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Significance and Use
4.1 Operation of commercial power reactors must conform to pressure-temperature limits during heatup and cooldown to prevent over-pressurization at temperatures that might cause non-ductile behavior in the presence of a flaw. Radiation damage to the reactor vessel is compensated for by adjusting the pressure-temperature limits to higher temperatures as the neutron damage accumulates. The present practice is to base that adjustment on the TTS produced by neutron irradiation as measured at the Charpy V-notch 41-J (30-ft·lbf) energy level. To establish pressure temperature operating limits during the operating life of the plant, a prediction of TTS must be made.
4.1.1 In the absence of surveillance data for a given reactor material (see Practice and ), the use of calculative procedures are necessary to make the prediction. Even when credible surveillance data are available, it will usually be necessary to interpolate or extrapolate the data to obtain a TTS for a specific time in the plant operating life. The embrittlement correlation presented herein has been developed for those purposes.
4.2 Research has established that certain elements, notably copper (Cu), nickel (Ni), phosphorus (P), and manganese (Mn), cause a variation in radiation sensitivity of reactor pressure vessel steels. The importance of other elements, such as silicon (Si), and carbon (C), remains a subject of additional research. Copper, nickel, phosphorus, and manganese are the key chemistry parameters used in developing the calculative procedures described here.
4.3 Only power reactor (PWR and BWR) surveillance data were used in the derivation of these procedures. The measure of fast neutron fluence used in the procedure is n/m2 (E > 1 MeV). Differences in fluence rate and neutron energy spectra experienced in power reactors and test reactors have not been accounted for in these procedures.
1.1 This guide presents a method for predicting values of reference transition temperature shift (TTS) for irradiated pressure vessel materials. The method is based on the TTS exhibited by Charpy V-notch data at 41-J (30-ft·lbf) obtained from surveillance programs conducted in several countries for commercial pressurized (PWR) and boiling (BWR) light-water cooled (LWR) power reactors. An embrittlement correlation has been developed from a statistical analysis of the large surveillance database consisting of radiation-induced TTS and related information compiled and analyzed by Subcommittee E10.02. The details of the database and analysis are described in a separate report (. ),, This embrittlement correlation was developed using the variables copper, nickel, phosphorus, manganese, irradiation temperature, neutron fluence, and product form. Data ranges and conditions for these variables are listed in . Section lists the materials included in the database and the domains of exposure variables that may influence TTS but are not used in the embrittlement correlation.
1.1.1 The range of material and irradiation conditions in the database for variables used in the embrittlement correlation:
188.8.131.52 Copper content up to 0.4 %.
184.108.40.206 Nickel content up to 1.7 %.
220.127.116.11 Phosphorus content up to 0.03 %.
18.104.22.168 Manganese content within the range from 0.55 to 2.0 %.
22.214.171.124 Irradiation temperature within the range from 255 to 300°C (491 to 572°F).
126.96.36.199 Neutron fluence within the range from 1 × 1021 n/m2 to 2 × 1024 n/m2 (E> 1 MeV).
188.8.131.52 A categorical variable describing the product form (that is, weld, plate, forging).
1.1.2 The range of material and irradiation conditions in the database for variables not included in the embrittlement correlation:
184.108.40.206 Type B Class 1 and 2, Grade B, Grade B (modified), and Class 2 and 3. Also, European and Japanese steel grades that are equivalent to these ASTM Grades.
220.127.116.11 Submerged arc welds, shielded arc welds, and electroslag welds having compositions consistent with those of the welds used to join the base materials described in .
18.104.22.168 Neutron fluence rate within the range from 3 × 1012 n/m2/s to 5 × 1016 n/m2/s (E > 1 MeV).
22.214.171.124 Neutron energy spectra within the range expected at the reactor vessel region adjacent to the core of commercial PWRs and BWRs (greater than approximately 500MW electric).
126.96.36.199 Irradiation exposure times of up to 25 years in boiling water reactors and 31 years in pressurized water reactors.
1.2 It is the responsibility of the user to show that the conditions of interest in their application of this guide are addressed adequately by the technical information on which the guide is based. It should be noted that the conditions quantified by the database are not distributed evenly over the range of materials and irradiation conditions described in , and that some combination of variables, particularly at the extremes of the data range are under-represented. Particular attention is warranted when the guide is applied to conditions near the extremes of the data range used to develop the TTS equation and when the application involves a region of the data space where data is sparse. Although the embrittlement correlation developed for this guide was based on statistical analysis of a large database, prudence is required for applications that involve variable values beyond the ranges specified in . Due to strong correlations with other exposure variables within the database (that is, fluence), and due to the uneven distribution of data within the database (for example, the irradiation temperature and flux range of PWR and BWR data show almost no overlap) neither neutron fluence rate nor irradiation time sufficiently improved the accuracy of the predictions to merit their use in the embrittlement correlation in this guide. Future versions of this guide may incorporate the effect of neutron fluence rate or irradiation time, or both, on TTS , as such effects are described in (. The irradiated material database, the technical basis for developing the embrittlement correlation, and issues involved in its application, are discussed in a separate report )(. That report describes the nine different )TTS equations considered in the development of this guide, some of which were developed using more limited datasets (for example, national program data (). If the material variables or exposure conditions of a particular application fall within the range of one of these alternate correlations, it may provide more suitable guidance. , )
1.3 This guide is expected to be used in coordination with several standards addressing irradiation surveillance of light-water reactor vessel materials. Method of determining the applicable fluence for use in this guide are addressed in Guides , , and Test Method . The overall application of these separate guides and practices is described in Practice .
1.4 The values stated in SI units are to be regarded as standard. The values given in parentheses are mathematical conversions to U.S. Customary units that are provided for information only and are not considered standard.
1.5 This standard guide does not define how the TTS should be used to determine the final adjusted reference temperature, which would typically include consideration of the transition temperature before irradiation, the predicted TTS , and the uncertainties in the shift estimation method.
1.6 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.
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 Number Code 27.120.10 (Reactor engineering)
UNSPSC Code 26140000(Atomic and nuclear energy machinery and equipment)
ASTM E900-15, Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials, ASTM International, West Conshohocken, PA, 2015, www.astm.orgBack to Top