Subsection NH Research Articles

Visualization of Thermal Fatigue Damage Distribution With Simplified Stress Range Calculations

This paper proposes simplified methods to evaluate fatigue damage in a component subjected to cyclic thermal loads to visualize damage distribution by using typical computer-aided engineering systems. The objective is to perform the evaluations on a standard desktop PC within a reasonably short computation time. Three simplified methods for defining elastic stress ranges are proposed in place of the method in the ASME Subsection NH procedures. A thermal fatigue test that was previously performed using a type-304 stainless steel (304SS) cylinder is simulated to validate the proposed methods. Heat transfer and elastic analyses are conducted. Simultaneously with the analyses, fatigue usage factors are calculated using user subroutines formulated in this study, including the three simplified methods and the ASME NH-based method. The calculated values of the fatigue usage factor are visualized using a graphical user interface (GUI) incorporated into a commercial finite-element analysis (FEA) code. The fatigue usage factor distribution obtained using the simplified methods could be calculated without requiring large amounts of memory and long computation time. In addition, the distribution of the fatigue usage factor was consistent with the distribution of cracks observed in the test.

Eurofer97 Creep-Fatigue assessment tool for ANSYS APDL and workbench

For the design of future fusion reactors the ferritic-martensitic steel Eurofer97 is the main candidate for in-vessel structural application to be able to withstand harsh environment like irradiation and cyclic loading under elevated temperatures. In case of high temperatures, not only fatigue but also creep damage becomes significant and has to be considered. In the past a so called Creep-Fatigue Assessment (CFA) tool has been developed for Finite Element software ANSYS APDL as post-processor within the frame of Engineering Data & Design Integration (EDDI) in EUROfusion. This tool was initially based on the elastic Creep-Fatigue rules of ASME Boiler Pressure Vessel Code (BPVC) Section III Division 1 Subsection NH Appendix T.Nowadays the tool is able to automatically identify the critical region regarding Creep-Fatigue damage in ANSYS APDL and Workbench using local stress, maximum elastic strain range and temperature from elastic thermo-mechanical Finite Element analysis. The use of stress linearization in elastic analysis allows the calculation of a modified equivalent strain range including inelastic effects to finally determine the allowable number of cycles, creep and fatigue damage fraction. For Creep-Fatigue Assessment (CFA) post-processing design fatigue curves, creep stress versus time to rupture curves, monotonic and isochronous stress versus strain curves have been used in the combination with Creep-Fatigue damage interaction diagram to describe the Creep-Fatigue behavior of Eurofer97.As it is well known that Eurofer97 shows cyclic softening, a modified Creep-Fatigue rule has been implemented in CFA tool to improve the underestimation of creep damage. This modified rule uses for calculation of creep damage the stress versus strain and design creep curves of cyclic softened material for some fraction of lifetime and an improved Creep-Fatigue damage interaction diagram of Eurofer97.In cases where the elastic analysis of ASME BPVC is too conservative, inelastic analysis can be used to calculate total strain directly instead of the prediction used in the elastic analysis. This inelastic approach for Creep-Fatigue Assessment is less complicated compared to the elastic route because only few steps according to ASME BPVC are necessary. Nevertheless more efforts for Finite Element simulation including inelastic material behavior are required.In Summary, this CFA tool can be used for fast Creep-Fatigue evaluation. It simply allows the fast identification of Creep-Fatigue damage for a structural component which is made of Eurofer97.

Development of a program for high-temperature design evaluation according to RCC-MRx

A web-based design evaluation program is developed for the elevated temperature design (ETD) analysis according to the RCC-MRx for Generation IV and fusion reactor systems. The program, named ‘HITEP_RCC-MRx,’ is composed of two modules: ‘HITEP_RCC-DBA,’ which computerizes the design-by-analysis (DBA) for class 1 components according to RB-3200 procedures, and ‘HITEP_RCC-PIPE,’ which computerizes the design-by-rule (DBR) analysis for class 1 piping according to RB-3600 procedures. Each module performs three major design evaluations on load-controlled stress, inelastic strain and creep-fatigue damage in a reliable and efficient way. The program is verified with a number of components and piping systems in sodium test facilities subjected to high temperature loading for the DBA module and the DBR module, respectively. Since HITEP_RCC-MRx is programmed with web-based language, it can operate on a smartphone as well as on a personal computer once it is connected to an internet or WIFI. Comparisons of the ETD codes of RCC-MRx and ASME Section III Subsection NH are made in terms of the programmability and reproduction of the evaluation results. A new creep model proposed for Grade 91 steel is incorporated in the program for comparison of creep behaviors between the creep models of RCC-MRx and the new Garofalo creep model for Grade 91 steel. Comparisons for verification between the evaluation results from HITEP_RCC-MRx and direct calculations using Excel confirm that the HITEP_RCC-MRx performs ETD evaluations in an efficient and reliable way.

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.

Comparison of elevated temperature design codes of ASME Subsection NH and RCC-MRx

The elevated temperature design (ETD) codes are used for the design evaluation of Generation IV (Gen IV) reactor systems such as sodium-cooled fast reactor (SFR), lead-cooled fast reactor (LFR), and very high temperature reactor (VHTR). In the present study, ETD code comparisons were made in terms of the material properties and design evaluation procedures for the recent versions of the two major ETD codes, ASME Section III Subsection NH and RCC-MRx. Conservatism in the design evaluation procedures was quantified and compared based on the evaluation results for SFR components as per the two ETD codes. The target materials are austenitic stainless steel 316 and Mod.9Cr-1Mo steel, which are the major two materials in a Gen IV SFR. The differences in the design evaluation procedures as well as the material properties in the two ETD codes are highlighted.

ANSYS Creep-Fatigue Assessment tool for EUROFER97 components

The damage caused by creep-fatigue is an important factor for materials at high temperatures. For in-vessel components of fusion reactors the material EUROFER97 is a candidate for structural application where it is subjected to irradiation and cyclic thermo-mechanical loads. To be able to evaluate fusion reactor components reliably, creep-fatigue damage has to be taken into account. In the frame of Engineering Data and Design Integration (EDDI) in EUROfusion Technology Work Programme rapid and easy design evaluation is very important to predict the critical regions under typical fusion reactor loading conditions. The presented Creep-Fatigue Assessment (CFA) tool is based on the creep-fatigue rules in ASME Boiler Pressure Vessel Code (BPVC) Section 3 Division 1 Subsection NH which was adapted to the material EUROFER97 and developed for ANSYS. The CFA tool uses the local stress, maximum elastic strain range and temperature from the elastic analysis of the component performed with ANSYS. For the assessment design fatigue and stress to rupture curves of EUROFER97 as well as isochronous stress vs. strain curves determined by a constitutive model considering irradiation influence are used to deal with creep-fatigue damage. As a result allowable number of cycles based on creep-fatigue damage interaction under given hold times and irradiation rates is obtained. This tool can be coupled with ANSYS MAPDL and ANSYS Workbench utilizing MAPDL script files.

Structural Integrity Role in Plant Life Extension - A Case Study

The steam generators at Advanced Gas-Cooled Reactor (AGR) nuclear power stations in the UK are potentially life-limiting components. Enhancing the capability to monitor the steam generators has been identified as having the potential to provide key evidence in justifying the extension of the generating lifetime of the stations. It has been proposed to install new temperature measuring instrumentation to monitor reactor gas temperature and to provide additional data regarding steam generator operating conditions. The modification will be to introduce thermocouples to the bore of an intact steam generator tube to facilitate temperature measurement at or near to the locations of interest. The modified steam generator tube will be sealed at the feed header upstand. Between the upper surface of the superheater header tubeplate and the wall of the superheater header, the thermocouple bundle and sheath will be contained within a rigid stainless steel guide tube. The guide tube will be attached at both ends by welds, each forming a pressure boundary. At the tubeplate a weld will separate the bore of the sealed guide tube from the steam space within the superheater header; a weld between the guide tube and the superheater header will separate the steam space within the superheater header from atmosphere outside the header. In order to obtain a better design, three 3-dimentional finite element models have been created using ABAQUS. A series of cyclic pressure, and start-up and shutdown thermal transient stress analyses have been carried out to provide stress values for structural integrity assessments to be conducted using ASME III, Subsection NH and R5.

소듐 시험루프 내 고온 압력용기의 크리프-피로 건전성 평가

본 연구에서는 한국원자력연구원 내에 설치될 예정인 소듐시험 시설인 SELFA(Sodium Thermalhydraulic Experiment Loop for Finned-tube Sodium-to-Air heat exchanger) 내에서 정상상태 가동온도가 510°C의 고온 압력용기인 팽창탱크에 대해 고온 건전성 평가를 수행하였다. 팽창탱크에 대해 3 차원 유한요소 해석에 기초하여 고온설계 기술기준인 ASME Section III Subsection NH 와 프랑스의 RCC-MRx 코드를 따라 크리프-피로 손상평가를 수행하였다. 평가결과 팽창탱크는 크리프-피로 설계 과도 하중 하에서 구조적 건전성을 유지하는 것으로 나타났다. 316L 스테인리스강 재질의 동 압력용기에 대해 정량적 코드 비교 분석을 수행하였다.

Creep-Fatigue Design Studies for Process Reactor Components Subjected to Elevated Temperature Service as per ASME-NH

A number of process reactors in refinery and petrochemical industry are constructed using low chrome alloys which are operating in the creep range and are in cyclic service. ASME B&PV Section VIII Code provides allowable stresses at elevated temperatures for design of reactor components, which are controlled by creep properties. However, fatigue design rules and fatigue exemption rules are not applicable, precluding construction of reactors using low chrome alloys at temperatures above 371°C (7000F). ASME B&PV Section III Division 1 - Subsection NH (ASME-NH) Code considers cyclic failure modes at elevated temperatures and provides creep-fatigue interaction rules and damage limits. In this paper, creep-fatigue damage under elevated temperatures is investigated for process reactor components using elastic analysis method of ASME-NH Code for a defined representative load cycle. Also the complexities in application procedures of ASME-NH rules with thermal and structural analysis results are described in detail.

High temperature design of finned-tube sodium-to-air heat exchanger in a sodium test loop

A sodium test loop called ‘SELFA’ (sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger) for simulating thermal hydraulic behavior of the FHX (finned-tube sodium-to-air heat exchanger) unit in a Korean prototype sodium-cooled fast reactor is planned to be constructed at KAERI (Korea Atomic Energy Research Institute). In this study, the elevated temperature design for a model FHX and creep–fatigue damage evaluation have been conducted for the model according to the design codes of ASME section III subsection NH and RCC-MRx based on full 3D finite element analyses. Design optimization for the finned-tubes and tube arrangements in the scaled-down FHX has been performed. The materials of the FHX and piping systems are austenitic stainless steel type 316. The design temperature of the SELFA test loop is 600°C and the design pressure is 1MPa. The damage evaluation results have shown that no creep–fatigue damage occurs in the present design of the FHX under the intended test conditions.

소듐냉각고속로 붕괴열교환기의 고온 설계 및 건전성 평가

In this study, high temperature design and creep-fatigue damage evaluation of a decay heat exchanger (DHX) in the decay heat removal systems of a sodium-cooled fast reactor (SFR) have been performed. Detail design and 3D finite element analysis have been conducted for the DHXs to be installed in active and passive decay heat removal systems in Korean Generation IV SFR, and the DHX installed in the STELLA-1(Sodium integral effect test loop for safety simulation and assessment) at KAERI (Korea Atomic Energy Research Institute). Evaluations of creep-fatigue damage based on full 3D finite element analyses were conducted for the two Mod.9Cr-1Mo steel heat exchangers according to the elevated temperature design codes of ASME Section III Subsection NH and RCC-MR code. Code comparisons were made based on the creep-fatigue damage evaluation and issues on conservatisms of the design codes were discussed.

High temperature design and damage evaluation of a helical type sodium-to-air heat exchanger in a sodium-cooled fast reactor

A high temperature design and creep-fatigue damage evaluation for a helical type sodium-to-air heat exchanger (AHX) in a 600MWe sodium-cooled fast reactor (SFR) has been conducted. The AHX is a safety-grade component in a passive decay heat removal (DHR) system. Safe and reliable decay heat removal is one of the most important tasks in the successful design of a sodium-cooled fast reactor. Therefore, independent, diverse and redundant DHR systems incorporating both passive and active mechanisms have been employed in the demonstration SFR developed by Korea Atomic Energy Research Institute (KAERI). One of the key DHR components is sodium-to-air heat exchanger, which has a helically coiled tube arrangements. A creep-fatigue damage evaluation has been performed according to the elevated temperature design codes of ASME Subsection NH and RCC-MRx based on a full 3D finite element analysis. The integrity of the heat exchangers under creep-fatigue loading was confirmed and code comparisons were made as well.

High Temperature Design and Damage Evaluation of MOD.9Cr-1Mo Steel Heat Exchanger

High temperature design and evaluation of creep-fatigue damage for sodium-to-sodium heat exchanger, decay heat exchanger (DHX), in a sodium test loop have been conducted. The DHX is a shell- and tube-type heat exchanger with an outer diameter of 21.7 mm and effective length of 1.73 m. The DHX shell and tube materials were Mod.9Cr-1Mo steel. The temperatures of shell inlet and shell outlet in the DHX are 510 °C and 308 °C, respectively, while the temperatures of tube inlet and outlet are 254 °C and 475 °C, respectively. Three-dimensional finite element analysis was conducted for the DHX, and evaluation of creep-fatigue damage at several critical locations of the heat exchanger was carried out according to the elevated temperature design codes of the ASME Section III Subsection NH and RCC-MR. Evaluations on the integrity of the DHX and code comparisons were carried out for the critical locations of the DHX. The evaluation results showed that damages at the tubesheet joints of tube-to-tubesheet and tubesheet-to-shell were not so critical in the present DHX model.

Review of the finite element models for a structural integrity evaluation of the sodium-cooled fast reactor high temperature piping

We compared the structural analysis feature of finite element (FE) models for the structural integrity evaluation of the sodium-cooled fast reactor (SFR) high temperature piping and to evaluate the structural integrity against the typical duty cycle event. To evaluate the structural integrity of the high temperature piping per ASME Subsection NH rules, the structural analysis should be carried out first by using a 3-dimensional structural model. The object FE models under consideration in this study are pipe element model, 3-D full model, and 3-D simplified model. The pipe element model is based on the 3-D beam element and effective in understanding overall deformation but less favorable to the detailed stress distribution. The 3-D full model consists of solid structure as well as the contained coolant inside the piping structure with the fluid element. The 3-D simplified model consists of structure shape only, but its material properties are recalculated to reflect the coolant weight effect. The loading conditions for the structural analyses are the mechanical load including dead weight and steady state thermal load. From the analysis results, the piping element model shows the smallest stress intensity, and the required time for FE analysis is also the shortest. The 3-D simplified model shows the most conservative stress intensity output but its calculation time is less than the 3-D full model.

Stress and Strain Locus of Perforated Plate in Inelastic Deformation—Strain-Controlled Loading Case

Plastic strain of structures having stress concentration is estimated by using the simplified method or the finite element elastic solutions. As the simplified methods used in codes and standards, we can cite Neuber’s formula in the work by American Society of Mechanical Engineers (1995, “Boiler and Pressure Vessel Code,” ASME-Code, Section 3, Division 1, Subsection NH) and by Neuber (1961, “Theory of Stress Concentration for Shear Strained Prismatic Bodies With Arbitrary Nonlinear Stress-Strain Law,” ASME, J. Appl. Mech., 28, pp.544–550) and elastic follow-up procedure in the work by Japan Society of Mechanical Engineers [2005, “Rules on Design and Construction for Nuclear Power Plants, 2005, Division 2: Fast Breeder Reactor” (in Japanese)]. Also, we will cite stress redistribution locus (SRL) method recently proposed as the other simplified method in the work by Shimakawa et al. [2002, “Creep-Fatigue Life Evaluation Based on Stress Redistribution Locus (SRL) Method,” JPVRC Symposium 2002, JPVRC/EPERC/JPVRC Joint Workshop sponsored by JPVRC, Tokyo, Japan, pp. 87–95] ad by High Pressure Institute of Japan [2005, “Creep-Fatigue Life Evaluation Scheme for Ferritic Component at Elevated Temperature,” HPIS C 107 TR 2005 (in Japanese)]. In the present paper, inelastic finite element analysis of perforated plate, whose stress concentration is about 2.2–2.5, is carried out, and stress and strain locus in inelastic range by the detailed finite element solutions is investigated to compare accuracy of the simplified methods. As strain-controlled loading conditions, monotonic loading, cyclic loading, and cyclic loading having hold time in tension under strain-controlled loading are assumed. The inelastic strain affects significantly life evaluation of fatigue and creep-fatigue failure modes, and the stress and strain locus is discussed from the detailed inelastic finite element solutions.

Buckling limit evaluations for reactor vessel of ABTR

We evaluated the buckling limit of a conceptually designed reactor vessel of the Advanced Burner Test Reactor (ABTR) for a horizontal safe shutdown earthquake (SSE) seismic load. In this evaluation, both seismic isolation and non-isolation designs were considered for thin reactor vessels subjected to elevated temperature services. For calculating the buckling load, two kinds of methods, a numerical simulation method using finite element analysis and an evaluation formula driven from a theoretical basis, were used. To consider the material aging effect caused by a 60-year design lifetime and 510°C normal operating temperature, an isochronous stress-strain curve corresponding to these conditions was used for the nonlinear elastic-plastic buckling analysis method. From the evaluation results of the buckling load, it was found that plasticity behavior significantly affects the buckling strength, but that the initial geometrical imperfections have little effect. Also, the non-seismic isolation design does not satisfy the buckling limit rules of the ASME BPV Section III, Subsection NH, but the seismic isolation design does satisfy it with sufficient margins.

Application of Negligible Creep Criteria to Candidate Materials for HTGR Pressure Vessels

Two of the proposed high temperature gas reactors (HTGRs) under consideration for a demonstration plant have the design object of avoiding creep effects in the reactor pressure vessel during normal operation. This work addresses the criteria for negligible creep in subsection NH, Division 1 of the ASME Boiler and Pressure Vessel Code, Sec. III, other international design codes, and some currently suggested criteria modifications and their impact on permissible operating temperatures for various reactor pressure vessel materials. The goal of negligible creep could have different interpretations depending on what failure modes are considered and associated criteria for avoiding the effects of creep. It is shown that for the materials of this study, consideration of localized damage due to cycling of peak stresses results in a lower temperature for negligible creep than consideration of the temperature at which the allowable stress is governed by the creep properties. In assessing the effect of localized cyclic stresses, it is also shown that consideration of cyclic softening is an important effect that results in a higher estimated temperature for the onset of significant creep effects than would be the case if the material were cyclically hardening. There are other considerations for the selection of vessel material besides avoiding creep effects. Of interest for this review are (1) the material’s allowable stress level and impact on the wall thickness (the goal being to minimize the required wall thickness) and (2) ASME code approval (inclusion as a permitted material in the relevant section and subsection of interest) to expedite regulatory review and approval. The application of negligible creep criteria to two of the candidate materials, SA533 and Mod 9Cr–1Mo (also referred to as Grade 91), and to a potential alternate, normalized and tempered 214 Cr–1Mo, is illustrated, and the relative advantages and disadvantages of the materials are discussed.

Evaluation of Creep-Fatigue Damage for Hot Gas Duct Structure of the NHDD Plant

Evaluation of creep-fatigue damage has been carried out for the hot gas duct (HGD) structure in the nuclear hydrogen development and demonstration (NHDD) plant. The core outlet and inlet temperature of the NHDD plant are 950°C and 490°C, respectively. Case studies on high temperature design codes of the draft code case for Alloy 617, ASME boiler and pressure vessel code section III subsection NH (ASME-NH), and RCC-MR were carried out for the inner tube of the HGD for the candidate materials of Alloy 617 and Alloy 800H. Technical issues in application of the draft code case to a high temperature structure are discussed for the Alloy 617 material. Code comparison between the ASME-NH and RCC-MR for Alloy 800H has been carried out. The candidate material of the outer pressure boundary (cross vessel) of the HGD is Mod.9Cr-1Mo steel. The damage evaluation, according to the ASME-NH and RCC-MR for the cross vessel of Mod.9Cr-1Mo steel, has been conducted and their results were compared.

Creep Effects on Design Below the Temperature Limits of ASME Section III Subsection NB

Some recent studies of material response have identified an issue that crosses over and blurs the boundary between ASME Boiler and Pressure Vessel Code Section III Subsection NB and Subsection NH. For very long design lives, the effects of creep show up at lower and lower temperature as the design life increases. Although true for the temperature at which the allowable stress is governed by creep properties, the effect is more apparent, e.g., creep effects show up sooner, at local structural discontinuities and peak thermal stress locations. This is because creep is a function of time, temperature, and stress, and the higher the localized stress, the lower in temperature creep begins to cause damage. If the threshold is below the Subsection NB to NH temperature boundary, 700°F for ferritic steels and 800°F for austenitic materials, then this potential failure mode will not be considered. Unfortunately, there is no experience base with very long lives at temperatures close to but under the Subsection NB to NH boundary to draw on. This issue is of particular interest in the application of Subsection NB rules of construction to some high temperature gas-cooled reactor concepts. The purpose of this paper is, thus, twofold: one part is about statistical treatment and extrapolation of sparse data for a specific material of interest, SA-533 Grade B Class 1; the other part is about how these results could impact current design procedures in Subsection NB.

A Review on Current Status of Alloys 617 and 230 for Gen IV Nuclear Reactor Internals and Heat Exchangers 1

Alloys 617 and 230 are currently identified as two leading candidate metallic materials in the down selection for applications at temperatures above 760°C in the Gen IV nuclear reactor systems. Qualifying the materials requires significant information related to codification, mechanical behavior modeling, metallurgical stability, environmental resistance, and many other aspects. In the present paper, material requirements for the Gen IV nuclear reactor systems are discussed; available information regarding the two alloys for the intended applications are reviewed and analyzed; and further R&D activities are suggested. In the United States the major requirement for qualifying the materials is to satisfy the ASME Subsection NH, with adequate considerations for NRC, ASME NQA-1, and Section XI. In comparison, Alloy 617 is more studied with larger existing databases in air and helium, while Alloy 230 may have highly desired potentials but needs more exploration. To provide a sound technical basis for the material selection decision, more data should be generated to characterize behaviors of both alloys in creep, loading rate sensitivity, fatigue, creep-fatigue, crack resistance, toughness, product form dependency, and metallurgical stability.