Evaluation of the Japanese Fatigue Test Data in Gr.91 for Elevated Temperature Design

The ASME Boiler and Pressure Vessel Code (ASME BPVC) Section III, Division 5, Subsection HB, Subpart B provided only one design fatigue curve for Grade 91 steel (Gr.91) at 540 °C (or 1000 °F) in 2019 and earlier versions. To overcome this disadvantage, The ASME Section III Working Group on Creep-Fatigue and Negligible Creep (WG-CFNC) had taken an action to incorporate the temperature-dependent design fatigue curves for Gr. 91 developed by Japan Society of Mechanical Engineers (JSME) into ASME BPVC Section III Division 5. As a result, the temperature dependent design fatigue curves are provided in the 2021 edition of the ASME BPVC. To clear the features of the best-fit fatigue curve equation developed by the JSME, 305 data stored in the database were analyzed. Details of the database and relationship between the best-fit fatigue curve equation and the data including the statistic values and the values of 95% and 99% lower confidence bound calculated by failure probability assessment were clarified through analysis. In addition to the best-fit fatigue curve equation, an equation for dynamic stress-strain response showing the behavior of Gr.91 steel under cyclic loading of is also provided based on the same database. Moreover, some additional available data of fatigue and creep-fatigue tests obtained in Japan are also provided for considering the creep-fatigue damage evaluation under elevated temperature condition.

Inelastic Material Models of Type 316H for Elevated Temperature Design of Advanced High Temperature Reactors

In this paper, the inelastic material models for Type 316H stainless steel, which is one of the principal candidate materials for elevated temperature design of the advanced high temperature reactors (HTRs) pressure retained components, are investigated and the required material parameters are identified to be used for both elasto-plastic models and unified viscoplastic models. In the constitutive equations of the inelastic material models, the kinematic hardening behavior is expressed with the Chaboche model with three backstresses, and the isotropic hardening behavior is expressed by the Voce model. The required number of material parameters is minimized to be ten in total. For the unified viscoplastic model, which can express both the time-independent plastic behavior and the time-dependent viscous behavior, the constitutive equations have the same kinematic and isotropic hardening parameters of the elasto-plastic material model with two additional viscous parameters. To identify the material parameters required for these constitutive equations, various uniaxial tests were carried out at isothermal conditions at room temperature and an elevated temperature range of 425–650 °C. The identified inelastic material parameters were validated through the comparison between tests and calculations.

Evaluation of Mean Stress Correction on Fatigue Curves of Grade 91 and Alloy 617 in ASME Section III Division 5

The current ASME Boiler and Pressure Vessel (B&PV) Code Section III, Division 5, Subsection HB, Subpart B has only one design fatigue curve for grade 91 steel (Gr. 91) at 540 °C (or 1000 °F). The ASME Section III Working Group on Creep-Fatigue and Negligible Creep (WG-CFNC) has taken an action to incorporate the temperature-dependent design fatigue curves for Gr. 91 developed by Japan Society of Mechanical Engineers (JSME) into ASME Section III Division 5. During the process, issues regarding the effect of mean stress on fatigue analysis, and how to consider the mean stress effect for elevated-temperature design, were brought up. To evaluate whether the design fatigue curves of Gr. 91 needed adjustment to account for mean stress, critical tests were designed and performed at 371 °C (700 °F) and 540 °C (1000 °F). This study is similar to the work performed on Alloy 617 when its fatigue design curves were established for temperature range of 538–704°C (1000–1300°F) as part of the Code Case package for Alloy 617 to be used as Class A construction material in Division 5. The effects of mean stress on Alloy 617 were evaluated at 550°C (1022°F). The results showed that the mean stresses introduced by the non-zero mean strain could not be maintained under strain-controlled fatigue and resulted in negligible effect on the fatigue life. Mean stress correction was not recommended for Alloy 617 fatigue design curves in Division 5. This study shows the same conclusion for Gr. 91.

HITEP_RCC-MRx Program for the Support of Elevated Temperature Design Evaluation and Defect Assessment

An integrated software platform of high-temperature design evaluation and defect assessment for a nuclear component and piping system subjected to high-temperature operation in creep regime has been developed. The program, called “HITEP_RCC-MRx,” is based on French nuclear grade high-temperature design code of RCC-MRx and enables a designer to conduct not only elevated temperature design evaluation but also elevated temperature defect assessment. HITEP_RCC-MRx consists of three modules: “HITEP_RCC-DBA,” which is programmed for the design-by-analysis (DBA) evaluation for class 1 pressure boundary components such as the pressure vessel and heat exchangers according to the RB-3200 procedures; “HITEP_RCC-PIPE,” which is programmed for the design-by-rule (DBR) evaluation according to the RB-3600 procedures; and “HITEP_RCC-A16,” which is programmed for high-temperature defect assessment according to the A16 procedures. The program has been verified with a number of related example problems on modules of DBA, Pipe, and A16. It was shown from the verification examples that integrated software platform of HITEP_RCC-MRx is capable of conducting three functions of an elevated temperature design evaluation for pressure boundary components and for piping, and an elevated defect assessment in an efficient and reliable way.

Development of the External Pressure Charts for 2¼Cr-1Mo and Mod.9Cr-1Mo Steel at Elevated Temperature Design

For the required thickness estimation against buckling in the elevated temperature design, the external pressure chart for two kinds of ferritic steel, 2 1/4Cr-1Mo and Mod.9Cr-1Mo steel, was developed. On the basis of the guideline described in the ASME BPVC Section II, Part D, Mandatory Appendix 3 with mechanical and physical properties provided in the JSME fast reactor code, the external pressure charts for each material were constructed. The minimum stress-strain curve for evaluating the external pressure chart was applied the stress-strain equation with design yield strength, Sy, provided by the JSME fast reactor code. As a result, three external pressure charts with digital values were proposed for elevated temperature design. Moreover, the rationalization effect from the current alternative was evaluated by the sample problem. This proposal resolves two issues. One is alternative use of chart for lower strength material over the 150 °C. The other is the external pressure chart above 480°C for which ferritic steels are not available.

Assessment of Creep Crack Growth Rates for Grade 91 Weld Joint at 550°C

This study presents assessment of creep crack growth rates (CCGRs) for the base metal (BM), weld metal (WM), and heat affected zone (HAZ) of Gr. 91 weld joint, which was prepared by a shield metal arc weld (SMAW) method. A series of tensile, creep, creep crack growth (CCG) tests were performed for the BM, WM, and HAZ at the identical temperature of 550°C. The CCGR laws for the BM, WM and HAZ were constructed and compared in terms of a C*-fracture parameter. In addition, the CCGR law tested for BM was compared to that of RCC-MRx code. For a given value of C*, the WM and HAZ were almost similar in the CCGR, but they were significantly faster than the BM. This reason was closely attributed to the higher creep rate in the WM and HAZ than the BM. Currently elevated temperature design (ETD) code in French, RCC-MRx was found to be non-conservative in the CCGR when compared with the present investigation.

Development of Software for the Implementation of Elastic and Simplified Inelastic Rules for Design of Class A Nuclear Components in Elevated Temperature Service

Software is being developed to aid assessment procedures of components under specified loading conditions in accordance with the elevated temperature design requirements for Class A components in ASME Boiler and Pressure Vessel Code, Section III, Rules for Construction of Nuclear Facility Components, Division 5, High Temperature Reactors, Subsection HB, Subpart B (HBB). There are many features and alternative paths of varying complexity in HBB. The initial focus of this program is a basic path through the various options for a single reference material, 316H stainless steel. However, the program will be structured for eventual incorporation all of the features and permitted materials of HBB. This paper focuses on a description of the overall program, particular challenges in developing numerical procedures for the assessment, and an overall description of the approach to computer program development.

Design and Integrity Evaluation of a Finned-Tube Sodium-to-Air Heat Exchanger in a Sodium Test Facility

A high-temperature design and an integrity evaluation for a finned-tube sodium-to-air heat exchanger (FHX) in a sodium test facility were conducted based on full 3D finite-element analyses, and comparisons of the design codes were made. A model FHX has been installed in a sodium test facility of sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger (SELFA) for simulating the thermal hydraulic behavior of the FHX unit in the prototype Gen IV sodium-cooled fast reactor (PGSFR). For the design evaluations, ASME Section VIII Div. 2 has been applied for the FHX as a whole. For parts of the FHX operating in the creep regime, nuclear grade elevated temperature design (ETD) codes of ASME Section III Subsection NH and RCC-MRx were additionally applied to evaluate the integrity against creep-fatigue damage. For parts of the FHX operating at low temperature, ASME Section III Subsection NB was applied additionally to evaluate the integrity upon load-controlled stresses and fatigue. The integrity of the FHX was confirmed based on the design evaluations as per the design codes. Code comparisons were made in terms of the chemical compositions, material properties, and conservatism. The conservatism was quantified and compared at the critical low temperature location between ASME Section VIII Div. 2 and ASME-NB, and at the critical high-temperature location between ASME-NH and RCC-MRx.