Effects of V and Carbides on the Temper Embrittlement of the 2.25Cr-1Mo Steel

Kim, Jeong Tae; Kim, Byung Hoon; Kong, Byeong Seo; Kim, Dong Jin

doi:10.1115/pvp2006-icpvt-11-93218

Cited by 3 publications

(4 citation statements)

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“…Clearly, more studies are needed to establish the effect of Si, Ti, B, V, and other impurities on thermal aging and resulting temper embrittlement of 2.25Cr-1Mo-0.25V steels. Very recently, Kim et al [2006] studied the susceptibility to temper embrittlement of 2.25Cr-1Mo and 2.25Cr-1Mo-0.25V steel (V=0.29wt.%) forgings of size 400 mm thick and 2.3 m inner diameter. In the vanadium modified steel, P= 0.006 wt.% while in the standard steel, P≈0.007 wt.…”

Section: Thermal Aging Effects On Mechanical Propertiesmentioning

confidence: 99%

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Preliminary materials selection issues for the next generation nuclear plant reactor pressure vessel.

Natesan

Shankar²,

Shah³

2007

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The NGNP, which is an advanced HTGR concept with emphasis on both electricity and hydrogen production, involves helium as the coolant and a closed-cycle gas turbine for power generation with a core outlet/gas turbine inlet temperature of 900-1000°C. In the indirect cycle system, an intermediate heat exchanger is used to transfer the heat from primary helium from the core to the secondary fluid, which can be helium, nitrogen/helium mixture, or a molten salt. The system concept for the VHTR can be a reactor based on the prismatic block of the GT-MHR developed by a consortium led by General Atomics in the U.S. or based on the PBMR design developed by ESKOM of South Africa and British Nuclear Fuels of U.K.This report has reviewed the available information on candidate materials for the construction of RPV and has made a preliminary assessment of several relevant factors to make a judicious selection of the material for the RPV. Some of the factors addressed in this report are availability of commercial alloys for RPV application, their status in ASME Codes for nuclear service, ASME code compliance during steady state and depressurized conduction cooling conditions, baseline mechanical properties, effects of thermal aging on mechanical properties and code compliance over long term, fabrication and welding issues, global assessment of availability/capability of vendors for RPV procurement, and technology gaps in our knowledge base to address some of the technical issues. The assessment included three primary candidate alloys namely, low alloy steel SA508 (UNS K12042), Fe-2.25Cr-1Mo-0.25V steel (UNS K31835), and modified 9Cr-1Mo steel (UNS K90901) for the RPV. Several conclusions were drawn from this assessment:Baseline Mechanical Properties: There is sufficient database available for the mechanical properties of SA-508 steel. There is limited data available on the thermal aging effects on the mechanical properties and additional information needs development on the long-term aging effects. The steel is approved for use up to 371°C (700°F) under the ASME Code. At present no data is available on the effects of impure helium on the long-term corrosion and mechanical properties of the material.There is adequate tensile data on the Fe-2.25Cr-1Mo-0.25V in the temperature range of interest in NGNP RPV. There is only limited creep data available for the steel and additional data are needed, especially at elevated temperatures that encompass depressurized conduction cooldown conditions. However, substantial data on the creep fatigue properties, thermal aging affects on the mechanical properties, performance characteristics in impure helium, and properties of thick section material are needed prior to its selection for NGNP RPV. Furthermore, the steel is approved under ASME Code Section VIII (for non nuclear applications) and is not approved under Section III for use in nuclear systems.Substantial database on the baseline mechanical properties is currently available for the Modified 9Cr-1Mo steel. Sufficient data are also avail...

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Section: Thermal Aging Effects On Mechanical Propertiesmentioning

confidence: 99%

“…A recent paper by Doosan Heavy Industries (DHI), South Korea [Kim et al 2006] discusses the aging response of a forged 2.3-m inner dia. and 400-mm thick 2.25Cr-1Mo-V material.…”

Section: Fe-225 Cr-1 Mo-v Steelmentioning

confidence: 99%

Preliminary materials selection issues for the next generation nuclear plant reactor pressure vessel.

Natesan

Shankar²,

Shah³

2007

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“…Recently, Kim et al (2006) studied the susceptibility to temper embrittlement of 2.25Cr-1Mo and 2.25Cr-1Mo-0.25V steel (V=0.29 wt.%) forgings of size 400 mm thick and 2.3 m inner diameter. In the vanadium modified steel, P= 0.006 wt.% while in the standard steel, P≈0.007 wt.…”

Section: Thermal Aging Effects On Mechanical Propertiesmentioning

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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“…Lei [ 10 ] determined, through the measurement of internal friction temperature curves, that high-temperature temper embrittlement belongs to the category of α-phase aging and exhibits reversibility. Kim [ 11 ] studied the effects of vanadium and carbides on the temper embrittlement of Cr-Mo steel.…”

Section: Introductionmentioning

confidence: 99%

Regulation Law of Tempering Cooling Rate on Toughness of Medium-Carbon Medium-Alloy Steel

Yang,

Xu,

Zhao

et al. 2023

Materials

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Temper embrittlement is a major challenge encountered during the heat treatment of high-performance steels for large forgings. This study investigates the microstructural evolution and mechanical properties of Cr-Ni-Mo-V thick-walled steel, designed for large forgings with a tensile strength of 1500 MPa, under different tempering cooling rates. Optical microscopy (OM), scanning electron microscopy (SEM), and electron backscatter diffraction (EBSD) were employed to analyze the microstructural features. The results demonstrate that the embrittlement occurring during air cooling after tempering is attributed to the concentration of impurities near Fe3C at the grain boundaries. The low-temperature impact toughness at −40 °C after water quenching reaches 29 J due to the accelerated cooling rate during tempering, which slows down the diffusion of impurity elements towards the grain boundaries, resulting in a reduced concentration and dislocation density and an increased stability of the grain boundaries, thereby enhancing toughness. The bainite content decreases and the interface between martensite and bainite undergoes changes after water quenching during tempering. These alterations influence the crack propagation direction within the two-phase microstructure, further modifying the toughness. These findings contribute to the understanding of temper embrittlement and provide valuable guidance for optimizing heat treatment processes to enhance the performance of high-performance steels in large forgings.

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Effects of V and Carbides on the Temper Embrittlement of the 2.25Cr-1Mo Steel

Cited by 3 publications

References 0 publications

Preliminary materials selection issues for the next generation nuclear plant reactor pressure vessel.

Preliminary materials selection issues for the next generation nuclear plant reactor pressure vessel.

Code qualification of structural materials for AFCI advanced recycling reactors.

Regulation Law of Tempering Cooling Rate on Toughness of Medium-Carbon Medium-Alloy Steel

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