French RSE-M and RCC-MR code appendices for flaw analysis: Presentation of the fracture parameters calculation—Part I: General overview

Marie, Stéphane; Chapuliot, S.; Kayser, Y.; Lacire, M.H.; Drubay, B.; Barthelet, B.; Delliou, Patrick Le; Rougier, Viviana Carolina; Naudin, C.; Gilles, Philippe; Triay, M.

doi:10.1016/j.ijpvp.2007.05.003

Cited by 17 publications

(9 citation statements)

References 9 publications

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“…It is similar to the J integral calculations employed for flaw evaluations in Light Water Reactor applications, Marie et al (2007). A16 has a wealth of formulas for determining the fracture mechanics parameters for specific components such as plates, pipes, and elbows.…”

Section: A16 Procedures Of Rcc-mrmentioning

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: A16 Procedures Of Rcc-mrmentioning

confidence: 99%

Code qualification of structural materials for AFCI advanced recycling reactors.

Natesan¹,

Li²,

Nanstad³

et al. 2012

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“…This section gives details of the reference stress calculation for cracked plates subjected to mechanical loadings presented in part I [3].…”

Section: Reference Stress Calculation For Mechanical Loadingsmentioning

confidence: 99%

“…All the developments have been integrated in the 2005 edition of RSE-M [1] and in the 2007 edition of RCC-MR [2]. This series of articles is composed of 5 parts: the first part [3] presents an overview of the methods proposed in the RCC-MR and RSE-M codes. Parts II-IV provide all the compendia for a specific component.…”

Section: Introductionmentioning

confidence: 99%

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French RSE-M and RCC-MR code appendices for flaw analysis: Presentation of the fracture parameters calculation—Part II: Cracked plates

Marie

Chapuliot

Kayser

et al. 2007

International Journal of Pressure Vessels and Piping

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“…All the developments have been integrated in the 2005 edition of RSE-M [1] and in 2007 edition of RCC-MR [2]. This series of papers is composed of five parts: the first [3] presents an overview of the methods in the RCC-MR and RSE-M codes. Parts II-IV provide compendia for specific components.…”

Section: Introductionmentioning

confidence: 99%

French RSE-M and RCC-MR code appendices for flaw analysis: Presentation of the fracture parameters calculation—Part IV: Cracked elbows

Marie

Chapuliot

Kayser

et al. 2007

International Journal of Pressure Vessels and Piping

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Two French nuclear codes include flaw assessment procedures: the RSE-M Code ''Rules for In-service Inspection of Nuclear Power Plant Components'' and the RCC-MR code ''Design and Construction rules for mechanical components of FBR nuclear islands and high temperature applications''. Development of analytical methods has been made for the last 10 years through a collaboration between CEA, EDF and AREVA-NP, and through R&D actions involving CEA and IRSN. These activities have led to unification of the common methods of the two codes. The calculation of fracture mechanics parameters, and in particular the stress intensity factor K I and the J integral, has been widely developed for industrial configurations. All the developments have been integrated in the 2005 edition of RSE-M and in 2007 edition of RCC-MR.This series of papers is composed of five parts: the first presents an overview of the methods proposed in the RCC-MR and RSE-M codes. Parts II-IV provide compendia for specific components. The geometries are plates (part II), pipes (part III) and elbows (part IV). Part V presents validation of the methods, with details on their accuracy. This paper presents the stress intensity factor and J calculation for cracked elbows. General data applicable for all defect geometries are first presented, and then, compendia for K I and s ref calculations are provided for the available defect geometries. r

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French RSE-M and RCC-MR code appendices for flaw analysis: Presentation of the fracture parameters calculation—Part I: General overview

Cited by 17 publications

References 9 publications

Code qualification of structural materials for AFCI advanced recycling reactors.

Code qualification of structural materials for AFCI advanced recycling reactors.

French RSE-M and RCC-MR code appendices for flaw analysis: Presentation of the fracture parameters calculation—Part II: Cracked plates

French RSE-M and RCC-MR code appendices for flaw analysis: Presentation of the fracture parameters calculation—Part IV: Cracked elbows

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