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DOE OSTI2016

Fast automatic 3D liver segmentation based on a three-level AdaBoost-guided active shape model

Purpose: A robust, automatic, and rapid method for liver delineation is urgently needed for the diagnosis and treatment of liver disorders. Until now, the high variability in liver shape, local image artifacts, and the presence of tumors have complicated the development of automatic 3D liver segmentation. In this study, an automatic three-level AdaBoost-guided active shape model (ASM) is proposed for the segmentation of the liver based on enhanced computed tomography images in a robust and fast manner, with an emphasis on the detection of tumors. Methods: The AdaBoost voxel classifier and AdaBoost profile classifier were used to automatically guide three-level active shape modeling. In the first level of model initialization, fast automatic liver segmentation by an AdaBoost voxel classifier method is proposed. A shape model is then initialized by registration with the resulting rough segmentation. In the second level of active shape model fitting, a prior model based on the two-class AdaBoost profile classifier is proposed to identify the optimal surface. In the third level, a deformable simplex mesh with profile probability and curvature constraint as the external force is used to refine the shape fitting result. In total, three registration methods—3D similarity registration, probability atlas B-spline, and their proposed deformable closest point registration—are used to establish shape correspondence. Results: The proposed method was evaluated using three public challenge datasets: 3Dircadb1, SLIVER07, and Visceral Anatomy3. The results showed that our approach performs with promising efficiency, with an average of 35 s, and accuracy, with an average Dice similarity coefficient (DSC) of 0.94 ± 0.02, 0.96 ± 0.01, and 0.94 ± 0.02 for the 3Dircadb1, SLIVER07, and Anatomy3 training datasets, respectively. The DSC of the SLIVER07 testing and Anatomy3 unseen testing datasets were 0.964 and 0.933, respectively. Conclusions: The proposed automatic approach achieves robust, accurate, and fast liver segmentation for 3D CTce datasets. The AdaBoost voxel classifier can detect liver area quickly without errors and provides sufficient liver shape information for model initialization. The AdaBoost profile classifier achieves sufficient accuracy and greatly decreases segmentation time. These results show that the proposed segmentation method achieves a level of accuracy comparable to that of state-of-the-art automatic methods based on ASM.

60 APPLIED LIFE SCIENCES↗

DOE OSTI2022

Design and execution of a verification, validation, and uncertainty quantification plan for a numerical model of left ventricular flow after LVAD implantation

Left ventricular assist devices (LVADs) are implantable pumps that act as a life support therapy for patients with severe heart failure. Despite improving the survival rate, LVAD therapy can carry major complications. Particularly, the flow distortion introduced by the LVAD in the left ventricle (LV) may induce thrombus formation. While previous works have used numerical models to study the impact of multiple variables in the intra-LV stagnation regions, a comprehensive validation analysis has never been executed. The main goal of this work is to present a model of the LV-LVAD system and to design and follow a verification, validation and uncertainty quantification (VVUQ) plan based on the ASME V&V40 and V&V20 standards to ensure credible predictions. The experiment used to validate the simulation is the SDSU cardiac simulator, a bench mock-up of the cardiovascular system that allows mimicking multiple operation conditions for the heart-LVAD system. The numerical model is based on Alya, the BSC’s in-house platform for numerical modelling. Alya solves the Navier-Stokes equation with an Arbitrary Lagrangian-Eulerian (ALE) formulation in a deformable ventricle and includes pressure-driven valves, a 0D Windkessel model for the arterial output and a LVAD boundary condition modeled through a dynamic pressure-flow performance curve. The designed VVUQ plan involves: (a) a risk analysis and the associated credibility goals; (b) a verification stage to ensure correctness in the numerical solution procedure; (c) a sensitivity analysis to quantify the impact of the inputs on the four quantities of interest (QoIs) (average aortic root flow $Q^{avg}_{Ao}$, maximum aortic root flow $Q^{max}_{Ao}$, average LVAD flow $Q^{avg}_{VAD}$, and maximum LVAD flow $Q^{max}_{VAD}$; (d) an uncertainty quantification using six validation experiments that include extreme operating conditions. Numerical code verification tests ensured correctness of the solution procedure and numerical calculation verification showed a grid convergence index (GCI)95% <3.3%. The total Sobol indices obtained during the sensitivity analysis demonstrated that the ejection fraction, the heart rate, and the pump performance curve coefficients are the most impactful inputs for the analysed QoIs. The Minkowski norm is used as validation metric for the uncertainty quantification. It shows that the midpoint cases have more accurate results when compared to the extreme cases. The total computational cost of the simulations was above 100 [core-years] executed in around three weeks time span in Marenostrum IV supercomputer. This work details a novel numerical model for the LV-LVAD system, that is supported by the design and execution of a VVUQ plan created following recognised international standards. We present a methodology demonstrating that stringent VVUQ according to ASME standards is feasible but computationally expensive.

59 BASIC BIOLOGICAL SCIENCES↗

DOE OSTI2014

Progress Report on Alloy 617 Isochronous Stress-Strain Curves

Isochronous stress-strain curves for Alloy 617 up to a temperature of 1000°C will be required to qualify the material for elevated temperature design in Section III, Division 1, Subsection NH of the ASME Boiler and Pressure Vessel Code. Several potential methods for developing these curves are reviewed in this report. It is shown that in general power-law creep is the rate controlling deformation mechanism for a wide range of alloy heats, test temperatures and stresses. Measurement of the strain rate sensitivity of Alloy 617 indicates that the material is highly strain rate sensitive in the tensile deformation range above about 750°C. This suggests that the concept of a hot tensile curve as a bounding case on the isochronous stress-strain diagrams is problematic. The impact of strain rate on the hot tensile curves is examined and it is concluded that incorporating such a curve is only meaningful if a single tensile strain rate (typically the ASTM standard rate of 0.5%/min) is arbitrarily defined. Current experimentally determined creep data are compared to isochronous stress-strain curves proposed previously by the German programs in the 1980s and by the 1990 draft ASME Code Case. Variability in how well the experimental data are represented by the proposed design curves that suggests further analysis is necessary prior to completing a new draft Code Case.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

Evaluation on the Feasibility of Using Ultrasonic Testing of Reactor Pressure Vessel Welds for Assessing Flaw Density/Distribution per 10 CFR 50.61a, Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock

This technical letter report provides the status of an assessment undertaken by PNNL at the request of the NRC to verify the capability of periodic ASME-required volumetric examinations of reactor vessels to characterize the density and distribution of flaws of interest for applying §50.61a on a plant-by-plant basis. The PTS rule, described in the Code of Federal Regulations, Title 10, Section 50.61 (§50.61), "Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events," establishes screening criteria to ensure that the potential for a reactor vessel to fail due to a PTS event is deemed to be acceptably low. Recently, the NRC completed a research program that concluded that the risk of through-wall cracking due to a PTS event is much lower than previously estimated. The NRC subsequently developed and promulgated an alternate PTS rule, §50.61a, that can be implemented by PWR licensees. The §50.61a rule differs from §50.61 in that it requires licensees who choose to follow this alternate method to analyze the results from periodic volumetric examinations required by the ASME Code, Section XI, Rules for Inservice Inspection (ISI) of Nuclear Power Plants.

alternate PTS rule↗

DOE OSTI2014

Creep of A508/533 Pressure Vessel Steel

ABSTRACT 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.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Multiaxial Creep-Fatigue and Creep-Ratcheting Failures of Grade 91 and Haynes 230 Alloys Toward Addressing Design Issues of Gen IV Nuclear Power Plants

The proposed research will develop systematic sets of uniaxial and multiaxial experimental data at a very high temperature (850-950°C) for Alloy 617. The loading histories to be prescribed in the experiments will induce creep-fatigue and creep-ratcheting failure mechanisms. These experimental responses will be scrutinized in order to quantify the influences of temperature and creep on fatigue and ratcheting failures. A unified constitutive model (UCM) will be developed and validated against these experimental responses. The improved UCM will be incorporated into the widely used finite element commercial software packages ANSYS. The modified ANSYS will be validated so that it can be used for evaluating the very high temperature ASME-NH design-by-analysis methodology for Alloy 617 and thereby addressing the ASME-NH design code issues.

36 MATERIALS SCIENCE↗

DOE OSTI2015

Report on FY15 Two-Bar Thermal Ratcheting Test Results

Alloy 617 is a reference structural material for very high temperature components of advanced-gas cooled reactors with outlet temperatures in the range of 900-950°C. In order for designers to be able to use Alloy 617 for these high temperature components, Alloy 617 has to be approved for use in Section III (the nuclear section) of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. A plan has been developed to submit a draft code for Alloy 617 to ASME Section III by 2015. However, the current rules in Subsection NH* for the evaluation of strain limits and creep-fatigue damage using simplified methods based on elastic analysis have been deemed inappropriate for Alloy 617 at temperatures above 1200°F(650°C). The rationale for this exclusion is that at higher temperatures it is not feasible to decouple plasticity and creep deformation, which is the basis for the current simplified rules. This temperature, 1200°F(650°C), is well below the temperature range of interest for this material in High Temperature Gas Cooled Reactor (HTGR) applications. The only current alternative is, thus, a full inelastic analysis which requires sophisticated material models which have been formulated but not yet verified. To address this issue, proposed code rules have been developed which are based on the use of elastic-perfectly plastic (EPP) analysis methods and which are expected to be applicable to very high temperatures. These newly proposed rules also address a long-term objective to provide an option for more simple, comprehensive and easily applied rules than the current so called “simplified” rules.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2014

Application of Combined Sustained and Cyclic Loading Test Results to Alloy 617 Elevated Temperature Design Criteria

Alloy 617 is a reference structural material for very high temperature components of advanced-gas cooled reactors with outlet temperatures in the range of 900-950°C . In order for designers to be able to use Alloy 617 for these high temperature components, Alloy 617 has to be approved for use in Section III (the nuclear section) of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. A plan has been developed to submit a draft code for Alloy 617 to ASME Section III by 2015. However, the current rules in Subsection NH for the evaluation of strain limits and creep-fatigue damage using simplified methods based on elastic analysis have been deemed inappropriate for Alloy 617 at temperatures above 1200°F (650°C). The rationale for this exclusion is that at higher temperatures it is not feasible to decouple plasticity and creep deformation, which is the basis for the current simplified rules. This temperature, 1200 °F, is well below the temperature range of interest for this material in High Temperature Gas Cooled Reactor (HTGR) applications. The only current alternative is, thus, a full inelastic analysis which requires sophisticated material models which have been formulated but not yet verified. To address this issue, proposed code rules have been developed which are based on the use of elastic-perfectly plastic (EPP) analysis methods and which are expected to be applicable to very high temperatures.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2015

Simulation of Internal Combustion Engines with High-Performance Computing Tools

Traditional Lagrangian spray modeling approaches for internal combustion engines are highly griddependent due to insufficient resolution in the near nozzle region. This is primarily because of inherent restrictions of volume fraction with the Lagrangian assumption together with high computational costs associated with small grid sizes. A state-of-the-art grid-convergent spray modeling approach was developed and implemented by Senecal et al. (ASME-ICEF2012-92043) in the CONVERGE software. The key features of the methodology include Adaptive Mesh Refinement (AMR), advanced liquid-gas momentum coupling, and improved distribution of the liquid phase, which enables use of cell sizes smaller than the nozzle diameter. This modeling approach was rigorously validated against nonevaporating, evaporating, and reacting data from the literature. The current numerical study focusses on further demonstration of the grid-convergent modeling approach for simulating a single-cylinder Cat® compression ignition engine. The simulated injector is characterized with a nominal nozzle exit diameter of 259 μm. Simulations using various minimum grid sizes (ranging from 125 μm to 1000 μm) are compared for engine performance and emissions parameters of interest such as pressure, heat release rate, ignition delay, NOx, HC, and soot emissions. The peak cell-count for the highest resolution simulation was on the order of 34 million. These computationally expensive simulations were facilitated at a high-performance computing facility at Argonne National Laboratory. METIS load-balancing algorithm was developed and implemented in Converge code for the simulations. Scaling studies were also performed. The validity of previously recommended grid settings (ASME-ICEF2012-92043) for accuracy/runtime trade-off is further assessed. Efficacy of a simplified combustion model is also compared against a detailed chemical kinetics based combustion model.

33 ADVANCED PROPULSION SYSTEMS↗

DOE OSTI2016

Progress Report on Alloy 617 Notched Specimen Testing

Creep behavior of Alloy 617 has been extensively characterized to support the development of a draft Code Case to qualify Alloy 617 in Section III division 5 of the ASME Boiler and Pressure Vessel Code. This will allow use of Alloy 617 in construction of nuclear reactor components at elevated temperatures and longer periods of time (up to 950°C and 100,000 hours). Prior to actual use, additional concerns not considered in the ASME code need to be addressed. Code Cases are based largely on uniaxial testing of smooth gage specimens. In service conditions, components will generally be under multi axial loading. There is also the concern of the behavior at discontinuities, such as threaded components. To address the concerns of multi axial creep behavior and at geometric discontinuities, notched specimens have been designed to create conditions representative of the states that service components experience. Two general notch geometries have been used for these series of tests: U notch and V notch specimens. The notches produce a tri axial stress state, though not uniform across the specimen. Characterization of the creep behavior of the U notch specimens and the creep rupture behavior of the V notch specimens provides a good approximation of the behavior expected of actual components. Preliminary testing and analysis have been completed and are reported in this document. This includes results from V notch specimens tested at 900°C and 800°C. Failure occurred in the smooth gage section of the specimen rather than at the root of the notch, though some damage was present at the root of the notch, where initial stress was highest. This indicates notch strengthening behavior in this material at these temperatures.

36 MATERIALS SCIENCE↗

DOE OSTI2017

High temperature ceramic heat exchanger for solid oxide fuel cell

A ceramic heat exchanger with high effectiveness and low-pressure drop has been developed for application as a cathode air preheater for a Solid Oxide Fuel Cell (SOFC). At the operating conditions of SOFCs, typical metallic alloys as those used in commercial heat exchangers may undergo chromium volatilization, which is a known cathode degradation mechanism that reduces SOFC performance and life. Use of ceramics such as alumina or alumina-silicate instead of chromium-containing metal is one approach to eliminate the effects of chromium on the SOFC cathode. This project leverages the geometric design of a heat exchanger previously prototyped and tested [1] to fabricate a novel heat exchanger made from a ceramic material. All calculations of heat exchanger performance requirements are based on state of the art SOFC operating conditions. The alumina-silicate material selected to fabricate the modular prototype heat exchanger was made via a heat transfer based trade-off analysis that shows that for low pressure-drop devices, flow convection is the limiting heat transfer mechanism, and therefore wall conductivity becomes less relevant to heat transfer effectiveness. The modular design presented allows for incremental aggregation of modules to target a broad range of operating conditions typical of present and upcoming SOFC applications (e.g., 25 to 400 kWe). The completed prototype was subjected to a series of static leak/pressure tests, dynamic pressure drop tests both at ambient and realistic service temperature conditions, and finally to a matrix of heat transfer experiments to fully characterize its heat transfer effectiveness (including tests at hot gas temperatures up to 1033 K or 1400oF). The stand-alone system was demonstrated in the laboratory to successfully satisfy the design criteria at realistic operating conditions, showing a total pressure drop at the design flow of 8 g/sec of 6 kPa (vs. 3.45 kPa approximate requirement). It was also shown to have heat transfer effectiveness of 76% (with 73% being the design requirement). Moreover, the unit was shown to operate leak-free, and the ceramic elements and seals were capable of withstanding the thermal cycling resulting from the testing procedure. The heat exchanger was then integrated into a modified fuel cell test stand and tested at steady state operation. Unfortunately, the level of performance achieved was lower than expected based on laboratory performance tests. This appears to have been the result of an unforeseen loosening of a heat exchanger structural component, which resulted into a leak that prevented proper supply of hot gas to the heat exchanger. However, in general terms, the heat exchanger/fuel cell integration process provides confidence that a future application of the unit with a SOFC is feasible, assuming structural issues are addressed. It is considered that the difficulties encountered at the last stage of the program provided invaluable input in locating shortcomings of the mechanical design and integration approach. A portion of this work was presented at the ASME Power & Energy Conference, in Charlotte, NC, on June 28th 2016 in a paper entitled “Development of a Ceramic Heat Exchanger for Application as Solid Oxide Fuel Cell Cathode Air Preheater” and the corresponding paper was published in the Proceedings of the ASME 2016 Power & Energy Conference, Paper No.: PowerEnergy2016-59333 [8].

20 FOSSIL-FUELED POWER PLANTS↗

DOE OSTI2017

FY17 Status Report on the Initial EPP Finite Element Analysis of Grade 91 Steel

This report describes a modification to the elastic-perfectly plastic (EPP) strain limits design method to account for cyclic softening in Gr. 91 steel. The report demonstrates that the unmodified EPP strain limits method described in current ASME code case is not conservative for materials with substantial cyclic softening behavior like Gr. 91 steel. However, the EPP strain limits method can be modified to be conservative for softening materials by using softened isochronous stress-strain curves in place of the standard curves developed from unsoftened creep experiments. The report provides softened curves derived from inelastic material simulations and factors describing the transformation of unsoftened curves to a softened state. Furthermore, the report outlines a method for deriving these factors directly from creep/fatigue tests. If the material softening saturates the proposed EPP strain limits method can be further simplified, providing a methodology based on temperature-dependent softening factors that could be implemented in an ASME code case allowing the use of the EPP strain limits method with Gr. 91. Finally, the report demonstrates the conservatism of the modified method when applied to inelastic simulation results and two bar experiments.

36 MATERIALS SCIENCE↗

DOE OSTI2018

Final Report on CFD and Thermal-Mechanical Stress Analysis of PWR Surge Line under Transient Condition Thermal Stratification and an Evolutionary Cyclic Plasticity Based Transformative Fatigue Evaluation Approach without Using S~N Curve: Rev. 1

This report presents an evolutionary cyclic plasticity based transformative approach for fatigue evaluation of safety critical components. This approach is generic which can be used not only for the present generation light water reactor components (which is our focus) but also for advanced reactor components. The approach is based on the fundamental concept of time-evolution of progressive fatigue damage rather than the end-of-life data based conventional S~N curve approaches. Series of fatigue tests on 316 stainless steel uniaxial specimens were conducted under strain-controlled and stress-controlled cyclic loading in two different environment such as in-air at 300 °C and PWR coolant water condition at 300 °C. Material models are developed to capture the uniaxial test data and are subsequently programmed into commercially available ABAQUS code to transform the uniaxial material behavior to multiaxial loading domain. The developed evolutionary cyclic-plasticity approach is verified by both analytical and 3D-FE modeling of uniaxial fatigue test specimens. Results show that the developed material models can capture the time-dependence of material hardening and softening with great accuracy. In addition, results show that the material models not only can capture the material behavior under constant amplitude loading but also can capture the load-sequence effect under variable and random amplitude loading. The ANL developed approach is one of its kind, which shows first in the fatigue research community that fatigue evaluation of a component (in our case verified for laboratory scale fatigue test specimens with great accuracy) can be performed for its entire fatigue life, including the hardening and softening behavior, without using a conventional S~N curve based approach. Additionally results are presented related to fatigue evaluation of real reactor components such as of a pressurized water reactor (PWR) surge line (SL), under upset transient conditions. SLs of PWR experience complex thermal stratification due to the mixing of differential temperature flow from hot leg and pressurizer. A detailed parametric study of thermal stratification in SL is performed through both standalone computational fluid dynamics (CFD) analysis and fluid-structure-interaction based coupled CFD and heat transfer analysis. Results show that the mass flow rate in SL substantially affects the extent of its thermal stratification. The resulting thermal boundaries conditioned are used as inputs for thermal-mechanical stress analysis and component level fatigue evaluation. Fatigue evaluation of SL pipe under example loading conditions was performed using the above-mentioned evolutionary-plasticity based approach and ASME code (for in-air life) and NUREG-6909 (for PWR environment life) based approaches. Results show that the evolutionary plasticity based approach estimate substantially different life compared to the conventional ASME based approach. To show the entirety of the proposed evolutionary fatigue-modeling framework, this report includes the results presented in semi-annual March report (ANL/LWRS-17/01). Sections 2-5 of March report are directly included as Sections 2-5 in this report, without any significant changes. However, Sections 6-9 of this report present the work performed during March-September of 2017. Whereas, Sections 1 and 10 present the organization of this report and the summary and future work, respectively. This report (ANL/LWRS-17/03 Rev. 1) is a slightly revised version of the earlier published report (ANL/LWRS-17/03). In this revised version, some of the results and associated discussion have been corrected/updated.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2018

Initial assessment of metallurgical interaction of clad/base metal systems

Molten Salt Reactors (MSRs), both liquid fuel and sold fuel (e.g., the fluoride salt-cooled high temperature reactors, FHR’s) offer several advantages relative to conventional, water-cooled nuclear power systems. However, currently approved ASME structural materials have not been optimized for use in molten salt environments and demonstrating materials performance in molten salts is challenging. Toward the goal of enabling MSR designs, corrosion resistant claddings offer an attractive solution to help mitigate corrosion concerns and enable the use of approved ASME materials. This report assesses metals and alloys that may be effective as corrosion resistant cladding with select structural alloys. The structural alloys of interest are capable of use at high temperatures ≥600°C for times ≥100,000 hours. Potential cladding alloys were assessed based on their corrosion resistance in both chloride and fluoride based salts, physical properties, mechanical properties, nuclear performance, and resistance to environmental-assisted cracking. Additionally both thermodynamic and kinetic computational modeling were used to assess interactions between the base metal and the cladding. Results to date have identified three candidate alloy systems that help enable MSR designs; nickel, molybdenum, and tungsten based claddings. This report discusses pros and cons of each class of cladding and outlines future work to develop these base metal / clad systems into industrial products suitable for use in commercial nuclear power systems.

36 MATERIALS SCIENCE↗

DOE OSTI2019

The Degradation of Strength under Varying Oxidizing Conditions for Nuclear Graphite

Strength behavior of fine and medium grain size nuclear graphite grades are presented after being subjected to varying oxidation conditions. This study provides valuable information for design and eventual licensing considerations for the oxidation behavior of graphite components within a gas-cooled High Temperature Reactor design and the eventual structural integrity of graphite core components. These tests address the underlying oxidized strength of nuclear graphite components for oxidized mass losses up to and beyond the current limits recommended for graphite components in the ASME code (maximum mass loss = 10%) over a range of oxidation temperatures (550° C – 750° C). Preliminary results generally demonstrate that low temperature oxidation can result in higher strength reductions than oxidation at higher temperatures for similar mass loss levels. Since the graphite component strength is a critical parameter necessary to determine the core structural integrity, determining the actual residual strength after oxidation is essential for determining actual component oxidation limits within the ASME code.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

Initial Development of an Improved Creep-Fatigue Design Method that Avoids the Separate Evaluation of Creep and Fatigue Damage and Eliminates the Requirement for Stress Classification

This report describes a method for capturing the effect of strain range, temperature, hold time, and elastic follow up on the cyclic life of a material undergoing uniaxial load. Developing design correlations of this type requires methods for extrapolating experimental data as a function of hold time and follow up factor. Various extrapolation methods are developed and discussed here and recommendations are developed for the final approaches to be adopted for design use. The report also describes comparisons between the new creep-fatigue design method, described here, and current creep-fatigue design methods in the ASME Boiler and Pressure Vessel Code and associated nuclear Code Cases. These comparisons demonstrate several advantages of the new method: it is easier to execute and is less over conservative than many current approaches. This report, along with a companion on extending these results to multiaxial load and on combining load cycles, represents the accumulation of several years of work on the development of this new approach to creep-fatigue design, combining an elastic perfectly plastic analysis method with simplified model tests measuring the effect of elastic follow up on creep-fatigue cyclic life. The conclusions to this report describe a path forward to developing the required design data for high temperature advanced reactor structural materials and guiding the adoption of the new method in the ASME Boiler and Pressure Vessel Code and related standards.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Initial development and verification of a primary load design method based on elastic-perfectly plastic analysis

This report describes a new primary load design method to supplement the current design-by-elastic-analysis rules in Section III, Division 5, Subsection HB, Subpart B of the ASME Boiler and Pressure Vessel Code, covering the design and construction of Class A high temperature nuclear structural components. The main objectives for the new design method are to provide a procedure that can be applied to components with complicated geometries and to simplify the primary load design process by providing rules more compatible with modern finite element analysis. The report includes a complete set of design rules, presented as a draft ASME Code Case, as well as a commentary on the rules and a set of verification problems. The verification design problems demonstrate the new primary load design rules produce safe, efficient components when compared to the current Division 5 approach, while greatly simplifying the design analysis process.

42 ENGINEERING↗

DOE OSTI2020

Initial High Temperature Inelastic Constitutive Model for Alloy 617

This report describes progress on a Alloy 617 inelastic constitutive model slated for inclusion in the ASME Boiler & Pressure Vessel Code in Code Case N-898 “Use of Alloy 617 (UNS N06617) for Class A Elevated Temperature Construction, Section III, Division 5” as part of an appendix on inelastic modeling. This appendix will provide a reference material model for A617 for use in conjunction with the ASME design by inelastic analysis criteria. This report focuses on the implementation and evaluation of an existing Alloy 617 model developed to represent the data collected at Idaho National Laboratory (INL) and Oak Ridge National Laboratory (ORNL) as part of the Alloy 617 Code Case experimental campaign. While the model accurately captures the cyclic stress/strain behavior of Alloy 617, it fails to capture the long-term creep and lower temperature tensile data. In addition, the model form is complicated and has over 50 configurable parameters. As such, we recommend the development of a simplified model, incorporating key features of the ORNL model but reducing the model complexity.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗