Codes and Standards - std0100t0191

Regulatory

Guide 1.4, "Assumptions Used for

Evaluating the Potential Radiological Consequences of a Loss-of-

Coolant Accident for Pressurized Water Reactors."

Guide 1.5, "Assumptions Used for

Regulatory

Evaluating the Potential Radiological Consequences of a Steam

Line Break Accident for Boiling Water Reactors."

Regulatory Guide 1.7, "Control of Combustible Gas

Concentrations in Containment Following a Loss-of-Coolant Acci-

dent."

Regulatory Guide

1.11, "Instrument Lines Pene-

trating Primary Reactor Containment."

Regulatory

1.53, "Application of the

Guide

Single-Failure Criterion to Nuclear Power Plant Protection Sys-

tems."

Regulatory

Guide

1.105, "Instrument Setpoints for

Safety-Related Systems."

Regulatory Guide 1.145, "Atmospheric Dispersion

Models for Potential Accident Consequence Assessments at Nuclear

Power Plants."

1.154, "Format and Content Of

Regulatory Guide

Plant-Specific Pressurized Thermal Shock Safety Analysis Reports

for Pressurized Water Reactors."

SRP

15.6.5,

Appendix A, "Radiological Con-

sequences of a Design Basis Loss-of-Coolant Accident Including

Containment Leakage Contribution.

SRP

15.6.5,

Appendix B, "Radiological Con-

sequences of a Design Basis Loss-of-Coolant Accident:

Leakage

from Engineered-Safety-Feature Components Outside Containment."

SRP

15.6.5,

Appendix D, "Radiological Con-

sequences of a Design Basis Loss-of-Coolant Accident:

Leakage

from Main Steam Isolation Valve Leakage Control System (BWR)."

15.6.3

Codes and Standards

ANSI/ISA-S67.04-1982 , "Setpoints for Nuclear

Safety-Related Instrumentation Used in Nuclear Power Plants."

ASME Boiler and Pressure Vessel Code, "Rules for

Construction of Nuclear Power Plant Components," Section III,

Division 1, Article NB-7000, "Protection Against Overpressure."

15-13