ASTM E2215 - 10

    Standard Practice for Evaluation of Surveillance Capsules from Light-Water Moderated Nuclear Power Reactor Vessels

    Active Standard ASTM E2215 | Developed by Subcommittee: E10.02

    Book of Standards Volume: 12.02


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    Significance and Use

    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.

    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 E185. Between its provisional adoption in 1961 and its replacement linked to this standard, Practice E185 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 E185 that covers the design of new surveillance programs and this standard practice that covers the testing and evaluation of surveillance capsules. Modifications to the standard test program and supplemental tests are described in Guide E636.

    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 E185.

    The radiation-induced changes in the properties of the vessel are generally monitored by measuring the Charpy index temperatures, 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 E185. The application of this data is the subject of Guide E900 and other documents listed in Section 2.

    Alternative methods exist for testing surveillance capsule materials. Some supplemental and alternative testing methods are available as indicated in Guide E636. Direct measurement of the fracture toughness is also feasible using the To Reference Temperature method defined in Test Method E1921 or J-integral techniques defined in Test Method E1820. Additionally, hardness testing can be used to supplement standard methods as a means of monitoring the radiation response of the materials.

    The methodology to be used in the analysis and interpretation of neutron dosimetry data and the determination of neutron fluence is defined in Practice E853.

    Guide E900 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.

    Guide E509 provides direction for development of a procedure for conducting an in-service thermal anneal of a light-water cooled nuclear reactor vessel and demonstrating the effectiveness of the procedure including a post-annealing vessel radiation surveillance program.

    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 E185, 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 are described in Guide E636.

    1.5 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.


    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

    ASMEBoilerandPressur 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


    ICS Code

    ICS Number Code 27.120.10 (Reactor engineering)

    UNSPSC Code

    UNSPSC Code 39121015(Reactors); 26142100(Nuclear reactor equipment)


    Referencing This Standard

    DOI: 10.1520/E2215-10

    ASTM International is a member of CrossRef.

    Citation Format

    ASTM E2215-10, Standard Practice for Evaluation of Surveillance Capsules from Light-Water Moderated Nuclear Power Reactor Vessels, ASTM International, West Conshohocken, PA, 2010, www.astm.org

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