ASTM E2215

4.1 Neutron radiation effects are considered in the design of light-water moderated nuclear power reactors. Changes in system operating parameters may be made throughout the service life of the reactor to account for these effects. A surveillance program is used to measure changes in the properties of actual vessel materials due to the irradiation environment. This practice describes the criteria that should be considered in evaluating surveillance program test capsules.

4.2 Prior to the first issue date of this standard, the design of surveillance programs and the testing of surveillance capsules were both covered in a single standard, Practice E 185. Between its provisional adoption in 1961 and its replacement linked to this standard, Practice E 185 was revised many times (1966, 1970, 1973, 1979, 1982, 1993 and 1998). Therefore, capsules from surveillance programs that were designed and implemented under early versions of the standard were often tested after substantial changes to the standard had been adopted. For clarity, the standard practice for surveillance programs has been divided into the new Practice E 185 that covers the design of new surveillance programs and this standard practice that covers the testing and evaluation of surveillance capsules. A future standard is planned which will recommend procedures for modifying and supplementing existing surveillance programs both in terms of design and testing.

4.3 This standard practice is intended to cover testing and evaluation of all light-water moderated reactor pressure vessel surveillance capsules. The practice is applicable to testing of capsules from surveillance programs designed and implemented under all previous versions of Practice E 185.

4.4 The radiation-induced changes in the properties of the vessel are generally monitored by measuring the Charpy transition temperature, the Charpy upper shelf energy and the tensile properties of specimens from the surveillance program capsules. The significance of these radiation-induced changes is described in Practice E 185. The application of this data is the subject of Guide E 900 and other documents listed in Section 2.

4.5 Alternative methods exist for testing surveillance capsule materials. Some supplemental and alternative testing methods are available as indicated in Practice E 636. Direct measurement of the fracture toughness is also feasible using the To Reference Temperature method defined in Test Method E 1921 or J-integral techniques defined in Test Method E 1820. Additionally hardness testing can be used to supplement standard methods as a means of monitoring the radiation response of the materials.

4.6 The methodology to be used in the analysis and interpretation of neutron dosimetry data and the determination of neutron fluence is defined in Practice E 853.

4.7 Guide E 900 describes the bases used to evaluate the radiation-induced changes in Charpy transition temperature for reactor vessel materials and provides a methodology for predicting future values.

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

A370 Test Methods and Definitions for Mechanical Testing of Steel Products
A751 Test Methods, Practices and Terminology for Chemical Analysis of Steel Products
E1214 Guide for Use of Melt Wire Temperature Monitors for Reactor Vessel Surveillance, E706 (IIIE)
E1253 Guide for Reconstitution of Irradiated Charpy Size Specimens
E170 Terminology Relating to Radiation Measurements and Dosimetry
E1820 Test Method for Measurement of Fracture Toughness
E185 Practice for Conducting Surveillance Tests for Light-Water Moderated Nuclear Power Reactor Vessels
E1921 Test Method for the Determination of Reference Temperature, T, for Ferritic Steels in the Transition Range
E208 Test Method for Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels
E21 Test Methods for Elevated Temperature Tension Tests of Metallic Materials
E23 Test Methods for Notched Bar Impact Testing of Metallic Materials
E482 Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance, E706 (IID)
E509 Guide for In-Service Annealing of Light-Water Cooled Nuclear Reactor Vessels
E560 Practice for Extrapolating Reactor Vessel Surveillance Dosimetry Results, E706 (IC)
E636 Practice for Conducting Supplemental Surveillance Tests for Nuclear Power Reactor Vessels, E706 (IH)
E693 Practice for Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements per Atom (DPA), (ID)
E706 Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standards, E706 (O)
E8 Test Methods for Tension Testing of Metallic Materials
E844 Guide for Sensor Set Design and Irradiation for Reactor Surveillance, E706 (IIC)
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)
American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Sections III and XI
ASME Boiler and Pressure Vessel Code Case N-629, Use of Fracture Toughness Test Data to Establish Reference Temperature for Pressure Retaining Materials, Section XI, Division 1
ASME Boiler and Pressure Vessel Code Case N-631 Use of Fracture Toughness Test Data to Establish Reference Temperature for Pressure Retaining Materials Other Than Bolting for Class 1 Vessels, Section III, Division 1

Index Terms

irradiation; nuclear reactor vessels (light water moderated); radiation exposure; surveillance (of nuclear reactor vessels); ICS Number Code 27.120.10

Citing ASTM Standards

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