NSUF 15-530: Characterization of CANDU Core Internals via Small Scale Mechanical Testing

This project intends to utilize small scale mechanical testing methods, i.e. micro-compression, micro-tensile, and micro-bend bar in order to obtain a quantitative understanding in terms of yield strength, work hardening rate, or failure mode, for example, of the degradation of mechanical properties of LWR core materials composed of nickel alloys after many years of service. This has previously been limited by the dearth of test data from test reactor experiments, but creating multiple, small test specimens from ex-service components provides a plethora of information that is both pertinent to a specific part of a given component and is not limited by the limitations of irradiating specifically fabricated surrogate materials test samples. In addition to obtaining fundamental knowledge about materials deformation processes, SSMT allows for the manufacture of specimens from specific regions of interest on both the macroscopic and microscopic scale. One can manufacture specimens from regions of the component operating under specific conditions of stress, temperature, and neutron flux. On the microscopic scale, one can select microscopic parts of deformed or welded sections of a component and grains or grain boundaries of specific orientation. Because the samples are on the order of microns in size, radiation exposure is limited and safety is enhanced. It is our goal to apply these methods to a large scale engineering problem, i.e. the weakening and embrittlement of nickel alloy components in power reactors due to He accumulation and enhanced radiation damage rates in a CANDU reactor that bound the conditions that may be experienced in any PWR or BWR environment. It is critical for the nuclear industry as a whole to be able to quantify the rate of decrease in the failure stress in comparison to the unirradiated state. Whereas embrittlement has been observed previously for nickel alloys irradiated at high temperatures greater than 500 oC, there have been few instances where power reactor components have been shown to exhibit similar degrees of embrittlement. What was once dismissed as a high temperature phenomenon is turning out to be exhibited by power reactor components, and it is essential to understand the failure mechanism and the rate of degradation in order to make long term predictions of mechanical failure through appropriate models. Multiple SSMT on parts of the CANDU Inconel X-750 garter spring, where temperature is the only variable, will provide valuable insight into the temperature dependence of embrittlement at power reactor operating temperatures. This understanding can then be applied to core components subject to He embrittlement in all reactor types. In conclusion, our work scope will obtain a qualitative and quantitative insight into low temperature, below 350 oC, He embrittlement and develop a novel technique for studying the degradation in mechanical properties applicable to all nuclear reactor core components. This work will take place over a six month period with all of the heavy duty hot FIB fabrication work taking place at INL and testing of the micro specimens at the UC Berkeley BNC user facility, concluding with a comprehensive report to the ATR-NSUF.

Additional Info

Field	Value
Abstract	This project intends to utilize small scale mechanical testing methods, i.e. micro-compression, micro-tensile, and micro-bend bar in order to obtain a quantitative understanding in terms of yield strength, work hardening rate, or failure mode, for example, of the degradation of mechanical properties of LWR core materials composed of nickel alloys after many years of service. This has previously been limited by the dearth of test data from test reactor experiments, but creating multiple, small test specimens from ex-service components provides a plethora of information that is both pertinent to a specific part of a given component and is not limited by the limitations of irradiating specifically fabricated surrogate materials test samples. In addition to obtaining fundamental knowledge about materials deformation processes, SSMT allows for the manufacture of specimens from specific regions of interest on both the macroscopic and microscopic scale. One can manufacture specimens from regions of the component operating under specific conditions of stress, temperature, and neutron flux. On the microscopic scale, one can select microscopic parts of deformed or welded sections of a component and grains or grain boundaries of specific orientation. Because the samples are on the order of microns in size, radiation exposure is limited and safety is enhanced. It is our goal to apply these methods to a large scale engineering problem, i.e. the weakening and embrittlement of nickel alloy components in power reactors due to He accumulation and enhanced radiation damage rates in a CANDU reactor that bound the conditions that may be experienced in any PWR or BWR environment. It is critical for the nuclear industry as a whole to be able to quantify the rate of decrease in the failure stress in comparison to the unirradiated state. Whereas embrittlement has been observed previously for nickel alloys irradiated at high temperatures greater than 500 oC, there have been few instances where power reactor components have been shown to exhibit similar degrees of embrittlement. What was once dismissed as a high temperature phenomenon is turning out to be exhibited by power reactor components, and it is essential to understand the failure mechanism and the rate of degradation in order to make long term predictions of mechanical failure through appropriate models. Multiple SSMT on parts of the CANDU Inconel X-750 garter spring, where temperature is the only variable, will provide valuable insight into the temperature dependence of embrittlement at power reactor operating temperatures. This understanding can then be applied to core components subject to He embrittlement in all reactor types. In conclusion, our work scope will obtain a qualitative and quantitative insight into low temperature, below 350 oC, He embrittlement and develop a novel technique for studying the degradation in mechanical properties applicable to all nuclear reactor core components. This work will take place over a six month period with all of the heavy duty hot FIB fabrication work taking place at INL and testing of the micro specimens at the UC Berkeley BNC user facility, concluding with a comprehensive report to the ATR-NSUF.
Award Announced Date	2014-12-04T00:00:00
Awarded Institution	None
Facility	None
Facility Tech Lead	Yaqiao Wu
Irradiation Facility	None
PI	Peter Hosemann
PI Email	[email protected]
Project Type	RTE
RTE Number	530