ASTM E900 - 02(2007)

    Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials, E706 (IIF)

    Active Standard ASTM E900 | Developed by Subcommittee: E10.02

    Book of Standards Volume: 12.02

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    Historical (view previous versions of standard)

    ASTM E900

    Significance and Use

    Operation of commercial power reactors must conform to pressure-temperature limits during heatup and cooldown to prevent over-pressurization at temperatures that might cause nonductile behavior in the presence of a flaw. Radiation damage to the reactor vessel beltline region is compensated for by adjusting the pressure-temperature limits to higher temperature as the neutron damage accumulates. The present practice is to base that adjustment on the increase in transition temperature produced by neutron irradiation as measured at the Charpy V-notch 30-ft·lbf (41-J) energy level. To establish pressure temperature operating limits during the operating life of the plant, a prediction of adjustment in transition temperature must be made.

    4.1.1 In the absence of surveillance data for a given reactor (see Practice E 185), the use of calculative procedures will be necessary to make the prediction. Even when credible surveillance data are available, it will usually be necessary to extrapolate the data to obtain an adjustment in transition temperature for a specific time in the plant operating life. The embrittlement correlation presented herein has been developed for those purposes.

    Research has established that certain elements, notably copper and nickel, cause a variation in radiation sensitivity of steels. The importance of other elements, such as phosphorus (P), remains a subject of additional research. Copper and nickel are the key chemistry parameters used in developing the calculative procedures described here.

    Only power reactor surveillance data were used in the derivation of these procedures. The measure of fast neutron fluence used in the procedure is n/cm2 (E > 1 MeV). Differences in the neutron fluence rate and neutron energy spectra experienced in power reactors and test reactors have not been applied in these procedures. The manner in which these factors were considered is addressed elsewhere.3

    1. Scope

    1.1 This guide presents a method for predicting reference transition temperature adjustments for irradiated light-water cooled power reactor pressure vessel materials based on Charpy V-notch 30-ftlbf (41-J) data. Radiation damage calculative procedures have been developed from a statistical analysis of an irradiated material database that was available as of May 2000. The embrittlement correlation used in this guide was developed using the following variables: copper and nickel contents, irradiation temperature, and neutron fluence. The form of the model was based on current understanding for two mechanisms of embrittlement: stable matrix damage (SMD) and copper-rich precipitation (CRP); saturation of copper effects (for different weld materials) was included. This guide is applicable for the following specific materials, copper, nickel, and phosphorus contents, range of irradiation temperature, and neutron fluence based on the overall database:

    1.1.1 MaterialsA 533 Type B Class 1 and 2, A302 Grade B, A302 Grade B (modified), A508 Class 2 and 3.

    Submerged arc welds, shielded arc welds, and electroslag welds for materials in .

    1.1.2 Copper contents within the range from 0 to 0.50 wt %.

    1.1.3 Nickel content within the range from 0 to 1.3 wt %.

    1.1.4 Phosphorus content within the range 0 to 0.025 wt %.

    1.1.5 Irradiation exposure temperature within the range from 500 to 570F (260 to 299C).

    1.1.6 Neutron fluence within the range from 1 1016 to 8 1019 n/cm2 (E > 1 MeV).

    1.1.7 Neutron energy spectra within the range expected at the reactor vessel core beltline region of light water cooled reactors and fluence rate within the range from 2 108 to 1 1012 n/cm2s (E > 1 MeV).

    1.2 The basis for the method of adjusting the reference temperature is discussed in a separate report.

    1.3 This guide is Part IIF of Master Matrix E 706 which coordinates several standards used for irradiation surveillance of light-water reactor vessel materials. Methods of determining the applicable fluence for use in this guide are addressed in Master Matrix E 706, Practices E 560 (IC) and Guide E 944 (IIA), and Test Method E 1005 (IIIA). The overall application of these separate guides and practices is described in Practice E 853 (IA).

    1.4 The values given in customary U.S. units are to be regarded as the standard. The SI values given in parentheses are for information only.

    1.5 This standard guide does not define how the shift in transition temperature should be used to determine the final adjusted reference temperature. (That would typically include consideration of the initial starting point, the predicted shift, and the uncertainty in the shift estimation method.)

    This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

    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

    E185 Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels

    E560 Practice for Extrapolating Reactor Vessel Surveillance Dosimetry Results, E 706(IC)

    E693 Practice for Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements Per Atom (DPA), E 706(ID)

    E706 Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standards, E 706(0)

    E853 Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results, E706(IA)

    E944 Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance, E 706 (IIA)

    E1005 Test Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance, E 706 (IIIA)

    ICS Code

    ICS Number Code 27.120.10 (Reactor engineering)

    UNSPSC Code

    UNSPSC Code 26140000(Atomic and nuclear energy machinery and equipment)

    DOI: 10.1520/E0900-02R07

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