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Title: Code qualification of structural materials for AFCI advanced recycling reactors.

Abstract

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 Boilermore » 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 interaction, microstructural stability during long-term thermal aging, material degradation in sodium environments and effects of neutron irradiation for both base metals and weld metals. An assessment of modified versions of Type 316 SS, i.e. Type 316LN and its Japanese version, 316FR, was conducted to provide a perspective for codification of 316LN or 316FR in Subsection NH. Current status and data availability of four new advanced alloys, i.e. NF616, NF616+TMT, NF709, and HT-UPS, are also addressed to identify the R&D needs for their code qualification for ARR applications. For both conventional and new alloys, issues related to high temperature design methodology are described to address the needs for improvements for the ARR design and licensing. Assessments have shown that there are significant data gaps for the full qualification and licensing of the ARR structural materials. Development and evaluation of structural materials require a variety of experimental facilities that have been seriously degraded in the past. The availability and additional needs for the key experimental facilities are summarized at the end of the report. Detailed information covered in each Chapter is given.« less

Authors:
; ; ; ;  [1];  [2]
  1. (Nuclear Engineering Division)
  2. (
Publication Date:
Research Org.:
Argonne National Lab. (ANL), Argonne, IL (United States)
Sponsoring Org.:
NE
OSTI Identifier:
1042575
Report Number(s):
ANL-AFCI-244
TRN: US1202876
DOE Contract Number:
DE-AC02-06CH11357
Resource Type:
Technical Report
Country of Publication:
United States
Language:
ENGLISH
Subject:
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; 36 MATERIALS SCIENCE; ALLOYS; CLINCH RIVER BREEDER REACTOR; CREEP; ECONOMICS; FLEXIBILITY; FRACTURE PROPERTIES; FUEL CYCLE; IRRADIATION; LICENSING; NEUTRONS; PHYSICAL PROPERTIES; POWER REACTORS; REACTOR COMPONENTS; RECYCLING; SAFETY MARGINS; SODIUM; STABILITY; STAINLESS STEELS; TENSILE PROPERTIES

Citation Formats

Natesan, K., Li, M., Majumdar, S., Nanstad, R.K., Sham, T.-L., and ORNL). Code qualification of structural materials for AFCI advanced recycling reactors.. United States: N. p., 2012. Web. doi:10.2172/1042575.
Natesan, K., Li, M., Majumdar, S., Nanstad, R.K., Sham, T.-L., & ORNL). Code qualification of structural materials for AFCI advanced recycling reactors.. United States. doi:10.2172/1042575.
Natesan, K., Li, M., Majumdar, S., Nanstad, R.K., Sham, T.-L., and ORNL). 2012. "Code qualification of structural materials for AFCI advanced recycling reactors.". United States. doi:10.2172/1042575. https://www.osti.gov/servlets/purl/1042575.
@article{osti_1042575,
title = {Code qualification of structural materials for AFCI advanced recycling reactors.},
author = {Natesan, K. and Li, M. and Majumdar, S. and Nanstad, R.K. and Sham, T.-L. and ORNL)},
abstractNote = {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 interaction, microstructural stability during long-term thermal aging, material degradation in sodium environments and effects of neutron irradiation for both base metals and weld metals. An assessment of modified versions of Type 316 SS, i.e. Type 316LN and its Japanese version, 316FR, was conducted to provide a perspective for codification of 316LN or 316FR in Subsection NH. Current status and data availability of four new advanced alloys, i.e. NF616, NF616+TMT, NF709, and HT-UPS, are also addressed to identify the R&D needs for their code qualification for ARR applications. For both conventional and new alloys, issues related to high temperature design methodology are described to address the needs for improvements for the ARR design and licensing. Assessments have shown that there are significant data gaps for the full qualification and licensing of the ARR structural materials. Development and evaluation of structural materials require a variety of experimental facilities that have been seriously degraded in the past. The availability and additional needs for the key experimental facilities are summarized at the end of the report. Detailed information covered in each Chapter is given.},
doi = {10.2172/1042575},
journal = {},
number = ,
volume = ,
place = {United States},
year = 2012,
month = 5
}

Technical Report:

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  • The objective of this letter is to inform you of recent progress on the development of advanced structural materials in support of advanced fast reactors and AFCI. As you know, the alloy development effort has been initiated in recent months with the procurement of adequate quantities of the NF616 and HT-UPS alloys. As the test alloys become available in the coming days, mechanical testing, evaluation of optimizing treatments, and screening of environmental effects will be possible at a larger scale. It is therefore important to establish proper quality assurance protocols for this testing effort in a timely manner to ensuremore » high technical quality throughout testing. A properly implemented quality assurance effort will also enable preliminary data taken in this effort to be qualified as NQA-1 during any subsequent licensing discussions for an advanced design or actual prototype. The objective of this report is to describe the quality assurance protocols that will be used for this effort. An essential first step in evaluating quality protocols is assessing the end use of the data. Currently, the advanced structural materials effort is part of a long-range, basic research and development effort and not, as yet, involved in licensing discussions for a specific reactor design. After consultation with Mark Vance (an ORNL QA expert) and based on the recently-issued AFCI QA requirements, the application of NQA-1 quality requirements will follow the guidance provided in Part IV, Subpart 4.2 of the NQA-1 standard (Guidance on Graded Application of QA for Nuclear-Related Research and Development). This guidance mandates the application of sound scientific methodology and a robust peer review process in all phases, allowing for the data to be qualified for use even if the programmatic mission changes to include licensing discussions of a specific design or prototype. ORNL has previously implemented a QA program dedicated to GNEP activities and based on an appropriately graded application of NQA-1 requirements at the site. The current program is being revised to incorporate changes imposed through the recently revised AFCI Technical Integration Office QA requirements. Testing conducted under the AFCI QA program for the advanced structural materials effort shall incorporate the following quality assurance expectations: (1) personnel are adequately trained to perform assigned work; (2) activities are controlled to ensure consistency of results; (3) records necessary to substantiate how the work was performed are maintained (dedicated laboratory notebooks will be used); (4) the pedigree and traceability of the various tested materials are maintained throughout the described processes using consistent sample numbering and adequate record keeping; (5) equipment with the potential to affect the quality of the planned work is calibrated and maintained in accordance with applicable operating requirements. In addition, all reporting or related dissemination by ORNL personnel of the results of the work described in this subcontract shall be conducted in accordance with the requirements described or referenced in the ORNL Standards Based Management System subject area entitled Scientific and Technical Information. Reporting or publications at other institutions will be conducted in accordance with the requirements of that institution. Successful implementation of these protocols will provide a sound basis for future decisions and research. In addition, these steps will also help ensure that results can also be applied to licensing discussions at a future date.« less
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