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Significance and Use
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 . Between its provisional adoption in 1961 and its replacement linked to this standard, Practice 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 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 .
4.3 This 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 .
4.4 The radiation-induced changes in the properties of the reactor pressure vessel are generally monitored by measuring the index temperatures, the 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 .
4.5 Alternative methods exist for testing surveillance capsule materials. Some supplemental and alternative testing methods are available as indicated in Guide . Direct measurement of the fracture toughness is also feasible using the To Reference Temperature method defined in Test Method or J-integral techniques defined in Test Method . Additionally, hardness testing can be used to supplement standard methods as a means of monitoring the irradiation response of the materials.
4.6 Practice describes a methodology that may be used in the analysis and interpretation of neutron dosimetry data and the determination of neutron fluence. Regulators or other sources may describe different methods.
4.7 Guide describes a method for predicting the TTS. Regulators or other sources may describe different methods for predicting TTS.
4.8 Guide 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.
1.1 This practice covers the evaluation of test specimens and dosimetry from light water moderated nuclear power reactor pressure vessel surveillance capsules.
1.2 Additionally, this practice provides guidance on reassessing withdrawal schedule for design life and operation beyond design life.
1.3 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 irradiation-induced changes in the ferritic steels that comprise the beltline of a light-water moderated nuclear reactor pressure vessel.
1.4 This practice along with its companion surveillance program practice, Practice , is intended for application in monitoring the properties of beltline materials in any light-water moderated nuclear reactor.
1.5 Modifications to the standard test program and supplemental tests are described in Guide .
1.6 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.
1.7 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.
2. Referenced Documents (purchase separately) The documents listed below are referenced within the subject standard but are not provided as part of the standard.
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
E636 Guide for Conducting Supplemental Surveillance Tests for Nuclear Power Reactor Vessels
E693 Practice for Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements Per Atom (DPA)
E844 Guide for Sensor Set Design and Irradiation for Reactor Surveillance
E853 Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results
E900 Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials
E1214 Guide for Use of Melt Wire Temperature Monitors for Reactor Vessel Surveillance
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 StandardsBoiler and Pressure Vessel Code, Section III Subarticle NB-2000, Rules for Construction of Nuclear Facility Components, Class 1 Components, Materials Boiler and Pressure Vessel Code, Section XI
ICS Number Code 27.120.10 (Reactor engineering)
|Link to Active (This link will always route to the current Active version of the standard.)|
ASTM E2215-18, Standard Practice for Evaluation of Surveillance Capsules from Light-Water Moderated Nuclear Power Reactor Vessels, ASTM International, West Conshohocken, PA, 2018, www.astm.orgBack to Top