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

et al. 2020

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The report provides the initial description of a new creep-fatigue design method for structural components in high temperature nuclear service. The new method is based on an integrated elastic-perfectly plastic (EPP) analysis and Simplified Model Test (SMT) approach that reduces over conservatism, improves the treatment of elastic follow up, and simplifies the design procedure, when compared with the current creep-fatigue design methods in ASME Boiler and Pressure Vessel Code. Developing the design charts for the EPP-SMT design method requires extrapolating SMT test data as a function of hold time and follow up factor. The report develops the preliminary design charts for Alloy 617 at temperatures between 800°C and 950°C by combining two extrapolation approaches developed in a previous work. The report also presents a comparative analysis between the EPP-SMT design method and the current ASME creep-fatigue design methods by evaluating design life of two sample geometries under different loading conditions. Results from the comparative analysis verify the EPP-SMT design charts but suggest the requirement of additional test data in the low strain range regime for improving the extrapolation procedure that will further reduce the over conservatism in the creep-fatigue damage evaluation. The report also concludes that the EPP-SMT design procedure can account for effect of primary load on creep-fatigue life by using a fixed, bounding value of follow up in constructing the design charts. The conclusions to this report describe the future work required to complete this new design method so it can be codified through a nuclear Code Case Preliminary description of a new creep-fatigue design method that reduces over conservatism and simplifies the high temperature design process

Section: Modified Coffin-shift Modelsupporting

confidence: 57%

Section: Modified Coffin-shift Modelmentioning

confidence: 98%

Section: Development Of Epp-smt Design Chartsmentioning

confidence: 99%

Section: Modified Coffin-shift Modelmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

See 3 more Smart Citations

Preliminary description of a new creep-fatigue design method that reduces over conservatism and simplifies the high temperature design process

Barua

et al. 2020

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Example Evaluation of a Representative Heat Pipe Test Article Design for Structural Acceptability using ASME Design Rules

2020

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This report describes a design analysis of a test article mimicking a core block of a heat-pipe microreactor using the ASME Boiler & Pressure Vessel Code Section III, Division 5, Subsection HB, Subpart B rules. These rules cover the design and construction of elevated temperature nuclear reactor structural components, like the core block of a heat-pipe microreactor emulated by the test article. The purposes of the report are to: (1) provide a step-by-step example of applying the ASME rules to a reasonable core block geometry and corresponding loading conditions, (2) to provide design feedback on the test article, and (3) to evaluate the ASME rules for use in the design of complex heat-pipe microreactor components. The test article design has a relatively short ASME design life, limited by high thermal stresses. The current ASME rules are adequate for evaluating the core block component designs, though they could be optimized to increase the design efficiency and reduce the amount of designer/analyst time required to execute the design process. The report covers both the current, 2019 edition ASME rules and modified, currently unapproved, rules designed to improve the design process for components with complex geometries and high secondary stresses, like the core block test article.

Design rules for 316H nuclear components cladded with nickel or tungsten

Barua

et al. 2021

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The existing Class A metallic materials qualified in the ASME Boiler & Pressure Vessel Code Section III, Division 5 rules for high temperature nuclear reactors are not ideally suited for long term molten salt corrosion resistance in high temperature molten salt reactors. Potentially, corrosion or corrosion driven environmentally assisted cracking could limit the design life of structural components in molten salt reactors. A solution to this limitation is the use of cladded components -overlay the Class A material with a thin layer of some non-qualified, corrosionresistant material. However, this necessitates the development of design methods for cladded components that do not require long-term testing of clad materials in order to support the nearterm deployment of molten salt reactors. This report develops such a methodology along with a complete set of rules presented in a format compatible with an ASME nuclear Code Case. These design rules are for 316H Class A components cladded with either nickel or tungsten. The report also discusses the development of a set of general criteria for selecting clad materials beyond the specific 316H/tungsten and 316H/nickel systems and general acceptance tests for checking the mechanical integrity of the clad/base metal interface. Finally, the report includes a set of fullydocumented sample problems detailing the application of the rules to high temperature cladded components.