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INTERNAL OR EXTERNAL MEMORANDUM  and MEMORANDUMS-CORRESPONDENCE

Boron +, Hydrogen +  and Lead +

10 CFR 20 +, 10 CFR 50 +, 10 CFR 52 +, 10 CFR 73 +, 10 CFR 50.12 +, 10 CFR 50.62 +, 10 CFR 52 Appendix B, Appendix B to Part 52-Design Certification Rule for the System 80+ Design +, 10 CFR 52 Appendix C, Appendix C to Part 52-Design Certification Rule for the Ap600 Design +, 10 CFR 73.55 +, 10 CFR 100 Appendix A +, 10 CFR 50.34#f +, 10 CFR 52.47#a2 +, 10 CFR 73.55#d +  and 10 CFR 73.55#c +

Emergency Preparedness +, Initiating Events +  and Physical Protection +

L-036 +, LD-89-088 +, LD-89-033, Forwards Design Certification Licensing Review Basis, for Review & Concurrence +  and LD-88-008, Forwards Proprietary Base Line Level 1 PRA for Sys 80r NSSS Design. Rept Presents Methodology & Results of Sys 80 PRA to Be Used for Evaluation of Design trade-offs for Sys 80+ Std Design +

51 FR 30028 +, 51 FR 28044 +, 54 FR 15372 +  and 50 FR 37138 +

Holahan G +  and Office of Nuclear Reactor Regulation +

ML20246E292 +, ML20246E292 +, ML20246E292 +, ML20246E292 +, ML20246E292 +, ML20246E292 +, ML20246E292 +, ML20246E292 +, ML20246E292 +, ML20246E292 +, ML20246E292 +  and ML20246E292 +

LD-89-088 +, LD-89-033, Forwards Design Certification Licensing Review Basis, for Review & Concurrence +  and LD-88-008, Forwards Proprietary Base Line Level 1 PRA for Sys 80r NSSS Design. Rept Presents Methodology & Results of Sys 80 PRA to Be Used for Evaluation of Design trade-offs for Sys 80+ Std Design +

NUREG/CR-4550 +, NUREG/CR-2300 +, NUREG/CR-2585 +, NUREG-1197 +, NUREG-0852, Sser Supporting Vendor Responses to Confirmatory Issue 2, Steam Generator Tube Rupture. Calculated Radiological Consequences of Postulated Steam Generator Tube Rupture Accident Meets 10CFR100.11 Dose Ref Values +, NUREG/CR-2815, Discusses PRA Guidance Documents NUREG/CR-2815 & NUREG/CR-3485.Recasting of Documents in SRP or Reg Guide Formats Recommended Upon Receipt of Public Comments If Documents Reflect Agency Views +, NUREG-0800, Document for February 22, 2024 ACRS Digital I&C Subcommittee Meeting +  and NUREG-1050, Forwards Regulatory Analysis of Generic Safety Issue 61, Safety/Relief Valve Discharge Line Break within Mark I or II Wetwell Airspace. Based on Results Discussed in NUREG/CR-4594,risk Lower than Previously Estimated +

ES +, ST +, SAP +  and ISP +

Steam Generator +, Reactor Coolant System +, Feedwater +, Secondary containment +, Service water +, Emergency Diesel Generator +, Main Transformer +, Main Turbine +, Shutdown Cooling +, Decay Heat Removal +, Main Condenser +, Control Rod +, Main Steam +, Containment Spray +  and Emergency Feedwater System +

Congel F +, Thadani A +, Richardson J +  and Office of Nuclear Reactor Regulation +

Safe Shutdown +, Ultimate heat sink +, Mission time +, Probabilistic Risk Assessment +, Earthquake +, Control of Heavy Loads +, Weak link +, Water hammer +  and Leak-Before-Break +