Paper 9. Sodium Heated Steam Generators: Near Term and Projected Information Needs-Ferritic Steels

Spalaris, C.N.; Challenger, K. D.; Dutina, D.; Ring, P.

doi:10.1680/fsffrsgv1.00490.0009

Ferritic Steels for Fast Reactor Steam Generators 1978

DOI: 10.1680/fsffrsgv1.00490.0009

|View full text |Cite

Paper 9. Sodium Heated Steam Generators: Near Term and Projected Information Needs-Ferritic Steels

C.N. Spalaris¹,

K. D. Challenger²,

D. Dutina³

et al.

Help me understand this report

Search citation statements

Order By: Relevance

Paper Sections

Select...

Citation Types

Supporting

Mentioning

Contrasting

Year Published

2012

Publication Types

Select...

Other1

Relationship

Self Cite0

Independent1

Authors

Journals

Cited by 1 publication

(1 citation statement)

References 2 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…A decrease in the decarburization rate constant with time has been observed in isothermally annealed Fe-2.25Cr-1Mo steel after prolonged sodium exposure (Spalaris et al 1978). The studies on the C-transfer behavior of Fe-9Cr-Mo steels in sodium show that these steels can either carburize or decarburize depending on the C activity in sodium (Chopra et al 1976, Menken et al 1978.…”

mentioning

confidence: 99%

Code qualification of structural materials for AFCI advanced recycling reactors.

Natesan¹,

Li²,

Nanstad³

et al. 2012

View full text Add to dashboard Cite

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...

show abstract

mentioning

confidence: 99%

Code qualification of structural materials for AFCI advanced recycling reactors.

Natesan¹,

Li²,

Nanstad³

et al. 2012

View full text Add to dashboard Cite

show abstract

scite is a Brooklyn-based organization that helps researchers better discover and understand research articles through Smart Citations–citations that display the context of the citation and describe whether the article provides supporting or contrasting evidence. scite is used by students and researchers from around the world and is funded in part by the National Science Foundation and the National Institute on Drug Abuse of the National Institutes of Health.

Contact Info

customersupport@researchsolutions.com

10624 S. Eastern Ave., Ste. A-614

Henderson, NV 89052, USA

Blog Terms and Conditions API Terms Privacy Policy Contact Cookie Preferences Do Not Sell or Share My Personal Information

Made with 💙 for researchers

Part of the Research Solutions Family.

Paper 9. Sodium Heated Steam Generators: Near Term and Projected Information Needs-Ferritic Steels

Cited by 1 publication

References 2 publications

Code qualification of structural materials for AFCI advanced recycling reactors.

Code qualification of structural materials for AFCI advanced recycling reactors.

Contact Info

Product

Resources

About