Active Standard ASTM E509 | Developed by Subcommittee: E10.02
Book of Standards Volume: 12.02
Historical (view previous versions of standard)
This guide covers the general procedures to be considered (as well as the demonstration of their effectiveness) for conducting in-service thermal anneals of light-water moderated nuclear reactor vessels. The purpose of this in-service annealing (heat treatment) is to improve the mechanical properties, especially fracture toughness, of the reactor vessel materials previously degraded by neutron embrittlement. This guide is designed to accommodate the variable response of reactor-vessel materials in post-irradiation annealing at various temperatures and different time periods, as well as to provide direction for the development of a vessel annealing procedure and a post-annealing vessel radiation surveillance program. Guidelines are provided to determine the post-anneal reference nil-ductility transition temperature (RTNDT), the Charpy V-notch upper shelf energy level, fracture toughness properties, and the predicted reembrittlement trend for these properties for reactor vessel beltline materials.
This abstract is a brief summary of the referenced standard. It is informational only and not an official part of the standard; the full text of the standard itself must be referred to for its use and application. ASTM does not give any warranty express or implied or make any representation that the contents of this abstract are accurate, complete or up to date.
Significance and Use
Reactor vessels made of ferritic steels are designed with the expectation of progressive changes in material properties resulting from in-service neutron exposure. In the operation of light-water-cooled nuclear power reactors, changes in pressure-temperature (P – T) limits are made periodically during service life to account for the effects of neutron radiation on the ductile-to-brittle transition temperature material properties. If the degree of neutron embrittlement becomes large, the restrictions on operation during normal heat-up and cool down may become severe. Additional consideration should be given to postulated events, such as pressurized thermal shock (PTS). A reduction in the upper shelf toughness also occurs from neutron exposure, and this decrease may reduce the margin of safety against ductile fracture. When it appears that these situations could develop, certain alternatives are available that reduce the problem or postpone the time at which plant restrictions must be considered. One of these alternatives is to thermally anneal the reactor vessel beltline region, that is, to heat the beltline region to a temperature sufficiently above the normal operating temperature to recover a significant portion of the original fracture toughness and other material properties that were lost as a result of neutron embrittlement.
Preparation and planning for an in-service anneal should begin early so that pertinent information can be obtained to guide the annealing operation. Sufficient time should be allocated to evaluate the expected benefits in operating life to be gained by annealing; to evaluate the annealing method to be employed; to perform the necessary system studies and stress evaluations; to evaluate the expected annealing recovery and reembrittlement behavior; to develop and functionally test such equipment as may be required to do the in-service annealing; and, to train personnel to perform the anneal.
Selection of the annealing temperature requires a balance of opposing conditions. Higher annealing temperatures, and longer annealing times, can produce greater recovery of fracture toughness and other material properties and thereby increase the post-anneal lifetime. The annealing temperature also can have an impact on the reembrittlement trend after the anneal. On the other hand, higher temperatures can create other undesirable property effects such as permanent creep deformation or temper embrittlement. These higher temperatures also can cause engineering difficulties, that is, core and coolant removal and storage, localized heating effects, etc., in preventing the annealing operation from distorting the vessel or damaging vessel supports, primary coolant piping, adjacent concrete, insulation, etc. See ASME Code Case N-557 for further guidance on annealing conditions and thermal-stress evaluations (2).
When a reactor vessel approaches a state of embrittlement such that annealing is considered, the major criterion is the number of years of additional service life that annealing of the vessel will provide. Two pieces of information are needed to answer the question: the post-anneal adjusted RTNDT and upper shelf energy level, and their subsequent changes during future irradiation. Furthermore, if a vessel is annealed, the same information is needed as the basis for establishing pressure-temperature limits for the period immediately following the anneal and demonstrating compliance with other design requirements and the PTS screening criteria. The effects on upper shelf toughness similarly must be addressed. This guide primarily addresses RTNDT changes. Handling of the upper shelf is possible using a similar approach as indicated in NRC Regulatory Guide 1.162. Appendix X1 provides a bibliography of existing literature for estimating annealing recovery and reembrittlement trends for these quantities as related to U.S. and other country pressure-vessel steels, with primary emphasis on U.S. steels.
A key source of test material for determining the post-anneal RTNDT, upper shelf energy level, and the reembrittlement trend is the original surveillance program, provided it represents the critical materials in the reactor vessel. Appendix X2 describes an approach to estimate changes in RTNDT both due to the anneal and after reirradiation. The first purpose of Appendix X2 is to suggest ways to use available materials most efficiently to determine the post-anneal RTNDT and to predict the reembrittlement trend, yet leave sufficient material for surveillance of the actual reembrittlement for the remaining service life. The second purpose is to describe alternative analysis approaches to be used to assess test results of archive (or representative) materials to obtain the essential post-anneal and reirradiation RTNDT, upper shelf energy level, or fracture toughness, or a combination thereof.
An evaluation must be conducted of the engineering problems posed by annealing at the highest practical temperature. Factors required to be investigated to reduce the risk of distortion and damage caused by mechanical and thermal stresses at elevated temperatures to relevant system components, structures, and control instrumentation are described in 5.1.3 and 5.1.4.
Throughout the annealing operation, accurate measurement of the annealing temperature at key defined locations must be made and recorded for later engineering evaluation.
After the annealing operation has been carried out, several steps should be taken. The predicted improvement in fracture toughness properties must be verified, and it must be demonstrated that there is no damage to key components and structures.
Further action may be required to demonstrate that reactor vessel integrity is maintained within ASME Code requirements such as indicated in the referenced ASME Code Case N-557 (2). Such action is beyond the scope of this guide.
1.1 This guide covers the general procedures to be considered for conducting an in-service thermal anneal of a light-water moderated nuclear reactor vessel and demonstrating the effectiveness of the procedure. The purpose of this in-service annealing (heat treatment) is to improve the mechanical properties, especially fracture toughness, of the reactor vessel materials previously degraded by neutron embrittlement. The improvement in mechanical properties generally is assessed using Charpy V-notch impact test results, or alternatively, fracture toughness test results or inferred toughness property changes from tensile, hardness, indentation, or other miniature specimen testing (1).
1.2 This guide is designed to accommodate the variable response of reactor-vessel materials in post-irradiation annealing at various temperatures and different time periods. Certain inherent limiting factors must be considered in developing an annealing procedure. These factors include system-design limitations; physical constraints resulting from attached piping, support structures, and the primary system shielding; the mechanical and thermal stresses in the components and the system as a whole; and, material condition changes that may limit the annealing temperature.
1.3 This guide provides direction for development of the vessel annealing procedure and a post-annealing vessel radiation surveillance program. The development of a surveillance program to monitor the effects of subsequent irradiation of the annealed-vessel beltline materials should be based on the requirements and guidance described in Practices E 185
1.4 The values stated in inch-pound or SI units are to be regarded separately as the standard.
1.5 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
E636 Guide for Conducting Supplemental Surveillance Tests for Nuclear Power Reactor Vessels, E 706 (IH)
E900 Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials, E706 (IIF)
E1253 Guide for Reconstitution of Irradiated Charpy-Sized Specimens
E2215 Practice for Evaluation of Surveillance Capsules from Light-Water Moderated Nuclear Power Reactor Vessels
ASME StandardsCodeCaseN-557, In-Place Dry Annealing of a PWR Nuclear Reactor Vessel (Section XI, Division 1)
Nuclear Regulatory Commission DocumentsNRCRegulatoryGuide1. Format and Content of Report for Thermal Annealing of Reactor Pressure Vessels
ICS Number Code 27.120.10 (Reactor engineering)