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Letter, Report  and Miscellaneous

Boron +  and Lead +

10 CFR 50.54#f +  and 10 CFR 50.54 +

05000219  +  and 00219  +

Turbine Trip +  and Reactor Trip +

J. J. Barton +  and GPU Nuclear Corp +

NRC Generic Letter 89-19, Request for Actions Related to Resolution of Unresolved Safety Issue A-47 "Safety Implication of Control Systems in LWR Nuclear Power Plants" Pursuant to 10 CFR 50.54(f) +  and NRC Generic Letter 88-20, Individual Plant Examination for Severe Accident Vulnerabilities +

ML060550285 +, ML060550285 +, ML060550285 +, ML060550285 +, ML060550285 +, ML060550285 +, ML060550285 +, ML060550285 +, ML060550285 +, ML060550285 +, ML060550285 +, ML060550285 +, ML060550285 +  and ML060550285 +

2.314815e-5 d (5.555556e-4 hours, 3.306878e-6 weeks, 7.61e-7 months)  +, 1.157407e-4 d (0.00278 hours, 1.653439e-5 weeks, 3.805e-6 months)  +, 4.62963e-5 d (0.00111 hours, 6.613757e-6 weeks, 1.522e-6 months)  +, 8.101852e-5 d (0.00194 hours, 1.157407e-5 weeks, 2.6635e-6 months)  +, 1.157407e-5 d (2.777778e-4 hours, 1.653439e-6 weeks, 3.805e-7 months)  +, 3.472222e-5 d (8.333333e-4 hours, 4.960317e-6 weeks, 1.1415e-6 months)  +, 2.546296e-4 d (0.00611 hours, 3.637566e-5 weeks, 8.371e-6 months)  +  and 2.777778e-4 d (0.00667 hours, 3.968254e-5 weeks, 9.132e-6 months)  +

NUREG-150 +, NUREG/CR-4448 +, NUREG-1174, Summary of CRGR Meeting 163 on 890524 Re Proposed Generic Ltr on Emergency Response Data Sys,Usi A-17 on Sys Interactions,Generic Issue A-30 on safety-related Dc Power Sys & USI A-47 on Control Sys +, NUREG-1335, Staff Evaluation Rept Concluding That Licensee IPE Complete Wrt Info Requested by GL 88-20 & Associated Guidance, NUREG-1335 +, NUREG-1217, Forwards NRC Resolution of USI A-47, Safety Implications of Control Sys, for Info.Resolution Consists of NUREG-1217, Evaluation of Safety Implications...Technical Findings Re USI A-47 & NUREG-1218, Regulatory Analysis.... W/O Encls +, NUREG-1218, Disapproves 900320 Response to Generic Ltr 89-19 Re Safety Implications of Control Sys in LWRs +  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 +

ST +, RT +, CP +, Emergency Operating Procedure +  and ES-20 +

Manual Scram +  and Automatic Scram +

Ohio +, Washington +  and New Jersey +

Reactor Coolant System +, Feedwater +, Secondary containment +, Service water +, Reactor Protection System +, Primary Containment Isolation System +, Primary containment +, Reactor Recirculation Pump +, Reactor Pressure Vessel +, Core Spray +, Reactor Water Cleanup +, Standby Gas Treatment System +, Automatic Depressurization System +, Standby Liquid Control +, Circulating Water System +, Main Steam Safety Valve +, Decay Heat Removal +, Main Steam Line +, Main Condenser +, Control Rod +, Main Steam +  and Containment Spray +

Office of Nuclear Reactor Regulation +  and Document Control Desk +

Reactor Vessel Water Level +, Mission time +, Internal Flooding +, Probabilistic Risk Assessment +, B.5.b +, Scram Discharge Volume +, Loss of Offsite Power +, Hardened vent +, Fuel cladding +  and Loss of condenser vacuum +