Browse wiki

Pages with broken file links, TEXT-SAFETY REPORT  and NRC TECHNICAL REPORT

Oxygen +, Hydrogen +, Zirconium +  and Lead +

10 CFR 50 +  and 10 CFR 50 Appendix R +

Initiating Events +  and Mitigating Systems +

05000295  +  and 05000304  +

Turbine Trip +  and Reactor Trip +

ML20134N832 +, ML20134N832 +, ML20134N832 +, ML20134N832 +, ML20134N832 +, ML20134N832 +, ML20134N832 +, ML20134N832 +, ML20134N832 +  and ML20134N832 +

4.62963e-5 d (0.00111 hours, 6.613757e-6 weeks, 1.522e-6 months)  +, 9.259259e-5 d (0.00222 hours, 1.322751e-5 weeks, 3.044e-6 months)  +, 2.777778e-4 d (0.00667 hours, 3.968254e-5 weeks, 9.132e-6 months)  +, 1.388889e-4 d (0.00333 hours, 1.984127e-5 weeks, 4.566e-6 months)  +  and 1.157407e-5 d (2.777778e-4 hours, 1.653439e-6 weeks, 3.805e-7 months)  +

NUREG/CR-3226 +, NUREG/CR-2515 +, NUREG/CR-3511 +, NUREG/CR-3428 +, NUREG/CR-2666 +, NUREG/CR-2787 +, NUREG/CR-2802 +, NUREG/CR-3085 +, NUREG/CR-3200 +, NUREG/CR-2239, Forwards Slides/Handout Used at 821110 Presentation of NUREG/CR-2239 Re Sandia Siting or Consequences Study. Transcript of Meeting in PDR Did Not Include Slides +, NUREG-0850, Forwards Vol 1 to NUREG-0850, Preliminary Assessment of Core Melt Accidents at Zion & Indian Point Nuclear Power Plants & Strategies for Mitigating Effects. Vol 1 Only Vol Available at Present +, NUREG-0880, Transcript of 860619 Commission Meeting in Washington,Dc Re Safety Goals.Pp 1-58.Related Info Encl +, NUREG-0611, Forwards Safety Evaluation Finding Auxiliary Feedwater Sys Adequately Designed,Maintained & Operated.Finding Predicated on Implementation of Listed Recommendations Agreed to During 860929-1002 Visit +, NUREG/CR-1659, Forwards Reactor Safety Study Methodology Applications Program Rept for plant,NUREG/CR-1659.Safety Evaluation Based on Rept & Other Info to Be Prepared by Nrc.Util Views Requested.W/O Encl +, NUREG/CR-3300, Informs That Results of Sandia & BNL Probabilistic Safety Study Review Contained in Draft NUREG/CR-3300,Vols I & Ii. W/O Encl +, NUREG-0666, 03-17-81 NUREG-0666, a Probabilistic Safety Analysis of DC Power Supply Requirements at Nuclear Power Plants +  and NUREG-1070, 07-18-84 ACRS Report on NUREG-1070, NRC Policy on Future Reactor Designs - Decisions on Severe Accident Issues +

ES +, ST +  and RT +

9x10-7 +, 9x10-s +, 2CCW-2SW +  and 1CCW-35W +

Ohio +, California +, Idaho +, New York +  and Arkansas +

Steam Generator +, Reactor Coolant System +, Feedwater +, Service water +, Reactor Protection System +, Main Steam Isolation Valve +, Auxiliary Feedwater +, Reactor Pressure Vessel +, Residual Heat Removal +, Emergency Core Cooling System +, Decay Heat Removal +  and Containment Spray +

Design basis earthquake +, Loss of Offsite Power +, Earthquake +, Backfit +  and Fatality +