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INTERNAL OR EXTERNAL MEMORANDUM  and MEMORANDUMS-CORRESPONDENCE

Boron +, Lead +  and Tin +

10 CFR 55 +, 10 CFR 50 Appendix A +, 10 CFR 50.46 +, 10 CFR 50 +, 10 CFR 50.62 +, 10 CFR 55.45 +, 10 CFR 50.34#f +, 10 CFR 50.62#c1 +, 10 CFR 50.34#r20 +, 10 CFR 52.47#b2 +  and 10 CFR 52.47#a1iv +

SECY-83-472 +  and SECY-94-084 +

Turbine Trip +  and Reactor Trip +

05000482/FSAR-15.4#6 +, Category:05000482/FSAR Chapter 15 +, 05000482/FSAR-7.4#1 +, 05000482/FSAR-3.9#4 +, 05000482/FSAR-9.1 +, 05000482/FSAR-1.9#4.2 +  and 05000482/FSAR-1.9#4.1 +

NRC (Affiliation Not Assigned) +  and Collins T +

GDC 34 +  and GDC 30 +

NRC Generic Letter 80-19, Resolution of Enhanced Fission Gas Release Concern +, NRC Generic Letter 80-14, LWR Primary Coolant System Pressure Isolation Valves +, NRC Generic Letter 80-01, NUREG-0630 "Cladding, Swelling and Rupture - Models for LOCA Analysis" +, NRC Generic Letter 81-21, Natural Circulation Cooldown +, NRC Generic Letter 83-22, Safety Evaluation of "Emergency Response Guidelines" +, NRC Generic Letter 83-11, Licensee Qualification for Performing Safety Analyses in Support of Licensing Actions +, NRC Generic Letter 84-21, Long Term Low Power Operation in Pressurized Water Reactors +, NRC Generic Letter 85-16, High Boron Concentrations +, NRC Generic Letter 85-05, Inadvertent Boron Dilution Events +, NRC Generic Letter 86-16, Westinghouse ECCS Evaluation Models +, NRC Generic Letter 86-07, Transmittal of NUREG-1190 Regarding the San Onofre Unit 1 Loss of Power and Water Hammer Event +, NRC Generic Letter 87-12, Loss of Residual Heat Removal While the Reactor Coolant System is Partially Filled +, NRC Generic Letter 88-17, Loss of Decay Heat Removal +, NRC Generic Letter 89-18, Resolution of Unresolved Safety Issues A-17, "Systems Interactions in Nuclear Power Plants" +, NRC Generic Letter 91-07, GI-23, "Reactor Coolant Pump Seal Failures" and Its Possible Effect on Station Blackout +  and NRC Generic Letter 93-04, Rod Control System Failure and Withdrawal of Rod Control Cluster Assemblies, 10 CFR 50.54(f) +

ML20248L701 +, ML20248L701 +, ML20248L701 +, ML20248L701 +, ML20248L701 +, ML20248L701 +, ML20248L701 +, ML20248L701 +, ML20248L701 +, ML20248L701 +, ML20248L701 +  and ML20248L701 +

NUREG/CR-5374 +, NUREG-0471, Responds to 781113 Request & Encl Generic Task Problem Descriptions(NUREG-0471)& NRC Prog for Resolution of Generic Issues Related to Nuc Pwr Plants(NUREG-0410). W/O Encl +, NUREG-1190, Transcript of Commission 860122 Briefing in Washington,Dc Re San Onofre & Status of Rancho Seco.Pp 1-105.Supporting Documentation & NUREG-1190, Loss of Power & Water Hammer Event at San Onofre,Unit 1 on 851121 Encl +, NUREG-1229, Forwards Revised Resolution of USI A-17, Sys Interactions in Nuclear Power Plants. Package Requested to Be Revised to State More Explicitly Proposed Actions Taken by Licensees & by Nrc.Issue Identified as Category 2 Action +, NUREG-0737, To NUREG-0737, Request for Extension of Time for Commission Order Dated June 14, 1984 +, 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 +  and NUREG-0630, 09-09-80 Cladding Swelling and Rupture Models for LOCA Analysis - NUREG-0630 +

Regulatory Guide 1.97 +, Regulatory Guide 1.45 +, Regulatory Guide 1.133 +  and Regulatory Guide 1.139 +

SECY-83-472 +  and SECY-94-084 +

Steam Generator +, Reactor Coolant System +, Feedwater +, Residual Heat Removal +, Emergency Core Cooling System +, Decay Heat Removal +  and Control Rod +

NRC (Affiliation Not Assigned) +  and Quay T +

Safe Shutdown +, Boric Acid +, Shutdown Margin +, Reactor Vessel Water Level +, Probabilistic Risk Assessment +, Loss of Offsite Power +, Pressure Boundary Leakage +, Core Exit Thermocouple +, Loss of condenser vacuum +  and Water hammer +

WCAP-14477 +, WCAP-1447 +  and WCAP-019 +

WCAP-14477 +, WCAP-1447 +  and WCAP-019 +