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DOE OSTI2018

Finite element analysis of compliant cladding and base metal systems

This report describes a roadmap towards a design method for high temperature nuclear cladded components that does not require long term testing of clad materials. Molten Salt Reactors (MSRs) will operate using corrosive fuel or coolant salts. The existing Class A metallic materials qualified for ASME Boiler and Pressure Vessel Code Section III, Division 5 design are not optimal for corrosion resistance when exposed to MSR salts. Corrosion could then severely limit the service life of MSRs. One option to alleviate this could be to Code qualify new corrosion resistant materials. However, this is a lengthy and expensive process and MSR developers have a near term need for a design solution that enables reactor designs with long service lives. An alternative would be to use cladded components – overlay some existing Class A material with a thin layer of some corrosion resistant material. This report describes the outlines of a design strategy that would enable this approach. The strategy relies on approximate design analysis methods for certain types of clad/base systems that avoids requiring long-term mechanical properties for the clad materials. This would allow the use of cladded components in the near term, without requiring long testing programs. However, it will limit the types of clad materials that can be used. This report describes the design framework, demonstrates the effectiveness of the approximate analysis methods, and confirms the approach using full scale finite element analysis of prototypical cladded high temperature components. It also describes a roadmap towards a complete design method and surveys available clad materials and bonding methods for suitably, considering only mechanical and not environmental effects.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2018

Quarter 2 Research Performance Progress Report (RPPR) - DOE Solar Program

Creep-fatigue deformation is an important consideration for a thermal receiver of a gas phase (GP) Concentrating Solar Power (CSP) system due to the constant static stress or pressure, diurnal cycling, and elevated service temperatures required for efficient operation. An accurate description of the creep-fatigue behavior, not available for five of the six candidate materials, is important for assessment of preliminary designs. This project will provide a detailed analysis of the creep-fatigue behavior and damage accumulation of a candidate structural material for a CSP solar thermal receiver to address a critical knowledge barrier for receiver design in the GP pathway concept identified in the CSP Gen3 Demonstration Roadmap. This effort includes the development of rules for the design of solar receiver components against high temperature creep-fatigue and ratcheting failure modes. The ASME Code rules for high temperature nuclear components will form the basis of the method but adjustments will be made to reflect the generally shorter, diurnal operating cycles of thermal receivers and the relative consequences of failure, comparing nuclear to solar components. This is report covers the second quarter's work, which addressed outstanding questions on alloy selection for Quarter 1, and began laying the ground work for mechanical testing.

14 SOLAR ENERGY↗

DOE OSTI2018

Development of a mechanistic thermal aging model for Grade 91

This report provides a summary on understanding and predicting the effects of long-term thermal aging on microstructure and tensile properties of G91 to corroborate the ASME Code rules in strength reduction due to elevated temperature service. The research is to support the design and long-term operation of G91 structural components in sodium-cooled fast reactors (SFRs). The report is a Level 2 deliverable in FY18 (M2NT-18AN050502061), under the Work Package NT-18AN05050206, “G91 Code Extension Testing” performed by the Argonne National Laboratory (ANL), as part of the Advanced Materials Development Program for Fast Reactors. The thermal aging study of G91 involves three types of thermally-aged specimens: (1) specimens machined from two heats of G91, G91-H1 and H30176 that have been tested for the Advanced Materials Development Programs in the past years. These specimens were aged at 550, 600, and 650°C for times up to ~64,000 h; (2) specimens fabricated from the heads of the archived crept specimens (13 in total) of six heats of G91, H5349, H30394, H30383, H10148, H30176. These specimens were aged at 427 - 538°C and for times up to 132,647 h; (3) specimens made from a tube removed from the Kingston coal-fired power plant after exposure for 155,000 h at 550°C. These three sets of specimens provide a comprehensive data set for understanding and predicting the effect of long-term thermal aging on microstructure and tensile properties of G91. Microstructural analysis of the thermally-aged G91 revealed that thermal aging causes significant microstructural changes: (1) the tempered martensite formed during tempering is unstable, and suffers significant recovery during aging, manifested by an increase in subgrain width, a decrease in subgrain length, and a decrease in subgrain aspect ratio (i.e. subgrains become more equiaxed) as the aging temperature and time increase. The subgrain recovery is facilitated by the reduction of dislocation density within subgrains; (2) M23C6 carbides in the as-received G91 are distributed along grain and subgrain boundaries and play an important role as pinning obstacles against subgrain coarsening. M23C6 carbides coarsen during aging, and the coarsening of M23C6 carbides reduces the pinning force on the boundaries; (3) MX carbonitrides in the as-received G91 are distributed uniformly within subgrains, providing the precipitation hardening effect. They coarsen with increasing aging temperature and time, but the coarsening rate is low; (3) a new phase, the Laves phase intermetallic forms during aging. The formation of the Laves phase removes the Mo solutes from the matrix, reducing the solid solution strengthening effect. The Laves phase particles grow rapidly once they are nucleated, and they precipitate primarily along boundaries. They may have a pinning effect on boundaries, but should provide little to the alloy strength. The evolution of each microstructural constituent during thermal aging can be well described by individual microstructural models. The estimated activation energy for MX coarsening was 175 kJ/mol, equivalent to the activation energy for boundary and core diffusion in a-Fe, while the estimated activation energy for subgrain coarsening was significantly lower, ~96 kJ/mol. Thermal aging of G91 results in the reduction in the yield stress and the ultimate tensile strength but has an insignificant effect on the uniform and total elongations. The reduction in tensile strength is more pronounced as the aging temperature and time increases. The in situ tensile tests with high-energy X-rays revealed that sub-boundary strengthening plays a dominant role, and the strengthening role of M23C6 particles may be accounted for by subgrian boundary hardening, and three strengthening mechanisms can be superimposed to describe the tensile strength of G91: subgrain boundary hardening, MX precipitation hardening, and Mo solid solution strengthening. A microstructure – strength model was establ

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2018

NSRD-12, Novel Mini-Tubular HEPA Media for Nuclear Facility Ventilation Systems

The overall goal is to improve safety of DOE nuclear facilities during fire scenarios by developing improved filtration technology, increasing performance while reducing pressure drop. Ceramic filtration technology can provide robust, passive mitigation against radiological releases.The primary objective of this report is to demonstrate that mini-tubular ceramic (MTC) media can reduce pressure drop across a filter compared to flow through a membrane. A secondary objective is to obtain a baseline filtration efficiency measurement. Reducing pressure drop across a filter through MTC design can be combined with nanofiber technology to improve filter performance. Developing MTC media can also solve processing challenges associated with manufacturing ceramic nanofiber membranes and coatings by electrospinning. This work will support developing specifications for industry codes and standards such as ASME AG-1, Subsection FO, Ceramic Filters.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

Gear Test Assembly - Status Report FY2018

The gear test assembly completed testing in air and distilled water successfully under a range of loading conditions. An external torque sensor assembly was designed and added to confirm the torque readings provided by the Parker control system. Hand calculations were performed to determine the time range that the heaters could be turned on without exceeding the ASME Boiler and Pressure Vessel code Fatigue Stress Limits for 1,000,000 cycles. Transient computational fluid dynamics analyses were performed to determine the heater cycle on/off time. Steady state computational fluid dynamics analyses were performed to determine the final temperature distribution throughout the insulated assembly using a thinner but more effective insulation configuration. The initial heating tests were completed (without insulation installed) to check the heater control system. Testing of the argon pressurized seal chambers were performed and indicated the need for re-machining of several components to eliminate some leaks. Nondestructive examination of the gears continued successfully using eddy current and ultrasonic inspection techniques.

42 ENGINEERING↗

DOE OSTI2018

A Deterministic and Probabilistic Framework for Forecasting of Time-Series Damage States and Associated End-of-lIfe of a Pressurizer Water Reactor Surge Line under Design-Basis and Grid-Load-Following Loading Conditions

Argonne National Laboratory in support of the Department of Energy’s Light Water Reactor Sustainability (LWRS) program. The present semi-annual report covers the period between May 2018 and September 2018. In this report, we present the work performed to further improve the capability for structural integrity prediction. Part of this work focused on validating a computational fluid dynamics model for modeling thermal stratification. We also present finite element (FE) model based thermal-mechanical stress analysis of a pressurized water reactor (PWR) surge line under grid-load-following loading condition. In addition, we present work related to end-of-life prediction of a PWR surge line based on ASME code and NUREG-6909 deterministic approaches under design-basis and grid-load-following conditions. Our research in this report period also majorly involved with the development of a strategy for probabilistic fatigue life estimation. First, we applied a Weibull probabilistic modeling approach based on end-of-life fatigue data (traditional S~N data under in-air and PWR water conditions) to estimate the probabilistic life for a given strain amplitude. Second, we developed a data analytics based Markov-Chain-Monte-Carlo (MCMC) model for probabilistic modeling of time-series fatigue damage states and the associated probabilistic end-of-life of a component. Unlike the Weibull type approach, the MCMC type may not require a large number of fatigue tests for estimating the end-of-life for given loading and environmental conditions. We demonstrated the use of this framework through symmetric fatigue tests (stress- or strain-ontrolled tests either under in-air or PWR-water conditions) obtained from our earlier experiments. The MCMC model can use base time-series data either obtained from a model or fatigue experiment. Since experimental data are typically more accurate than model-calculated data, we used experiment-based time-series data to demonstrate the MCMC framework. However, when performing a complex experiment is not possible (due to time and costs involved), a model-based approach can be used to generate the time-series data. Similar model-based data are presented in our earlier report (e.g. for cyclic hardening damage states of a PWR surge line under constant amplitude and design-basis loading [1]). However, for the MCMC work in this report, we used experiment-based time-series damage states obtained from fatigue experiments under design-basis and grid-load-following loading conditions. The experimental loading inputs were based on thermal-mechanical FE model results of a typical PWR surge line. Then, we used the experimental time-series damage state data for forecasting the probabilistic time-series damage states and end-of-life of the PWR surge line. Based on this MCMC modeling strategy, and with the assumed loading, environment and thermal-mechanical boundary conditions, we have estimated that a 316SS-pure base metal PWR surge line would survive at least 159 years with zero failure probability.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2019

Thermal Hydraulic Experimental Test Article: Status Report for FY2019

The Thermal Hydraulic Experimental Test Article (THETA) is a facility that will be used to develop sodium components and instrumentation as well as acquire experimental data for validation of reactor thermal hydraulic and safety analysis codes. The facility will simulate nominal conditions as well as protected/unprotected loss of flow accidents in an SFR. High fidelity distributed temperature profiles of the developed flow field will be acquired with Rayleigh backscatter based optical fiber temperature sensors. The facility is being designed in partnership with systems code experts to tailor the experiment to ensure the most relevant and highest quality data for code validation. Figure 1 illustrates the main components of the primary system. THETA will be installed in the Mechanisms Engineering Test Loop (METL) with the primary system in the 28 inch Test Vessel #4, the secondary system in the 18-inch Test Vessel #6, and its sodium-to-air heat exchanger will occupy the location for 18-inch Test Vessel #5 (Test Vessel #5 will not be installed), Figure 2. The design work for the facility has been completed and is currently in the procurement, manufacturing and initial testing stage. The THETA primary system, including the redan, inlet and outlet plenum, 38 kW electric core simulant heater, submerged flowmeter, pump, and all instrumentation has either been procured or is currently being tested at Argonne. The secondary sodium system for cooling the tube-side sodium of the intermediate heat exchanger has all been analyzed according to ASME code and is in the final development stages before procurement.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2019

Creep-Fatigue Behavior and Damage Accumulation of a Candidate Structural Material for Concentrating Solar Power Thermal Receiver Quarter 6 Report

Creep-fatigue deformation is an important consideration for a thermal receiver in Concentrating Solar Power (CSP) systems due to the constant static stress or pressure, diurnal cycling, and elevated service temperatures required for efficient operation. An accurate description of the creep-fatigue behavior, not available for five of the six candidate materials, is important for assessment of preliminary designs. This project will provide a detailed analysis of the creep-fatigue behavior and damage accumulation of a candidate structural material for a CSP solar thermal receiver to address a critical knowledge barrier for receiver designs identified in the CSP Gen3 Demonstration Roadmap. This effort includes the development of rules for the design of solar receiver components against high temperature creep-fatigue and ratcheting failure modes. The ASME Code rules for high temperature nuclear components will form the basis of the method but adjustments will be made to reflect the generally shorter, diurnal operating cycles of thermal receivers and the relative consequences of failure, comparing nuclear to solar components.

14 SOLAR ENERGY↗

DOE OSTI2019

Planned FY19 Creep and Fatigue Design Curve Testing of Alloy 709 Base Metal

This report summarizes the status of planned FY 2019 creep and fatigue testing of the first commercial heat of Alloy 709 base metal at Oak Ridge National Laboratory (ORNL). A comprehensive master creep testing matrix was generated to support the ASME code qualification of Alloy 709. The testing activities and research are carried out at Argonne National Laboratory, Idaho National Laboratory and ORNL. ORNL was tasked to carry out a subset of the creep Code Case testing matrix for electroslag remelt (ESR) with solution annealing (S/A) at 1100°C in FY 2019. In this reporting cycle, 25 creep tests were started and 9 of them produced rupture data. The remaining 16 tests of ESR with S/A at 1100°C, along with 6 long-term creep tests of argon-oxygen-decarburization with S/A at 1100°C, are ongoing.

36 MATERIALS SCIENCE↗

DOE OSTI2019

Crack Growth Rate and Large Plate Demonstration of Chloride-induced Stress Corrosion Cracking in Spent Nuclear Fuel Storage Canisters

This report describes the body of work involving testing and analyses performed in FY18 and FY19 at the Savannah River National Laboratory and at Korea University to determine chloride-induced stress corrosion crack (CISCC) growth rates in weldments of austenitic stainless steel plate material. The primary work involved: several separate crack growth rate test campaigns with new testing setups using laboratory-scale corrosion specimens; analyses to estimate the stress intensity factors for flaws postulated to be present in the residual stress fields of axial and circumferential weldments in a prototypic dry storage canister; and establishment of a “large plate” (approximately 51×46 cm with thickness 16 mm and weight about 30 kg) CISCC test that uses a plate with a circumferential weldment harvested from the Sandia National Laboratories’ full-size mockup dry storage canister. The tests for crack growth rate were conducted on ASTM E1861 bolt-load compact tension specimens in a setup designed to allow initially dried salt deposits to deliquesce and infuse the brine to the crack front under conditions relevant to the canister storage environments (e.g., temperature and humidity). The crack growth rate test results matched previous experimental data in the open literature. These literature data had been compiled by SNL and EPRI to develop a temperature-dependent (K-insensitive) crack growth rate model to support flaw disposition in the ASME Boiler and Pressure Vessel Section XI Code Case N-860, “Examination Requirements and Acceptance Standards for Spent Nuclear Fuel Storage and Transportation Containment Systems.” The knowledge of the salt, temperature, humidity conditions for CISCC including the threshold stress intensity factor, $K_{ISCC}$, were used to setup the large plate test in which a set of partthrough- wall and through-wall electrical discharge machined (EDM) defects were positioned in the weld residual stress field of the welded plate material from the mockup canister. A salt solution was applied to this plate specimen with an air-brush spray, then dried, and then exposed to a constant 75% RH condition at an ambient room temperature of 22 °C. The salt loading and environmental condition is expected to be aggressive to cause CISCC. The visual results at the surface of the plate show corrosion product staining (brown rust appearance) but with no CISCC emanating from the machined notches after approximately 5 months exposure to date. The plate test will continue in FY20 and phased-array ultrasonic testing and other NDE techniques will be applied to identify and characterize CISCC that may emanate from the machined defects. The notches and any CISCC from them will be sectioned at the conclusion of the test to verify NDE characterization, and to develop effective through-wall crack growth rates from cracks that would likely be branched.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Thermal Hydraulic Experimental Test Article - Status Report for FY2019 (Rev. 1)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2020

Design Guidance for High Temperature Concentrating Solar Power Components

This report provides guidance for the design of components for concentrating power facilities operating at high temperatures and undergoing high, secondary thermal stresses relative to the applied primary pressure stress. The design rules were developed for the design of Generation 3 CSP tubular receivers manufactured from Alloy 740H, but are generally applicable to a wide range of component types undergoing similar loads. Part 1 of the report provides procedural design rules for components to be used in conjunction with the 2019 edition of the ASME Boiler & Pressure Vessel Code. Part 2 provides corresponding design data for Alloy 740H. Part 3 of the report is a commentary describing the rationale behind the design rules and the data underlying the design material properties. Finally, Part 4 provides an extensive set of worked sample problems detailing the application of the rules to CSP components.

24 POWER TRANSMISSION AND DISTRIBUTION↗

DOE OSTI2020

V28 Test Report

The V28 containment vessel was procured by the US Army Recovered Chemical Material Directorate (RCMD) as a replacement vessel for use on the P2 Explosive Destruction Systems. It is the fourth EDS vessel to be fabricated under Code Case 2564 of the ASME Boiler and Pressure Vessel Code, which provides rules for the design of impulsively loaded vessels. The explosive rating for the vessel, based on the Code Case, is nine (9) pounds TNT- equivalent for up to 637 detonations. This report documents the results of explosive tests that were done on the vessel at Sandia National Laboratories in Albuquerque New Mexico to qualify the vessel for explosive use. The primary qualification test consisted of six, 1.5 pounds charges of Composition C-4 (equivalent to 11.25 pounds TNT) distributed around the vessel in accordance with the User Design Specification. This test was repeated due to a lack of proper clamp settings. Two additional tests using less explosive were performed, one identical in configuration to a test performed in the V27 qualification series as a baseline for comparison, and one where the separation distance of the charges was increased to extend the V27 analysis of distributed load effects on the P2 vessel. All vessel acceptance criteria were met.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗

DOE OSTI2019

Development of a multiaxial deformation measure and creep-fatigue damage summation for multiple load cycle types in support of an improved creep-fatigue design method

This report describes methods for extending the elastic perfectly plastic (EPP) simplified model test (SMT) approach to cover multiaxial load conditions and methods to combine EPP+SMT use fractions for multiple loading cycles. The completion of this work, in conjunction with companion reports on constructing and validating SMT design curves, marks the development of a complete methodology for assessing creep-fatigue damage in high temperature nuclear reactor structural components. This EPP+SMT method aims to better account for elastic follow up effects, simplify the design process, and reduce overconservatism when compared to current simplified bounding approaches in the ASME Boiler and Pressure Vessel Code Section III, Division 5, Subsection HB, Subpart B covering high temperature nuclear reactors.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Assessment of the Material Performance for TCR Applications (Progress Report)

The objective of Argonne’s materials research activities for the Transformational Challenge Reactor (TCR) program is to improve the understanding of material properties from additive manufacturing with the focus on understanding the creep and fatigue properties of additively manufactured materials. The material work conducted by Argonne provides support in developing and qualifying advanced materials and manufacturing processes to allow for innovative reactor design and licensing for the TCR. In FY 2020, Oak Ridge National Laboratory (ORNL) provided six ASTM standard sized creep specimens of additively-manufactured 316L stainless steel (AM 316L SS) for evaluation of the effect of post-build heat treatment on the creep property. Two different teat treatments were conducted on the as-built specimens, namely 1050°C/1h and 650°C/1h before creep testing. All specimens were creep tested at 550°C and 275 MPa to understand the effect of heat treatments. We have also received AM 316L SS printed in various forms including rods and sheets, wrought 316 L SS (as a reference material), and AM IN718 for tensile, creep and fatigue testing. Both ASTM standard sized specimens and subsized specimens are being fabricated or tested to contribute to a comprehensive mechanical property database of AM materials and to provide location-specific property data to the TCR digital platform. Mechanical testing at Argonne is conducted to ASME NQA-1 or its lab equivalent for quality assurance. This report presents the results of the creep tests at 550°C and microstructural characterization for the assessment of the effect of the post-build heat treatment on the creep behavior of AM 316L SS. It also discusses ongoing efforts on obtaining location-dependent properties of AM 316L SS.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Initial development of a high temperature life prediction method directly accounting for variability in material properties

This report presents the initial development of a life prediction method that accounts for the variability of the material properties in Grade 91 steel. We account for material variability by both fitting a variable 3-parameter Weibull distribution to the rupture data of the material and by quantifying the effect that the creep deformation properties have on stresses via a Monte Carlo approach. To ensure a reasonable time frame when applying Monte Carlo to the creep stress, we propose a methodology that models the material as an extremely viscous Stokes flow with a non-Newtonian viscosity, therefore solving the problem in a single step. We use the final probabilistic model for structural failure developed in this document to evaluate a Flat Head Vessel designed with the Section III, Division 5 ASME Boiler Pressure and Vessel Code rules with a 100,000h design life and assess its probability of premature failure.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2020

Initial development for the introduction of provisions for irradiation effects in Division 5 rules

Heat pipe microreactor concept involves the use of a solid core block to encapsulate the nuclear fuel and the passive extraction of heat from the core block to the heat exchanger block by advanced heat pipes. The core block is a key component of the heat pipe microreactor design. It integrates the functions of reactor vessel, core structural components and fuel cladding of traditional Gen IV advanced reactor designs into one single structure. The effects of irradiation on the structural failure modes are outside the scope of the rules of the ASME Boiler and Pressure Vessel Code, Section III, Division 5, Subsection HB, Subpart B for metallic Class A components. Methods to account for the irradiation effect were investigated in order to develop design methods that can be applied to the core block design in the heat pipe microreactor. This report presents an initial development of an approach to account for the irradiation effects in the design rules.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2020

Thermal Hydraulic Experimental Test Article (FY2020 Status Report)

The Thermal Hydraulic Experimental Test Article (THETA) is a facility that will be used to develop sodium components and instrumentation as well as acquire experimental data for validation of reactor thermal hydraulic and safety analysis codes. The facility will simulate nominal conditions as well as protected/unprotected loss of flow accidents in a sodium-cooled fast reactor (SFR). High fidelity distributed temperature profiles of the developed flow field will be acquired with Rayleigh backscatter based optical fiber temperature sensors. The facility is being designed in partnership with systems code experts to tailor the experiment to ensure the most relevant and highest quality data for code validation. THETA is comprised of a traditional primary and secondary system. The primary system is submerged in the pool of sodium and consists of a pump, electrically heated core, intermediate heat exchanger, and connected piping and thermal barriers (redan). The secondary system, located outside of the sodium pool, consists of a pump, sodium to air heat exchanger, and connected piping and valves. Figure 1 illustrates the main components of the primary system. THETA will be installed in the Mechanisms Engineering Test Loop (METL) with the primary system in the 28 inch Test Vessel #4, Figure 2. The design work for the facility has been completed. Since the FY19 THETA status report, the THETA primary system components have been assembled and tested using de-ionized water. Some results of this testing at select Richardson and Reynolds numbers are included herein. The custom THETA sodium flowmeter (permanent magnet submersible sodium flowmeter) has been built and calibrated using sodium in an auxiliary facility against a NIST traceable vortex shedder type flowmeter. The permanent magnet flowmeter showed excellent performance, following the theoretical equation for measured voltage quite well, with an error < 3.5% at flow rates ranging from 3-24 GPM at temperatures up to 400 °C. The secondary sodium system for heat rejection from intermediate heat exchanger has been analyzed according to ASME code. All secondary side piping, fittings and a U-stamped air-to-sodium heat exchanger have been procured.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗