1. Scope

1.1 This practice covers the evaluation of test specimens and dosimetry from light water moderated nuclear power reactor pressure vessel surveillance capsules.

1.2 This practice is one of a series of standard practices that outline the surveillance program required for nuclear reactor pressure vessels. The surveillance program monitors the radiation-induced changes in the ferritic steels that comprise the beltline of a light-water moderated nuclear reactor pressure vessel.

1.3 This practice along with its companion surveillance program practice, Practice E 185, is intended for application in monitoring the properties of beltline materials in any light-water moderated nuclear reactor.

1.4 Modifications to the standard test program and supplemental tests will be described in a separate Standard that is under development to accompany this standard practice and Practice E 185.

2. Referenced Documents (purchase separately) The documents listed below are referenced within the subject standard but are not provided as part of the standard.

ASTM Standards

A370 Test Methods and Definitions for Mechanical Testing of Steel Products

E8/E8M Test Methods for Tension Testing of Metallic Materials

E21 Test Methods for Elevated Temperature Tension Tests of Metallic Materials

E23 Test Methods for Notched Bar Impact Testing of Metallic Materials

E170 Terminology Relating to Radiation Measurements and Dosimetry

E185 Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels

E208 Test Method for Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels

E509 Guide for In-Service Annealing of Light-Water Moderated Nuclear Reactor Vessels

E560 Practice for Extrapolating Reactor Vessel Surveillance Dosimetry Results, E 706(IC)

E636 Guide for Conducting Supplemental Surveillance Tests for Nuclear Power Reactor Vessels, E 706 (IH)

E853 Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results, E706(IA)

E900 Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials, E706 (IIF)

E1214 Guide for Use of Melt Wire Temperature Monitors for Reactor Vessel Surveillance, E 706 (IIIE)

E1253 Guide for Reconstitution of Irradiated Charpy-Sized Specimens

E1820 Test Method for Measurement of Fracture Toughness

E1921 Test Method for Determination of Reference Temperature, To, for Ferritic Steels in the Transition Range

ASME Standards

Keywords

irradiation; nuclear reactor vessels (light water moderated); radiation exposure; surveillance (of nuclear reactor vessels); Irradiance/irradiation--nuclear applications; Nuclear reactor vessels--light-water cooled; Nuclear reactor vessels--surveillance; Radiation exposure--nuclear materials/applications;

ICS Code

ICS Number Code 27.120.10 (Reactor engineering)

DOI: 10.1520/E2215-02

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