mahmut cinbiz | Penn State University (original) (raw)

Papers by mahmut cinbiz

OSTI OAI (U.S. Department of Energy Office of Scientific and Technical Information), Sep 21, 2022

International Journal of Hydrogen Energy

Microscopy and Microanalysis

Journal of Nuclear Materials

The Minerals, Metals & Materials Series, 2017

This study investigates the failure mechanisms of advanced oxidation resistant FeCrAl nuclear fue... more This study investigates the failure mechanisms of advanced oxidation resistant FeCrAl nuclear fuel cladding at high-strain rates, similar to conditions characteristic of design basis reactivity initiated accidents (RIAs). During a postulated RIA, the nuclear fuel cladding may be subjected to complex loading which can cause multiaxial strain states ranging from plane-strain to equibiaxial tension. To achieve those accident conditions, the samples were deformed by the expansion of high strength Inconel alloy tube under pre-specified pressure pulses, simulating strains rates occurring in a postulated RIA. The mechanical response of the advanced claddings, in the unirradiated state with ample ductility, was compared to that of hydrided zirconium-based nuclear fuel cladding. The hoop strain evolution pulses were collected in situ; the permanent diametral strains of both accident tolerant fuel (ATF) claddings and the current nuclear fuel alloys were determined after rupture. Both zirconium-based alloys and FeCrAl alloys exhibited ductile behavior. FeCrAl model alloys without microstructural control and strengthening mechanism were used in this demonstration study that showed reduced diametral strain (less than 0.15) compared to the diametral strain for the unirradiated zirconium-based alloy (approximately 0.2).

Transactions of the American Nuclear Society - Volume 122, 2020

Journal of Nuclear Materials, 2020

The reactivity initiated accident (RIA) is a postulated accident in light water reactors initiate... more The reactivity initiated accident (RIA) is a postulated accident in light water reactors initiated by a rapid increase in reactivity which causes an increase in fission rate and fuel temperature. One potential mode of fuel system failure during RIA is pellet cladding mechanical interaction (PCMI) due to the rapid thermal expansion of the fuel pellet. Simulated PCMI experiments were performed by rapidly pressurizing iron-chromium-aluminum (FeCrAl) cladding tube samples using a hydraulic modified burst test system, causing the specimens to burst under a biaxial loading path. Deformation and rupture of the specimens were tracked with a telecentric lens and high-speed camera system. Outer surface strains were calculated using digital image correlation (DIC) on speckle patterns painted on the specimens' outer surfaces. Experiments were conducted at approximately 275 C, representative of hot zero power RIA conditions. The failure hoop strain was DIC-calculated between approximately 1.8e3.4%, corresponding to quasi-static energy depositions of approximately 110e260 calories per gram UO 2 assuming initial pellet-cladding contact. These conditions are very similar to the proposed energy deposition limit of 150 calories per gram UO 2 for unirradiated zirconium-based cladding in US Nuclear Regulatory Commission (NRC) Draft Regulatory Guide 1327. A very small strain rate dependence was observed in the data, with the magnitude of the failure hoop strain decreasing slightly with increasing strain rate. This observed dependence may be relevant because the FeCrAl cladding strain rate will be approximately 20% higher than in zirconium-based cladding due to the harder neutron spectrum and resultant shorter neutron generation time.

Journal of Nuclear Materials, 2018

The thermal expansion coefficient of d hydrides and the evolution of the d-spacing of d hydride p... more The thermal expansion coefficient of d hydrides and the evolution of the d-spacing of d hydride platelets in Zircaloy-4 during heat treatments were investigated by conducting synchrotron x-ray diffraction experiments. Identical experiments enabled a direct comparison of the d-spacing of d zirconium hydride platelets with the d-spacing of the powder d hydrides. By analyzing the experimental data of this study and the data available in the literature, the thermal expansion coefficient of pure d hydrides was determined as 14.1 10 À6 C À1. A direct comparison of the d-spacings of d hydride platelets in CWSR Zry-4 sheet and the powder d hydride samples showed an evolution of temperature-dependent strains within the d hydride precipitates in which the strain components normal to the platelet edges exceed that normal to the platelet face during cooling/precipitation but not during heating/dissolution above approx. 200 C.

Annals of Nuclear Energy, 2019

An advanced tube-burst test system was designed to test samples of nuclear fuel cladding under co... more An advanced tube-burst test system was designed to test samples of nuclear fuel cladding under conditions relevant to a postulated design-basis reactivity insertion accident (RIA) in light-water reactors. The system is based on the ''driver tube" concept and allows for high-speed testing with internal pressure impulses of 10-1000 ms. To measure strain, the system was equipped with an ultra-high-speed video camera with a telecentric lens providing high focal depth, to compensate for specimen shift. A mirror system was developed to provide 360°view of the tube specimen into a single camera. Images taken during the test allow the use of a digital image correlation approach in strain evaluation. Several experiments were conducted to assess the performance of the experimental setup, and results are reported.

Journal of Nuclear Materials, 2018

Failure behavior of SiC/SiC composite tubes under strain rates similar to the pellet-cladding mec... more Failure behavior of SiC/SiC composite tubes under strain rates similar to the pellet-cladding mechanical interaction phase of reactivity-initiated accidents,

Annals of Nuclear Energy, 2017

Pellet-cladding mechanical interaction due to thermal expansion of nuclear fuel pellets during a ... more Pellet-cladding mechanical interaction due to thermal expansion of nuclear fuel pellets during a reactivity-initiated accident (RIA) is a potential mechanism for failure of nuclear fuel cladding. To investigate the mechanical behavior of cladding during an RIA, we developed a mechanical pulse-controlled modified-burst test instrument that simulates transient events with a pulse width from 10 to 300 ms. This paper includes validation tests of unirradiated and pre-hydrided ZIRLO TM cladding tubes. A ZIRLO TM cladding sample with 168 wt. ppm of hydrogen showed ductile behavior and failed at the maximum limits of the mechanical test setup; hoop strain to failure was greater than 9.2%. ZIRLO TM samples showed high resistance to failure even at very high hydrogen contents (1466 wt. ppm). When the hydrogen content was increased to 1554 wt. ppm, ''brittle-like" behavior was observed at a hoop strain of 2.5%. Preliminary scoping tests at room temperature of tubes fabricated from FeCrAl, a candidate material for accident-tolerant cladding were conducted to reproduce the pulse behavior of transient test reactors during integral tests. The preliminary FeCrAl tests are informative from the perspective of characterizing the test rig and supporting the design of RIA integral tests for current cladding materials and for candidate accident-tolerant cladding materials.

Journal of Alloys and Compounds, 2017

The mechanical properties of zirconium hydrides were studied using nano-indentation technique bet... more The mechanical properties of zirconium hydrides were studied using nano-indentation technique between 25 and 400 C. Temperature dependency of reduced elastic modulus and hardness of dand ε-zirconium hydrides were obtained by conducting nanoindentation experiments on bulk hydride samples with independently heating capability of indenter and heating stage. The reduced elastic modulus of d-zirconium hydride (H/Zr ratio ¼ 1.61) decreased from~113 GPa at room temperature tõ 109 GPa at 400 C, while its hardness decreased significantly from 4.1 GPa to 2.41 GPa in the same temperature range. For ε-zirconium hydrides (H/Zr ratio ¼ 1.79), the reduced elastic modulus decreased from 61 GPa at room temperature to 54 GPa at 300 C, while its hardness from 3.06 GPa to 2.19 GPa.

Nuclear Technology, Oct 17, 2022

OSTI OAI (U.S. Department of Energy Office of Scientific and Technical Information), Sep 21, 2022

International Journal of Hydrogen Energy

Microscopy and Microanalysis

Journal of Nuclear Materials

The Minerals, Metals & Materials Series, 2017

This study investigates the failure mechanisms of advanced oxidation resistant FeCrAl nuclear fue... more This study investigates the failure mechanisms of advanced oxidation resistant FeCrAl nuclear fuel cladding at high-strain rates, similar to conditions characteristic of design basis reactivity initiated accidents (RIAs). During a postulated RIA, the nuclear fuel cladding may be subjected to complex loading which can cause multiaxial strain states ranging from plane-strain to equibiaxial tension. To achieve those accident conditions, the samples were deformed by the expansion of high strength Inconel alloy tube under pre-specified pressure pulses, simulating strains rates occurring in a postulated RIA. The mechanical response of the advanced claddings, in the unirradiated state with ample ductility, was compared to that of hydrided zirconium-based nuclear fuel cladding. The hoop strain evolution pulses were collected in situ; the permanent diametral strains of both accident tolerant fuel (ATF) claddings and the current nuclear fuel alloys were determined after rupture. Both zirconium-based alloys and FeCrAl alloys exhibited ductile behavior. FeCrAl model alloys without microstructural control and strengthening mechanism were used in this demonstration study that showed reduced diametral strain (less than 0.15) compared to the diametral strain for the unirradiated zirconium-based alloy (approximately 0.2).

Transactions of the American Nuclear Society - Volume 122, 2020

Journal of Nuclear Materials, 2020

The reactivity initiated accident (RIA) is a postulated accident in light water reactors initiate... more The reactivity initiated accident (RIA) is a postulated accident in light water reactors initiated by a rapid increase in reactivity which causes an increase in fission rate and fuel temperature. One potential mode of fuel system failure during RIA is pellet cladding mechanical interaction (PCMI) due to the rapid thermal expansion of the fuel pellet. Simulated PCMI experiments were performed by rapidly pressurizing iron-chromium-aluminum (FeCrAl) cladding tube samples using a hydraulic modified burst test system, causing the specimens to burst under a biaxial loading path. Deformation and rupture of the specimens were tracked with a telecentric lens and high-speed camera system. Outer surface strains were calculated using digital image correlation (DIC) on speckle patterns painted on the specimens' outer surfaces. Experiments were conducted at approximately 275 C, representative of hot zero power RIA conditions. The failure hoop strain was DIC-calculated between approximately 1.8e3.4%, corresponding to quasi-static energy depositions of approximately 110e260 calories per gram UO 2 assuming initial pellet-cladding contact. These conditions are very similar to the proposed energy deposition limit of 150 calories per gram UO 2 for unirradiated zirconium-based cladding in US Nuclear Regulatory Commission (NRC) Draft Regulatory Guide 1327. A very small strain rate dependence was observed in the data, with the magnitude of the failure hoop strain decreasing slightly with increasing strain rate. This observed dependence may be relevant because the FeCrAl cladding strain rate will be approximately 20% higher than in zirconium-based cladding due to the harder neutron spectrum and resultant shorter neutron generation time.

Journal of Nuclear Materials, 2018

The thermal expansion coefficient of d hydrides and the evolution of the d-spacing of d hydride p... more The thermal expansion coefficient of d hydrides and the evolution of the d-spacing of d hydride platelets in Zircaloy-4 during heat treatments were investigated by conducting synchrotron x-ray diffraction experiments. Identical experiments enabled a direct comparison of the d-spacing of d zirconium hydride platelets with the d-spacing of the powder d hydrides. By analyzing the experimental data of this study and the data available in the literature, the thermal expansion coefficient of pure d hydrides was determined as 14.1 10 À6 C À1. A direct comparison of the d-spacings of d hydride platelets in CWSR Zry-4 sheet and the powder d hydride samples showed an evolution of temperature-dependent strains within the d hydride precipitates in which the strain components normal to the platelet edges exceed that normal to the platelet face during cooling/precipitation but not during heating/dissolution above approx. 200 C.

Annals of Nuclear Energy, 2019

An advanced tube-burst test system was designed to test samples of nuclear fuel cladding under co... more An advanced tube-burst test system was designed to test samples of nuclear fuel cladding under conditions relevant to a postulated design-basis reactivity insertion accident (RIA) in light-water reactors. The system is based on the ''driver tube" concept and allows for high-speed testing with internal pressure impulses of 10-1000 ms. To measure strain, the system was equipped with an ultra-high-speed video camera with a telecentric lens providing high focal depth, to compensate for specimen shift. A mirror system was developed to provide 360°view of the tube specimen into a single camera. Images taken during the test allow the use of a digital image correlation approach in strain evaluation. Several experiments were conducted to assess the performance of the experimental setup, and results are reported.

Journal of Nuclear Materials, 2018

Failure behavior of SiC/SiC composite tubes under strain rates similar to the pellet-cladding mec... more Failure behavior of SiC/SiC composite tubes under strain rates similar to the pellet-cladding mechanical interaction phase of reactivity-initiated accidents,

Annals of Nuclear Energy, 2017

Pellet-cladding mechanical interaction due to thermal expansion of nuclear fuel pellets during a ... more Pellet-cladding mechanical interaction due to thermal expansion of nuclear fuel pellets during a reactivity-initiated accident (RIA) is a potential mechanism for failure of nuclear fuel cladding. To investigate the mechanical behavior of cladding during an RIA, we developed a mechanical pulse-controlled modified-burst test instrument that simulates transient events with a pulse width from 10 to 300 ms. This paper includes validation tests of unirradiated and pre-hydrided ZIRLO TM cladding tubes. A ZIRLO TM cladding sample with 168 wt. ppm of hydrogen showed ductile behavior and failed at the maximum limits of the mechanical test setup; hoop strain to failure was greater than 9.2%. ZIRLO TM samples showed high resistance to failure even at very high hydrogen contents (1466 wt. ppm). When the hydrogen content was increased to 1554 wt. ppm, ''brittle-like" behavior was observed at a hoop strain of 2.5%. Preliminary scoping tests at room temperature of tubes fabricated from FeCrAl, a candidate material for accident-tolerant cladding were conducted to reproduce the pulse behavior of transient test reactors during integral tests. The preliminary FeCrAl tests are informative from the perspective of characterizing the test rig and supporting the design of RIA integral tests for current cladding materials and for candidate accident-tolerant cladding materials.

Journal of Alloys and Compounds, 2017

The mechanical properties of zirconium hydrides were studied using nano-indentation technique bet... more The mechanical properties of zirconium hydrides were studied using nano-indentation technique between 25 and 400 C. Temperature dependency of reduced elastic modulus and hardness of dand ε-zirconium hydrides were obtained by conducting nanoindentation experiments on bulk hydride samples with independently heating capability of indenter and heating stage. The reduced elastic modulus of d-zirconium hydride (H/Zr ratio ¼ 1.61) decreased from~113 GPa at room temperature tõ 109 GPa at 400 C, while its hardness decreased significantly from 4.1 GPa to 2.41 GPa in the same temperature range. For ε-zirconium hydrides (H/Zr ratio ¼ 1.79), the reduced elastic modulus decreased from 61 GPa at room temperature to 54 GPa at 300 C, while its hardness from 3.06 GPa to 2.19 GPa.

Nuclear Technology, Oct 17, 2022