Proposal for maintenance optimization scheme based on system based code concept

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code currently only allows five materials for use in construction of nuclear components for high temperature service. These are: 2.25Cr-1Mo and V-modified 9Cr-1Mo steels, Types 304 and 316 stainless steels and the high nickel Alloy 800H. Since 2005, the US high temperature gas-cooled reactor program has been characterizing elevated temperature mechanical properties of Alloy 617 as the leading candidate construction material for the intermediate heat exchanger. After analysis of these experimental results, along with historical data and additional results available through the Generation IV International Forum, Very High Temperature Reactor, Materials Program Management Board Materials Handbook, a draft ASME Code Case to allow nuclear construction with Alloy 617 for temperatures up to 1750°F (954°C) has been developed. This report contains the Code Case for Low Temperature Service Construction of Section III, Division 5, Subsection HB, Subpart A, Class A and Subsection HC, Subpart A, Class B components, which has been approved in Section II, Materials, and Section III, Rules for Construction of Nuclear Facility Components. Supporting technical justification for the low temperature Code Case is also included. This Code Case allows use of Alloy 617 up to 800°F (425°C). This report also contains an updated draft of a Section III, Division 5, Subsection HB, Subpart B, Class A Code Case for Alloy 617 to qualify it for use in construction of nuclear components up to 1750°F (954°C) for service life up to 100,000 hours. The draft contained in Appendix 4, subject to editorial revision and approval by the ASME Special Task Group on Alloy 617 Code Qualification, will be submitted for approval by letter ballot by the appropriate ASME Committees. The technical justification supporting the Code Case is presented in Appendix 5 of this report. This background document is part of the information package that will be submitted with the Code Case for ballot.

Due to its exceptional high-temperature properties, Alloy 709 (UNS 31025), an advanced austenitic stainless steel, has been selected by the U.S. Department of Energy, Office of Nuclear Energy, Advanced Reactor Technologies Program for qualification as a Class A construction material in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, Division 5 for high temperature reactor applications. This initiative encompasses a comprehensive staged qualification plan for code cases with progressively longer design lives of 100,000, 300,000, and 500,000 hr, as long-term test data become available. A significant portion of this effort focuses on creating a data package that is essential for establishing material-specific design parameters for Alloy 709, intended for use in sodium fast reactors, molten salt reactors, and gas-cooled reactors. Data generation for the initial 100,000-hr code case was completed at the end of 2024. A key component of this data package involves conducting fatigue and creep-fatigue testing at elevated temperatures to develop fatigue design curves and the damage envelope for the creep-fatigue interaction diagram (D-diagram), both of which are common to all three code cases. Strain-controlled fatigue and creep-fatigue tests have been conducted on three commercial heats of Alloy 709 plates at the Oak Ridge National Laboratory and the Idaho National Laboratory. The comprehensive code case testing matrix includes strain ranges from 0.25 to 3%, hold times of up to 10 hr, and testing temperatures of 21, 427, 649, 760, 816, and 954 °C (70, 800, 1200, 1400, 1500, and 1750 °F). This paper summarizes the results of cyclic testing and provides a comprehensive analysis of the test results. It outlines the methodology for developing temperature-dependent fatigue curves and presents draft fatigue design curves for inclusion in the code cases. Furthermore, creep-fatigue test data with various tension hold times are used to support the development of the creep-fatigue interaction diagram. Both sets of design parameters are crucial for the code case data package.

Evaluation of potential Reactor Pressure Vessel (RPV) steels has been carried out as part of the pre-conceptual Very High Temperature Reactor (VHTR) design studies. These design studies have generally focused on American Society of Mechanical Engineers (ASME) Code status of the steels, temperature limits, and allowable stresses. Initially, three candidate materials were identified by this process: conventional light water reactor (LWR) RPV steels A508 and A533, 2¼Cr-1Mo in the annealed condition, and Grade 91 steel. The low strength of 2¼Cr-1Mo at elevated temperature has eliminated this steel from serious consideration as the VHTR RPV candidate material. Discussions with the very few vendors that can potentially produce large forgings for nuclear pressure vessels indicate a strong preference for conventional LWR steels. This preference is based in part on extensive experience with forging these steels for nuclear components. It is also based on the inability to cast large ingots of the Grade 91 steel due to segregation during ingot solidification, thus restricting the possible mass of forging components and increasing the amount of welding required for completion of the RPV. Grade 91 steel is also prone to weld cracking and must be post-weld heat treated to ensure adequate high-temperature strength. There are also questions about the ability to produce, and very importantly, verify the through thickness properties of thick sections of Grade 91 material. The availability of large components, ease of fabrication, and nuclear service experience with the A508 and A533 steels strongly favor their use in the RPV for the VHTR. Lowering the gas outlet temperature for the VHTR to 750°C from 950 to 1000°C, proposed in early concept studies, further strengthens the justification for this material selection. This steel is allowed in the ASME Boiler and Pressure Vessel Code for nuclear service up to 371°C (700°F); certain excursions above that temperature are allowed by Code Case N-499-2 (now incorporated as an appendix to Section III Division 5 of the Code). This Code Case was developed with a rather sparse data set and focused primarily on rolled plate material (A533 specification). Confirmatory tests of creep behavior of both A508 and A533 are described here that are designed to extend the database in order to build higher confidence in ensuring the structural integrity of the VHTR RPV during off-normal conditions. A number of creep-rupture tests were carried out at temperatures above the 371°C (700°F) Code limit; longer term tests designed to evaluate minimum creep behavior are ongoing. A limited amount of rupture testing was also carried out on welded material. All of the rupture data from the current experiments is compared to historical values from the testing carried out to develop Code Case N-499-2. It is shown that the A508/533 basemetal tested here fits well with the rupture behavior reported from the historical testing. The presence of weldments significantly reduces the time to rupture. The primary purpose of this report is to summarize and record the experimental results in a single document.

3 - Materials

The Oak Ridge National Laboratory (ORNL) is a US Department of Energy (DOE) facility that is managed by UT-Battelle, LLC. In February 2006, DOE promulgated worker safety and health regulations to govern contractor activities at DOE sites. These regulations, which are provided in 10 CFR 851, Worker Safety and Health Program, establish requirements for worker safety and health program that reduce or prevent occupational injuries, illnesses, and accidental losses by providing DOE contractors and their workers with safe and healthful workplaces at DOE sites. The regulations state that contractors must achieve compliance no later than May 25, 2007. According to 10 CFR 851, Subpart C, Specific Program Requirements, contractors must have a structured approach to their worker safety and health programs that at a minimum includes provisions for pressure safety. In implementing the structured approach for pressure safety, contractors must establish safety policies and procedures to ensure that pressure systems are designed, fabricated, tested, inspected, maintained, repaired, and operated by trained, qualified personnel in accordance with applicable sound engineering principles. In addition, contractors must ensure that all pressure vessels, boilers, air receivers, and supporting piping systems conform to (1) applicable American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (2004) Sections I through XII, including applicable code cases; (2) applicable ASME B31 piping codes; and (3) the strictest applicable state and local codes. When national consensus codes are not applicable because of pressure range, vessel geometry, use of special materials, etc., contractors must implement measures to provide equivalent protection and ensure a level of safety greater than or equal to the level of protection afforded by the ASME or applicable state or local codes. This report documents the work performed to address legacy pressure vessel deficiencies and comply with pressure safety requirements in 10 CFR 851. It also describes actions taken to develop and implement ORNL’s Pressure Safety Program.

The required thickness of welding tees is neither specified in ASME (2012, “Factory-Made Wrought Buttwelding Fittings,” American Society of Mechanical Engineers, New York, Standard No. B16.9-2012) nor is a clear calculation method provided in codes such as ASME (2016, “Process Piping,” American Society of Mechanical Engineers, New York, Standard No. B31.3-2016). This can lead to uncertainty regarding the pressure capacity of a tee fitting, particularly one that has suffered from erosion or corrosion. Code methods including area replacement (ASME, 2016, “Process Piping,” American Society of Mechanical Engineers, New York, Standard No. B31.3-2016) or pressure-area (ASME, 2015, “Boiler and Pressure Vessel Code Section VIII Division 2,” American Society of Mechanical Engineers, New York, Standard No. BPVC-VIII-2-2015; BSI, 2014, “Unfired Pressure Vessels Part 3: Design,” BSI, London, UK, Standard No. BS EN 13445-3) do not directly account for the effect which the curvature of the crotch region may have on the stress state in the tee. The approach adopted in this work is to liken the geometry of the tee crotch to the intrados of a torus or pipe bend. The shell theory applicable to the torus is adapted for the tee in order to derive a relationship for circumferential membrane stress. An equivalent tube radius is assigned by determining the local radius of shell curvature in the plane passing through the crotch center of the curvature. The actual stresses in the tee crotch are significantly reduced by the adjoining straight portions. This effect is difficult to quantify theoretically and has thus been investigated by means of finite element analysis (FEA)-based assessments. An empirical relationship was then established providing a conservative correlation between the theoretical stresses and the program calculated local stress intensities.

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC) Committee has recently developed a new Section XI (Nuclear Components Inspection) Division 2 Code named “Reliability and Integrity Management (RIM).” RIM incorporates a new concept known as “System-Based Code (SBC)” originally due to Asada and his colleagues (2001–2004), where an integrated approach from design to service inspection is introduced using three new types of statistical quantities: (1) “system reliability index,” or “system co-reliability target” for any system consisting of structures and components, (2) “structural co-reliability,” for any structure, and (3) “component co-reliability” for any component, where co-reliability is defined as “1 – reliability” and is equal to failure probability. In a recent paper published in the International Journal of Pressure Vessels and Piping (Vol. 173 (2019), pp. 79–93). Fong, Heckert, Filliben, and Freiman developed a new theory of fatigue and creep rupture life modeling for metal alloys at room and elevated temperatures such that the co-reliability of a smooth component can be estimated from fatigue and creep rupture test data with simple loading histories. In this paper, we extend the theory to include a methodology to estimate the failure probability (or, co-reliability) of a stainless steel 316L(N) component undergoing a complex loading history such as a thermal fatigue cycle. To illustrate an application of this new modeling approach, we present a numerical example using (a) the experimental test data of 7 specimens of S.S. 316L(N) in creep at 565 C as published by Kilian Wasmer in his Ph.D. thesis (Imperial College London, 2003), (b) the experimental test data of 10 specimens of S.S. 316L(N) in creep at 650 C also by Wasmer (Ph.D. thesis, 2003), and (c) a thermal fatigue loading history of 1 hour of creep at 650 C at an operating creep stress of 73 MPa, and 300 hours of creep at 565 C at the same creep stress to simulate a hypothetical powerplant thermal fatigue operation. The significance and limitations of this new damage-state-based approach to modeling creep and to estimate the component failure probability (or, co-reliability) of steel components in creep at elevated temperatures with complex histories are discussed.

Fatigue design of Class-1 piping of NPP is carried out as par Section-III of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel code. The fatigue design criteria of ASME are based on the concept of safety factor, which does not provide means for management of uncertainties for consistently reliable designs. In this regard, a work is taken up to estimate the implicit reliability level associated with fatigue design criteria of Class-1 piping specified by ASME Section III, NB-3650. The methodology employed for reliability evaluation is FORM, HORSM and MCS, where limit state function is derived using mean fatigue curve developed by Argonne National Laboratory. A number of important aspects related to reliability of various piping product and joints are discussed, and implicit reliability level is evaluated.

Fatigue is a common failure mode in mechanical components that are subject to repeated load cycles. In nuclear plants the cyclic loading is usually associated with variations of pressure and temperature. The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code provide fatigue evaluation procedures in ASME III Section NB-3220. It is required to calculate a Cumulative Usage Factor (CUF). The Code criterion for acceptance is that the calculated CUF must be less than one. The ASME method is deterministic and it is assumed that uncertainties are accounted for by largely un-quantified inherent safety margins in the assessment. The conservatism in the ASME III Code fatigue evaluation methods may come from the assessment procedure used and/or the fatigue design curve. In this paper the inputs to an ASME III fatigue assessment that is based upon a finite element determination of stresses are outlined and assessed. Based upon these inputs possible areas of conservatism within the analysis are then discussed and conclusion drawn on the inherent margins. In addition a probabilistic model that uses Monte Carlo (MC) methods is developed. By assuming the scatter in the fatigue data, used to generate the design fatigue curve, to be the most significant random variable, the MC model was used to calculate the conditional probability of initiation associated with CUFs from a number of plant representative examples.

Section III, Division 1 and Section VIII, Division 2 of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code provide procedures for demonstrating shakedown using elastic-plastic analysis. While these procedures may be used in place of elastic analysis procedures, they are typically employed after the elastic analysis and simplified elastic-plastic analysis limits have been exceeded. In using the Section III, Division 1 and Section VIII, Division 2 procedures for elastic-plastic shakedown analyses, three concerns are raised. First, the Section III, Division 1 procedure is vague, which can result in inconsistent results between analysts. Second, the acceptance criteria contained in both procedures are vague, which can also result in inconsistent results between analysts. Lastly, differences in the procedures and acceptance criteria can result in demonstration of component elastic-plastic shakedown under Section III, Division 1 but not under Section VIII, Division 2. The authors presume that the ASME Code intends to provide similar design and analysis conclusions, which may not be a correct assumption. To demonstrate these concerns, a nozzle benchmark design subject to a representative thermal and pressure transient was evaluated using the two Code elastic-plastic shakedown procedures. Shakedown was successfully demonstrated using the Section III, Division 1 procedure. However, shakedown could not be demonstrated using the Section VIII, Division 2 procedure. The conflicting results seem to indicate that, for the nozzle design evaluated, the Section VIII, Division 2 procedure is considerably more conservative than the Section III, Division 1 procedure. To further assess the conservative nature of the Section VIII, Division 2 procedure, the nozzle benchmark design was evaluated using the same thermal transient, but without a pressure load. While shakedown was technically not observed using the Section VIII, Division 2 acceptance criteria, engineering judgment concluded that shakedown was demonstrated. Based on the results of all the evaluations, recommendations for modifications to both procedures were presented for consideration.

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code now suggests the use of fatigue crack growth analysis in the evaluation of indications found during in-service inspection of nuclear components. In this paper the role of crack growth analysis in the evaluation process is reviewed in some detail, and the background and philosophy of its implementation is discussed. Each of the steps in the crack growth analysis process is discussed in order to point out the assumptions possible and their implications. A detailed consideration of crack shape change during growth is presented, as well as a statement of guidelines for choosing a reference crack growth law. The ASME approach is compared with the approaches taken in other industries and, finally, a number of areas in need of further work are highlighted.

Fatigue crack growth experiments were conducted on low alloy steel submerged arc weld metals representative of those used in the fabrication of pressure vessels. Variations of wire composition, flux type, and postweld heat treatments were used to produce weld deposits of different microstructures and tensile and toughness properties. Fatigue crack growth experiments were conducted on 25.4-mm (1-in.) thick modified wedge opening loaded (WOL) specimens at room temperature and 288 °C (550°F). A compliance calibration curve was developed at room temperature and adjusted for use with elevated temperature compliance measurements. Results of these experiments indicate: (a) a slight effect of flux type on crack growth rates, (b) incomplete tempering or relief of residual welding stress, or both, has an appreciable effect on crack growth behavior, (c) comparable or lower crack growth rates at 288°C (550°F) compared to room temperature, (d) comparable or lower crack growth rates than those predicted in Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, and (e) variations in crack growth rates due to wire, flux, and postweld heat treatment were less pronounced than the variations in toughness and tensile properties.

One of the main concerns in nuclear power plant operation is protecting the reactor pressure vessel against unstable, brittle fracture. The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code and U.S. Nuclear Regulatory Commission (NRC) regulations prescribe methods for determining plant operating pressure-temperature curves and radiation embrittlement limits that define safety margins for both normal and accident conditions. These limits must be updated periodically using a surveillance testing program to reflect the condition of the steel vessel wall after many years of neutron irradiation. In older plants, these limits may impose significant operational restrictions. Surveillance data may also be useful in demonstrating that, in some plants, the effects of embrittlement are minimal and the life of the vessel can be extended while maintaining adequate margins of safety for plant operation. Decisions involving both life attainment and life extension place new demands on the surveillance program. In some cases, the new demands can be met by minor program adjustments; in other cases, major adjustments are needed to obtain the requisite data. This paper discusses various approaches for managing radiation embrittlement in support of life attainment and life extension decisions. Strategies such as integrated programs, replacement surveillance capsules, and supplemental dosimetry measurements are addressed.

The Advanced Reactor Technologies (ART) Program has established a multi-year plan to develop Alloy 709 advanced stainless steel (A709), generate the data package and develop material-specific design parameters in qualifying it as a new structural material for Class A component design in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, Division 5, High Temperature Reactors. In collaboration with material vendors, the Advanced Materials Development activities under ART have successfully scaled the A709 plate form production from a laboratory heat of 500 pounds to commercial heats totaling 133,000 pounds of A709 plate fabricated from three heats. The goal of the overall A709 development program is to establish the necessary microstructural and mechanical properties relationship for A709 to ultimately develop fabrication parameters for other product forms such as bars, piping and forging using the available ART A709 materials. The objective of this A709 development work at ORNL in FY2023 is to experimentally determine grain coarsening behavior for the A709 heats and to experimentally generate the continuous cooling precipitation (CCP) diagram for A709 using the as-rolled commercial heat plate materials. Integral to this work is the characterization of the as-rolled materials and the determination of an effective solution annealing process. This report summarizes the work performed to identify the grain coarsening temperature for commercial heat 58776-3RB fabricated by G. O. Carlson and heat 529900-02 fabricated by Allegheny Technologies Incorporated (ATI) Flat Rolled Products.

A study of buckling analyses of piping structures