ASTM E185

Predictions of neutron radiation effects on pressure vessel steels are considered in the design of light-water moderated nuclear power reactors. Changes in system operating parameters often are made throughout the service life of the reactor vessel to account for radiation effects. Due to the variability in the behavior of reactor vessel steels, a surveillance program is warranted to monitor changes in the properties of actual vessel materials caused by long-term exposure to the neutron radiation and temperature environment of the reactor vessel. This practice describes the criteria that should be considered in planning and implementing surveillance test programs and points out precautions that should be taken to ensure that: (1) capsule exposures can be related to beltline exposures, (2) materials selected for the surveillance program are samples of those materials most likely to limit the operation of the reactor vessel, and (3) the tests yield results useful for the evaluation of radiation effects on the reactor vessel.

The methodology to be used in estimation of neutron exposure obtained for reactor vessel surveillance programs is defined in Guide E 482, which establishes the bases to be used to evaluate both the design and future condition of the reactor vessel.

The design of a surveillance program for a given reactor vessel must consider the existing body of data on similar materials in addition to the specific materials used for that reactor vessel. The amount of such data and the similarity of exposure conditions and material characteristics will determine their applicability for predicting radiation effects.

1. Scope

1.1 This practice covers procedures for designing a surveillance program for monitoring the radiation-induced changes in the mechanical properties of ferritic materials in the beltline of light-water moderated nuclear power reactor vessels. This practice includes the minimum requirements for the design of a surveillance program, selection of vessel material to be included, and a schedule for evaluation of materials.

1.2 This practice was developed for all light-water moderated nuclear power reactor vessels for which the predicted maximum fast neutron fluence (E > 1 MeV) at the end of the design lifetime (EOL) exceeds 1 x 10¹⁷ n/cm² (1 x 10²¹ n/m²) at the inside surface of the reactor vessel.

1.3 This practice applies only to the planning and design of surveillance programs for reactor vessels designed and built after the effective date of this practice. Previous versions of Practice E 185 apply to earlier reactor vessels.

1.4 This practice does not provide specific procedures for monitoring the radiation induced changes in properties beyond the design life, but the procedure described may provide guidance for developing such a surveillance program.

Note 1—The increased complexity of the requirements for a light-water moderated nuclear power reactor vessel surveillance program has necessitated the separation of the requirements into three related standards. Practice E 185 describes the minimum requirements for a surveillance program. Practice E 2215, "Standard Practice for the Evaluation of Surveillance Capsules from Light-Water Moderated Nuclear Power Reactor Vessels" describes the procedures for testing and evaluation of surveillance capsules removed from a surveillance program as defined in the current or previous editions of Practice E 185. Another standard guide for supplementing existing light-water moderated nuclear power reactor vessel surveillance programs is under preparation. A summary of the many major revisions to Practice E 185 since its original issuance is contained in Appendix X1.

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
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
E2215 Practice for the Evaluation of Surveillance Capsules from Light-Water Moderated Nuclear Power Reactor Vessels
E23 Test Methods for Notched Bar Impact Testing of Metallic Materials
E399 Test Method for Plane-Strain Fracture Toughness of Metallic Materials
E482 Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance, E706 (IID)
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 Exposure in Ferritic Steels in Terms of Displacements per Atom (DPA), (ID)
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

radiation-induced; ferritic; vessel; light-water; reactor vessels; ICS Number Code 27.120.10

Citing ASTM Standards

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