Low Cycle Fatigue Behavior of Types 304 and 316 Stainless Steel Tested in Sodium at 550°C

Zeman, G.J.; Smith, D.L.

doi:10.13182/nt79-a32164

Cited by 14 publications

(5 citation statements)

References 14 publications

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“…The available data indicate that both the environment and microstructural changes due to sodium exposure influence the low-cycle fatigue behavior of austenitic SSs , Zeman and Smith 1979, Flagella et al 1978, Flagella and Kahrs 1976, Flagella 1983, Cowgill 1985, Husslage and Drenth 1980. In general, a low-oxygen sodium environment has a beneficial effect on the fatigue life of Types 304 and 316 SS.…”

Section: Fatigue and Creep-fatiguementioning

confidence: 99%

Code qualification of structural materials for AFCI advanced recycling reactors.

Natesan¹,

Li²,

Nanstad³

et al. 2012

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This report summarizes the further findings from the assessments of current status and future needs in code qualification and licensing of reference structural materials and new advanced alloys for advanced recycling reactors (ARRs) in support of Advanced Fuel Cycle Initiative (AFCI). The work is a combined effort between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) with ANL as the technical lead, as part of Advanced Structural Materials Program for AFCI Reactor Campaign. The report is the second deliverable in FY08 (M505011401) under the work package "Advanced Materials Code Qualification".The overall objective of the Advanced Materials Code Qualification project is to evaluate key requirements for the ASME Code qualification and the Nuclear Regulatory Commission (NRC) approval of structural materials in support of the design and licensing of the ARR. Advanced materials are a critical element in the development of sodium reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility, but also is essential for the economics of future advanced sodium reactors. Code qualification and licensing of advanced materials are prominent needs for developing and implementing advanced sodium reactor technologies. Nuclear structural component design in the U.S. must comply with the ASME Boiler and Pressure Vessel Code Section III (Rules for Construction of Nuclear Facility Components) and the NRC grants the operational license. As the ARR will operate at higher temperatures than the current light water reactors (LWRs), the design of elevated-temperature components must comply with ASME Subsection NH (Class 1 Components in Elevated Temperature Service). However, the NRC has not approved the use of Subsection NH for reactor components, and this puts additional burdens on materials qualification of the ARR. In the past licensing review for the Clinch River Breeder Reactor Project (CRBRP) and the Power Reactor Innovative Small Module (PRISM), the NRC/Advisory Committee on Reactor Safeguards (ACRS) raised numerous safety-related issues regarding elevated-temperature structural integrity criteria. Most of these issues remained unresolved today. These critical licensing reviews provide a basis for the evaluation of underlying technical issues for future advanced sodium-cooled reactors.Major materials performance issues and high temperature design methodology issues pertinent to the ARR are addressed in the report. The report is organized as follows: the ARR reference design concepts proposed by the Argonne National Laboratory and four industrial consortia were reviewed first, followed by a summary of the major code qualification and licensing issues for the ARR structural materials. The available database is presented for the ASME Code-qualified structural alloys (e.g. 304, 316 stainless steels, 2.25Cr-1Mo, and mod.9Cr-1Mo), including physical properties, tensile properties, impact properties and fracture toughness, creep, fatigue, creep-fatigue in...

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Section: Fatigue and Creep-fatiguementioning

confidence: 99%

Code qualification of structural materials for AFCI advanced recycling reactors.

Natesan¹,

Li²,

Nanstad³

et al. 2012

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“…The fatigue life of Type 316SS in sodium is significantly longer than in air, as shown in Fig. 4 [34][35][36][37][38]. For the sodium exposure conditions used in these studies, the specimens developed a carbon concentration profile that varied from 0.4 wt% at the surface to the initial concentration in the steels at a depth of 0.01-0.02 cm.…”

Section: Fatigue and Creep-fatigue Performancementioning

confidence: 99%

Sodium effects on mechanical performance and consideration in high temperature structural design for advanced reactors

Natesan

Chopra

et al. 2009

Journal of Nuclear Materials

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“…As a result of the failure of the welded joints during in-service results in leaks in the pipes of the welding joints so that production stops and suffers losses because they have to shut down or stop production. [7] If there is a shutdown or cessation of operations, it will be very detrimental because there is lost time manhours, and material losses due to the cessation of production of products produced if production continues, for example at an oil refinery if a leak occurs, the effect of the loss is how many barrels per day, how much Employee salaries must be paid daily and are most at risk if a leak or cessation of an operation has an impact on the safety of workers' lives or pollutes the environment, so the effect is not only a matter of material but there are legal consequences for a company in the industrial world. Therefore, an in-depth study or comprehensive research is needed to solve cases of failure during the use of equipment (in service failure), especially welding dissimilar welding joints in the industrial world.…”

Section: Introductionmentioning

confidence: 99%

Study Effect of Holding Time at Post Weld Heat Treatment (PWHT) on Joint Dissimilar Welding ASTM 335 P11 to ASTM 304L under Controlled Condition

Ismadi,

Winarto

2024

KEM

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Welding joint Austenitic stainless steels can undergo intergranular corrosion attack in some service conditions, such as oil refineries, petrochemical plants, and industrial furnaces operating at high temperatures (between 500°C and 675°C). This research focuses on fabricating dissimilar metal welds to avoid failure due to sensitization and investigate stress-relieved carbon steel by Post-weld heat treatment (PWHT). This work studies a dissimilar weld joint (DWJ) of ASTM 335 grade P11 joint to ASTM 304L and PWHT after welding. These welding processes are used in multi-pass gas tungsten arc welding (GTAW) using filler metal R309L. PWHT at temperatures of 550 °C with holding times of 40, 50, and 60 minutes and then slow cooling in atmospheric air temperature, It was used to compare the results holding time carbon steel ASTM A335 Gr. P11 while processing. Holding time at 50 minutes has the maximum result of PWHT with 182 HV and has a good distribution of perlite with fine grain and increasing holding time also increases formed carbide on the grain boundary of stainless steel 304L.

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Low Cycle Fatigue Behavior of Types 304 and 316 Stainless Steel Tested in Sodium at 550°C

Cited by 14 publications

References 14 publications

Code qualification of structural materials for AFCI advanced recycling reactors.

Code qualification of structural materials for AFCI advanced recycling reactors.

Sodium effects on mechanical performance and consideration in high temperature structural design for advanced reactors

Study Effect of Holding Time at Post Weld Heat Treatment (PWHT) on Joint Dissimilar Welding ASTM 335 P11 to ASTM 304L under Controlled Condition

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