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ANSI 13.1-1969 +  and ANSI N13.1-1969 +

ASME Section I +  and ASME Section II +

Strontium +, Iodine +, Nitrogen +, Oxygen +, Hydrogen +, Xenon +, Lead +, Silver +, Tellurium +, Barium +  and Cerium +

10 CFR 55 +, 10 CFR 50 Appendix A +, 10 CFR 26 +, 10 CFR 50.46 +, 10 CFR 100 +, 10 CFR 50 +  and 10 CFR 73.2 +

PLA-1621, Forwards Response to Generic Ltr 82-33,Suppl 1 to NUREG-0737, Requirements for Emergency Response Capability. W/One Oversize Encl.Aperture Card Is Available in PDR +, SECY-81-513 +, PLA-1312 +  and PLA-8081, 1 to Updated Final Safety Analysis Report, Questions and Responses 121.1 Through 121.21 +

05000387  +  and 05000388  +

Reactor Trip +  and Turbine Trip +

GDC 19, Control room +  and GDC 54 +

NRC Generic Letter 82-33, Supplement 1 to NUREG-0737 - Emergency Response Capability +  and NRC Generic Letter 81-10, Post-TMI Requirements for the Emergency Operations Facility +

ML23292A186 +, ML23292A186 +, ML23292A186 +, ML23292A186 +, ML23292A186 +, ML23292A186 +, ML23292A186 +, ML23292A186 +, ML23292A186 +, ML23292A186 +, ML23292A186 +, ML23292A186 +, ML23292A186 +, ML23292A186 +, ML23292A186 +, ML23292A186 +  and ML23292A186 +

1.851852e-4 d (0.00444 hours, 2.645503e-5 weeks, 6.088e-6 months)  +, 2.777778e-4 d (0.00667 hours, 3.968254e-5 weeks, 9.132e-6 months)  +, 8.333333e-4 d (0.02 hours, 1.190476e-4 weeks, 2.7396e-5 months)  +, 1.041667e-4 d (0.0025 hours, 1.488095e-5 weeks, 3.4245e-6 months)  +, 8.101852e-5 d (0.00194 hours, 1.157407e-5 weeks, 2.6635e-6 months)  +  and 2.314815e-5 d (5.555556e-4 hours, 3.306878e-6 weeks, 7.61e-7 months)  +

IEEE 344-1975 +  and IEEE 323-1974 +

Tritium +, Am-241 +, Co-60 +, Cs-137 +, Sr-90 +, Y-90 +, Kr-85 +, I-131 +, Mo-99 +, Pu-238 +  and Pu-239 +

ML17138A954 +, ML17138A923 +  and ML17138B433 +

General Electric +  and Target Rock +

NUREG-700 +, NUREG-0635, Forwards SER Re TMI Item II.K.3.30 for C-E Plants,Per Conditional Acceptance of CEN-203(P),Rev 1, Response to NRC Action Plan Item II.K.3.30 Justification of Small Break LOCA Methods +, NUREG-0694, Advises That Augmented Natural Circulation Training Should Be Performed.Descriptions of Natural Circulation Tests Should Be Included in Fsar.Detailed Test Procedures Must Be Submitted 60 Days Prior to Scheduled Test Performance +, NUREG-0801, Forwards Comments on Program Plan for Detailed Control Room Design Review (Dcrdr).Program Plan Addresses All Required Elements of DCRDR.In-progress Audit Not Planned +, NUREG-0696, Forwards SER Accepting Proposed Alternate Emergency Operations Facility Location +, NUREG-0654, Correction to Amendment Nos. 221 and 206 to Authorize Revision of the Emergency Plan Based on NUREG-0654/FEMA-REP-1, Revision 2 +, NUREG-0588, Responds to 870709 Request for NRR Assistance in Determining Adequacy of Two Test Repts Re Acceptance Criteria of NUREG-0588,Category I.Qualification of Kulka Terminal Blocks Demonstrated & Not Used in Instrumentation Circuits +, NUREG-0585, Forwards NUREG-0578, TMI-2 Lessons Learned Task Force Status Report & Short-Term Requirements, & NUREG-0585, TMI-2 Lessons Learned Task Force Final Rept, in Response to .W/O Encl +, NUREG-0737, To NUREG-0737, Request for Extension of Time for Commission Order Dated June 14, 1984 +  and NUREG/CR-1580, Notice of Availability of Draft NUREG/CR-1580, Human Engineering Guide to Control Room Evaluation +

EP-AD-013 +  and TOP +

Regulatory Guide 1.183 +, Regulatory Guide 1.97 +, Regulatory Guide 1.78 +, Regulatory Guide 1.3 +, Regulatory Guide 1.33 +, Regulatory Guide 1.70 +, Regulatory Guide 1.7 +, Regulatory Guide 1.23 +, Regulatory Guide 1.89 +  and Regulatory Guide 1.96 +

NEDO-24708A +, NEDO-24934 +  and NEDO-24782 +

141F013A-S +  and 07-79-01B +

Manual Scram +  and Automatic Scram +

Tennessee +, California +  and Alabama +

Reactor Coolant System +, Feedwater +, Secondary containment +, Service water +, High Pressure Coolant Injection +, Emergency Diesel Generator +, Main Steam Isolation Valve +, Auxiliary Feedwater +, Reactor Core Isolation Cooling +, Primary containment +, Shutdown Cooling +, Reactor Recirculation Pump +, Safety Parameter Display System +, Reactor Pressure Vessel +, Core Spray +, Reactor Building Ventilation +, Residual Heat Removal +, Standby Gas Treatment System +, Automatic Depressurization System +, Standby Liquid Control +, Intermediate Range Monitor +, Emergency Core Cooling System +, Decay Heat Removal +, Main Steam Line +, Safety Relief Valve +, Main Condenser +, Control Rod +, Main Steam +, Containment Spray +...

TS-3.7.4 +, TS-3.7.3 +, TS-5.5.2 +  and TS-5.2.2.e +

Reactor Vessel Water Level +, Grab sample +, Anticipated operational occurrence +, Scram Discharge Volume +, Shift Technical Advisor +, Loss of Offsite Power +, Excess Flow Check Valve +, Earthquake +, Fuel cladding +, Thermal fatigue +, Potassium iodide +  and Control Room Habitability +