Codes and Standards - std0100t0189

Regulatory Guide

1.154, "Format and Content of

Plant-Specific Pressurized Thermal Shock Safety Analysis Reports

for Pressurized Water Reactors."

15.5.3

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."

379-1972, "IEEE Trial-Use Guide for the

IEEE

Application of the Single-Failure Criterion to Nuclear Power Gen-

erating Station Class 1E Systems." (This standard was revised and

issued as IEEE 379-1988, "IEEE Standard Application of the

Single-Failure Criterion to Nuclear Power Generating Station

Class 1E Systems.")

15.5.4

Supplemental Information

NUREG-0800,

USNRC "Standard Review Plan," June

1987.

SRP

15.5.1, "Inadvertent Operation of ECCS

That Increases Reactor Coolant Inventory."

SRP 15.5.2, "Chemical and Volume Control Sys-

tem Malfunction That Increases Reactor Coolant Inventory."

NUREG-0744, Rev. 1, "Resolution of Task A-11 Reac-

tor Vessel Materials Toughness Safety Issue," Vols. 1 and 2,

USNRC, October 1982.

NUREG-0660, "NRC Action Plan Developed as a Result

of the TMI-2 Accident," USNRC, Vol. 1, May 1980 and Vol. 1, Rev.

1, August 1980.

NUREG-0224,

"Final Report on Reactor Vessel Pres-

sure Transient Protection for Pressurized Water Reactors," USNRC,

September 1978.

This section identifies criteria for initiating events

that involve inadvertent opening of pressure relief valves, fail-

ure of small lines carrying primary coolant outside containment,

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