Station, Redacted Version of Revision 29 to Updated Safety Analysis Report, Chapter 10.0 - Steam and Power Conversion System
WM 16-0008, Wolf Creek Generating Station, Redacted Version of Revision 29 to Updated Safety Analysis Report, Chapter 6.0 - Engineered Safety Features
RC-16-0083, Virgil C. Summer, Unit 1 - Updated Final Safety Analysis Report, Chapter 10, Steam and Power Conversion System
NL-15-183, Watts Bar Nuclear Plant, Unit 2, Amendment 114 to Final Safety Analysis Report, Section 10.0 - Main Steam and Power Conversion System
RS-14-006, Clinton, Unit 1, Updated Safety Analysis Report, Revision 16, Chapter 5 - Reactor Coolant System and Connected Systems
RS-14-006, Clinton, Unit 1, Updated Safety Analysis Report, Revision 16, Chapter 1 - Introduction and General Description of Plant
RS-13-009, Clinton, Unit 1, Updated Safety Analysis Report, Revision 15, Chapter 5 - Reactor Coolant System and Connected Systems
RS-11-159, Quad Cities, Units 1 and 2 - Updated Final Safety Analysis Report (Ufsar), Revision 11, Chapter 05 - Reactor Coolant System and Connected Systems
RS-11-100, Dresden, Units 2 and 3, Updated Final Safety Analysis Report (Ufsar), Revision 9, Chapter 5.0, Reactor Coolant System and Connected Systems
RS-08-045, LaSalle, Units 1 and 2, Updated Final Safety Analysis Report (Ufsar), Revision 17, Appendix a Through Appendix N
RC-18-0080, Virgil C. Summer Nuclear Station, Unit 1, Updated Final Safety Analysis Report, Chapter 10.0, Steam and Power Conversion System
RC-18-0080, Virgil C. Summer Nuclear Station, Unit 1, Updated Final Safety Analysis Report, Chapter 3.0, Design of Structures, Components, Equipment and Systems
M200041, GE-Hitachi ABWR Design Control Document Tier 1 & 2, Rev. 7 - Chapter 05 - Reactor Coolant System and Connected Systems
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
ST-HL-AE-2093, Forwards Revised FSAR Section 3.12, Conformance W/Nrc Reg Guides. Section Updated to Reflect Util Position on All Applicable Reg Guides
RBG-21536, Forwards Rev to FSAR & Tech Specs Resulting from Final Review of Fracture Toughness & Reactor Vessel Matl Surveillance Program Requirements Per 10CFR50.Revs Will Be Included in Future Amend to FSAR
ET-NRC-93-3795, Forwards Responses to NRC RAIs on AP600 from Ltrs 920923, 1001 & 09.Remainder of RAIs Contained in NRC Ltrs of 920923 & 1001 Will Be Provided Prior to 930123
RBG-16-672, Forwards Response to Draft SER Open Items Identified by NRC Mechanical Engineering Branch & Response to NRC Requesting Further Clarification on Design of Pipe/Component Supports
LD-93-011, Forwards Responses to 22 Issues Re Closure of Sys 80+ Draft SER Issues
LD-92-123, Forwards Section 17.3, Design Reliability Assurance Program Plan for Sys 80+ Standard Design, Rev 1 & Responses to 79 Issues Identified in Dser for Sys 80+
LD-96-029, Forwards Finalized Package of Six Draft Changes to Sys 80+ Std Plant Design Control Document.Changes Will Be Formally Incorporated Into Sys 80+ Design Control Document at Conclusion of Design Certification Rulemaking
LD-96-028, Submits Addl Draft Changes for Staff Review & Approval. Changes Listed on Attached Table W/Details Shown on Encl Design Control Document Pages
LD-92-063, Forwards Info to Supplement 920325 Responses to NRC Request for Addl Info Re Analysis of Ruptured Reactor Coolant Pump Seal Cooler Tube & Single Failure Assumptions for Analysis of moderate-energy-line Break
LD-95-013, Transmits Supplementary Revs for Incorporation Into DCD & Affidavit.Changes Resolve Inconsistencies Identified by NRC & by ABB-CE
LD-91-010, Responds to NRC 881223 Request for Addl Info Re CESSAR-DC. Forwards Proposed Revisions to CESSAR-DC
LD-90-008, Forwards Response to 881216 Request for Addl Info Re CESSAR-DC,Chapters 3,4,5 & 6 Re Turbine Missiles,Control Element Drive Structural Matls,Cleaning & Contamination Protection Procedures & Reactor Internals Matls
RAIO-0818-61388, LLC Response to NRC Request for Additional Information No. 492 (Erai No. 9551) on the NuScale Design Certification Application
RBG-16-413, Forwards Response to Draft SER Open Item Re Reactor Vessel Matls Surveillance Program.Response Will Be Incorporated Into Future FSAR Amend