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 10 ¹⁷ n/cm ² (1 x 10 ²¹ n/m ² ) 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 1The 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 (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

A751 Test Methods, Practices, and Terminology for Chemical Analysis 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

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

E482 Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance, E706 (IID)

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

E844 Guide for Sensor Set Design and Irradiation for Reactor Surveillance, E 706 (IIC)

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

E2215 Practice for Evaluation of Surveillance Capsules from Light-Water Moderated Nuclear Power Reactor Vessels

ASME Standards

Keywords

Nuclear applications/materials--steel; Nuclear reactor vessels--light-water cooled; Nuclear reactor vessels--surveillance; Radiation exposure--nuclear materials/applications; Repair by welding (for steel for nuclear/special applications); Steel screens ;

ICS Code

ICS Number Code 27.120.10 (Reactor engineering)

DOI: 10.1520/E0185-02

ASTM International is a member of CrossRef.