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

Natesan, Kapilan; Shankar, Prem; Shah, V.N.

doi:10.2172/925328

Cited by 17 publications

(13 citation statements)

References 43 publications

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“…Several DOE-NE programs, such as the FCRD, 1 ARC, 2,3a LWRS, 4 and NGNP [5][6][7][8] programs, are investigating new fuels and materials for advanced and existing reactors. A key objective of such programs is to understand the performance of these fuels and materials when irradiated.…”

Section: Fuels and Materials Irradiationsmentioning

confidence: 99%

“…8 shows examples of metallic fuel swelling and restructuring and indicates that swelling and restructuring occur early in irradiation. Short-term irradiations of MOX fuel and minor actinide-containing MOX fuel conducted by Japan Atomic Energy Agency (JAEA) show that restructuring and redistribution begins within the first 24 hours of irradiation and can even be seen as early as 10 minutes 10.…”

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confidence: 99%

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NEET In-Pile Ultrasonic Sensor Enablement-FY 2012 Status Report

Daw¹,

Rempe²,

Tittmann³

et al. 2012

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Several programs funded by the Department of Energy Office of Nuclear Energy (DOE-NE), such as the Fuel Cycle Research and Development, Advanced Reactor Concepts, Light Water Reactor Sustainability, and Next Generation Nuclear Plant programs, are investigating new fuels and materials for advanced and existing reactors. A key objective of such programs is to understand the performance of these fuels and materials when irradiated. The Nuclear Energy Enabling Technology (NEET) Advanced Sensors and Instrumentation (ASI) in-pile instrumentation development activities are focused upon addressing cross-cutting needs for DOE-NE irradiation testing by providing higher fidelity, real-time data, with increased accuracy and resolution from smaller, compact sensors that are less intrusive. Ultrasonic technologies offer the potential to measure a range of parameters, including geometry changes, temperature, crack initiation and growth, gas pressure and composition, and microstructural changes, under harsh irradiation test conditions. There are two primary issues that current limit in-pile deployment of ultrasonic sensors. The first is transducer survivability. The ability of ultrasonic transducer materials to maintain their useful properties during an irradiation must be demonstrated. The second issue is signal processing. Ultrasonic testing is typically performed in a lab or field environment, where the sensor and sample are accessible. Due to the harsh nature of in-pile testing, and the variety of measurements that are desired, an enhanced signal processing capability is needed to make in-pile ultrasonic sensors viable. The NEET ASI program is funding the Ultrasonic Transducer Irradiation and Software Enhancements project, which is a collaborative effort between the Idaho National Laboratory,

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Section: Fuels and Materials Irradiationsmentioning

confidence: 99%

mentioning

confidence: 99%

NEET In-Pile Ultrasonic Sensor Enablement-FY 2012 Status Report

Daw¹,

Rempe²,

Tittmann³

et al. 2012

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“…A report entitled "Preliminary Materials Selection issues for the next Generation Nuclear Plant Reactor Pressure Vessel" (Natesan et al, September 2006) [10] reviewed the available information on candidate materials for the construction of the RPV and made a preliminary assessment of several factors relevant to judicious material selection. Some of the factors addressed in this report were, baseline mechanical properties, including the effects of thermal aging, impure helium and thick sections; availability of commercial alloys, with a global assessment of potential RPV vendors; welding issues; and ASME code compliance, including the status of candidate alloys in the ASME Codes for nuclear service and an evaluation of suitability during steady state and depressurized conduction cool-down (DCC) conditions.…”

Section: Reactor Pressure Vesselmentioning

confidence: 99%

Next Generation Nuclear Plant Materials Research and Development Program Plan, Revision 4

Hayner¹,

Bratton²,

Mizia³

et al. 2007

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Executive SummaryDOE has selected the High Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 950°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic or pebble-bed reactor and use low-enriched uranium, TRISO-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development (R&D) Program is responsible for performing R&D on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Some of the general and administrative aspects of the R&D Plan include:Expand American Society of Mechanical Engineers (ASME) Codes and American Society for Testing and Materials (ASTM) Standards in support of the NGNP Materials R&D Program.Define and develop inspection needs and the procedures for those inspections.Support selected university materials related R&D activities that would be of direct benefit to the NGNP Project. GraphiteThe Next Generation Nuclear Plant (NGNP) will be a helium cooled High Temperature Gas Reactor (HTGR) with a large graphite core. Graphite physically contains the fuel and comprises the majority of the core volume. Graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gascooled reactors. This development has resulted in graphite being established as a viable structural material for HTGRs. While the general characteristics necessary for producing nuclear grade graphite are understood, historical "nuclear" grades no longer exist. New grades must be fabricated, characterized and irradiated to demonstrate that current grades of graphite exhibit acceptable non-irradiated and irradiated properties upon which the thermomechanical design of the structural graphite in NGNP is based. This technology development plan establishes the research and development activities and associated rationale necessary to qualify nuclear grade graphite for use within the Next Generation Nuclear Plant (NGNP) reactor.Background information from past graphite reactor experience, other relevant graphite grades, and the technology developed for past gas reactors is presented to provide a perspective on what has been achieved previously in this area of research. Next the technology required to qualify the graphite for use in NGNP will be developed based on the historical graphite fabrication and performance database, the anticipated NGNP graphite design service conditions, and gaps in the fabrication and performance database. The resultant data needs are outlined and justified from the perspective of reactor design, reactor performance, or the reactor safety case. The approach will allow direct comparison between data...

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“…To replicate the industrial large open die forging operation of SA508 grade 3 steel components, isothermal uniaxial compression tests were carried out over a specific range of test temperatures (880-1130°C) and true strain rates (0.001-1/s) with a true strain of unity. The standard strain rates in any open die forging process are typically in the range of 1-500/s (Ref 7); however, it is particularly in the range of 0.001-10/s for various steels (including the one utilized in the current study) used in the nuclear plant RPVs ( Ref 4,8,9). There is a continuous drive to improve the existing numerical models for predicting accurate strains and microstructure evolution during hot forging processes.…”

Section: Introductionmentioning

confidence: 99%

Microstructural Evolution of SA508 Grade 3 Steel during Hot Deformation

Mandal

Lalvani

Barrow

et al. 2020

J. of Materi Eng and Perform

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SA508 grade 3 steel is widely used in the manufacture of large-scale forged components for nuclear reactor applications. Numerical models have already been established to simulate industrial forging process of grade 3 steel; however, limited information is available on the microstructural evolution of this steel during hot forging operation. This work focuses on the flow behavior and related microstructural evolution in grade 3 steel with detailed analysis on the interfacial friction, texture and hardness evolution. Uniaxial hot compression tests were conducted over a range of test temperatures (880-1130°C) and true strain rates (0.001-1/s), representative of the industrial hot forging conditions. Two different deformation mechanisms, MDRX at the lowest forging temperature and DRV along with DRX at the highest forging temperature, were observed showing marked impact on the final microstructure and hardness. A random fiber-type weak deformation texture was observed irrespective of the test temperatures and strain rates used. The microstructural changes from the as-received to the various deformed conditions were quantified. The quantitative data are the key to obtain accurate parameters for DRV and DRX processes that affect the accuracy of the mathematical models.

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

Cited by 17 publications

References 43 publications

NEET In-Pile Ultrasonic Sensor Enablement-FY 2012 Status Report

NEET In-Pile Ultrasonic Sensor Enablement-FY 2012 Status Report

Next Generation Nuclear Plant Materials Research and Development Program Plan, Revision 4

Microstructural Evolution of SA508 Grade 3 Steel during Hot Deformation

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