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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 163 records · Page 9
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

TEM Study of Zircaloy 2 with FeCrAl Layer under Simulated BWR Environment

Zircaloy 2 is widely used as nuclear fuel cladding in light-water nuclear reactors, which is resistive to corrosion, creep, and swelling. Due to water chemical reactivity and high-temperature steam exposure, oxidation in zirconium cladding is possible during transients. To mitigate the zirconium oxidation and the hydrogen generation at elevated temperature, we have coated Zircaloy 2 with oxide resistant FeCrAl thin film layers using magnetron sputtering. Transmission electron microscopy (TEM), scanning transmission electron microscopy (STEM), and electron energy-dispersive X-ray spectroscopy (EDX) are used to understand the behavior of the protective FeCrAl layer under typical BWR operation conditions. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Features of temperature control of fuel element cladding for pressurized water nuclear reactor “WWER-1000” while simulating reactor accidents

During the experiments simulating NPR (nuclear power reactor) accidents with a coolant loss fuel elements behavior in a steam-hydrogen medium was studied at the temperature changed with the rate from 1 to 100K/s within the range of 300÷1500 °C. Indications of the thermocouples fixed on the cladding notably differ from real values of the cladding temperatures in the area of measuring junction due to thermal resistance influence of the transition zones “cladding-junction” and “junction-coolant”. The estimating method of a measurement error was considered which can provide adequate accounting of the influence factors. The method is based on thermal probing of a thermocouple by electric current flashing through thermoelements under the coolant presence or absence, a response time registration and processing, calculation of thermal inertia value for a thermocouple junction. A formula was derived for calculation of methodical error under stationary mode and within the stage of linear increase in temperature, which will determine the conditions for the cladding depressurization. Some variants of the formula application were considered, and the values of methodical errors were established which reached ∼5% of maximum value by the final moment of the stage of linear increase in the temperature.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Development and property evaluation of nuclear grade wrought FeCrAl fuel cladding for light water reactors

Development of nuclear grade, iron-based wrought FeCrAl alloys has been initiated for light water reactor (LWR) fuel cladding to serve as a substitute for zirconium-based alloys with enhanced accident tolerance. Ferritic alloys with sufficient chromium and aluminum additions can exhibit significantly improved oxidation kinetics in high-temperature steam environments when compared to zirconium-based alloys. In the first phase, a set of model FeCrAl alloys containing 10–20Cr, 3–5Al, and 0–0.12Y in weight percent, were prepared by conventional arc-melting and hot-working processes to explore the effect of composition on the properties of FeCrAlY alloys. It was found that the tensile properties were insensitive to the alloy compositions studied; however, the steam oxidation resistance strongly depended on both the chromium and the aluminum contents. The second phase development focused on strengthening Fe-13Cr-5Al with minor alloying additions of molybdenum, niobium, and silicon. Combined with an optimized thermo-mechanical treatment, a thermally stable microstructure was produced with improved tensile properties at temperatures up to 741°C.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Silicon carbide reinforced zirconium based cladding

A method for making an improved nuclear fuel cladding tube includes reinforcing a Zr alloy tube by first winding or braiding ceramic yarn directly around the tube to form a ceramic covering, then physically bonding the ceramic covering to the tube by applying a first coating selected from the group consisting of Nb, Nb alloy, Nb oxide, Cr, Cr oxide, Cr alloy, or combinations thereof, by one of a thermal deposition process or a physical deposition process to provide structural support member for the Zr tube, and optionally applying a second coating and optionally applying a third coating by one of a thermal deposition process or a physical deposition process. If the tube softens at 800° C.-1000° C., the structural support tube will reinforce the Zr alloy tube against ballooning and bursting, thereby preventing the release of fission products to the reactor coolant.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Light Water Reactor Sustainability Program Status of Silicon Carbide Joining Technology Development

Advanced, accident tolerant nuclear fuel systems are currently being investigated for potential application in currently operating light water reactors (LWR) or in reactors that have attained design certification. Evaluation of potential options for accident tolerant nuclear fuel systems point to the potential benefits of silicon carbide (SiC) relative to Zr-based alloys, including increased corrosion resistance, reduced oxidation and heat of oxidation, and reduced hydrogen generation under steam attack (off-normal conditions). If demonstrated to be applicable in the intended LWR environment, SiC could be used in nuclear fuel cladding or other in-core structural components. Achieving a SiC-SiC joint that resists corrosion with hot, flowing water, is stable under irradiation and retains hermeticity is a significant challenge. This report summarizes the current status of SiC-SiC joint development work supported by the Department of Energy Light Water Reactor Sustainability Program. Significant progress has been made toward SiC-SiC joint development for nuclear service, but additional development and testing work (including irradiation testing) is still required to present a candidate joint for use in nuclear fuel cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Milestone 2.8: Preliminary Radiolytic Gas Generation Measurements from Helium-Backfilled Samples

The Department of Energy (DOE) is currently managing nearly 13 metric tons of aluminum-clad spent nuclear fuel (ASNF) with the intention of extended (> 50 years) dry storage in helium-backfilled canisters. Due to in-reactor and cooling pond conditions, oxyhydroxide corrosion layers have formed on the surface of the ASNF elements. These corrosion layers are susceptible to radiolysis and the formation of molecular hydrogen gas (H 2 ) due to the fuel’s inherent radiation field. Consequently, a rigorous evaluation of the effect of helium gas on radiolytic H 2 production is necessary to support the Technical Considerations and Challenges for Extended (> 50 yrs) Dry Storage of ASNF program, especially as previous Task 2 - Oxyhydroxide Layer Radiolytic Gas Generation Resolution work demonstrated a significant effect of gas composition on the radiolytic yield (G-value) of H 2 . Here we report preliminary G-values for the radiolytic formation of H 2 from the gamma irradiation of pre-corroded aluminum alloy 1100 coupons flame sealed in helium environments. Irradiations yielded G(H 2 ) values of (5.1 ± 0.5) x 10 –4 and (9.4 ± 0.9) x 10 –4 µmol J –1 for pristine coupons, and (10.1 ± 0.4) x 10 –4 and (15.1 ± 1.2) x 10 –4 µmol J –1 for pre-corroded coupons for 0% and 50% relative humidity, respectively. These helium environment G(H 2 ) values are between 28% and 58% higher than previously reported values for argon environments. This enhancement is attributed to the significant difference in first ionization energy between helium (24.59 eV) and argon (15.76) facilitating additional processes, e.g., Penning ionization. These new preliminary helium environment G(H 2 ) values will be employed by Task 3 - Sealed and Vented System Episodic Breathing and Gas Generation Prediction to model the effect of radiolytic H 2 accumulation in helium environments to evaluate the practicality of the extended storage standard canister design.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2022

Milestone 1.2.11: H 2 Production from Surrogate Non-Native Corrosion Plumes on Aluminum 6061-T6 Fuel Cladding Surrogates

Thick, localized, “non-native” corrosion plumes have been identified on Advanced Test Reactor fuel elements, raising concern on their impact on the radiolytic formation of molecular hydrogen gas (H 2 ) from aluminum-clad spent nuclear fuel (ASNF) under proposed extended (> 50 years) dry storage conditions. Here, we report our findings on H 2 generation from the gamma irradiation (up to 52 MGy) of surrogate “non-native” corrosion plume coupons: ambient-temperature-corroded (~350 days in water) aluminum alloy 6061 (AA6061) coupons in helium gas environments with ~0% added relative humidity. Additionally, we provide a comparison of proposed ASNF drying techniques— vacuum drying only, vacuum drying + 100 °C for 4 hr, and vacuum drying + 220 °C for 4 hr—on the yield of H 2 from these surrogate systems. The presented data indicates that similar amounts of H 2 (~2 × 10–3 µmol J–1) are formed from gamma irradiated AA6061 coupons corroded under different temperature regimes, i.e., ambient/350 days vs. 90 C/30 days. These findings validate current, complimentary modelling predictions based on high-temperature-corrosion irradiation data only. Further, the application of a heat treatment procedure (100 and 220 °C), in conjunction with vacuum drying, accelerated the rate at which a steady-state H 2 yield was attained, in comparison to vacuum only, due to the removal of H 2 precursors in the form of adsorbed waters. Interestingly, within the confidence limits of our measurements, negligible difference in total H 2 yield was found between the two investigated heat treatment procedures.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2016

Synchrotron Characterization of the Oxidation of Advanced Steel Cladding Alloys

Advanced steel alloys are proposed as nuclear fuel claddings due to their high corrosion resistance under extreme conditions of pressure and temperature. APMT is an iron based alloy (Fe-22wt.%Cr-5.8wt.%Al) and Alloy 33 is a mixture of iron, chromium, and nickel (32wt.%Fe-33wt.%Cr-31wt.%Ni). This paper focuses on studying the corrosion mechanisms for these alloys under exposure to superheated steam at 800 deg. C and 1000 deg. C using the synchrotron based X-ray diffraction (XRD) and Hard X-ray photoelectron spectroscopy (HAXPES) techniques. HAXPES data was collected using 4 keV X-rays with a 1x1 mm spot size at X24A beamline. Specimens were mounted on the sample holder. The specimens were tilted 5 degrees relative to the beam and the estimated probe depth (for this fixed energy) is 63 A. The X-ray wavelength for XRD studies at X14A beamline is 0.7792 A with a beam size of of 2 x 0.2 mm{sup 2}. Specimens were exposed to steam environment at 800 and 1000 deg. C for 8, 24, and 48 hours. For Alloy 33; a layer of mixed Cr{sub 2}O{sub 3} and FeCr{sub 2}O{sub 4} oxides is formed after exposure to steam. Specimen tilting indicated that Cr{sub 2}O{sub 3} is formed near the metal/oxide interface and FeCr{sub 2}O{sub 4} was formed on the oxide surface. The oxide peak intensities increased with an increase in steam temperature. The specimen oxidized at 1000 deg. C had well-developed oxide peaks compared to the specimen oxidized at 800 deg. C. For APMT; a layer of mixed α-Al{sub 2}O{sub 3} and monoclinic θ-Al{sub 2}O{sub 3} oxides is formed on the outer surface. The peak intensities of the monoclinic θ-Al{sub 2}O{sub 3} phase, decreases relative to the peaks from the Al{sub 2}O{sub 3} phase with increasing tilt angle. This indicates that the surface is dominated by the θ-Al{sub 2}O{sub 3}. After exposure to steam at 800 deg. C, oxidation of Alloy 33 results in the formation of Cr{sub 2}O{sub 3} oxide with time and absence of both metallic and substoichiometric phases is observed. No nickel peaks are seen on the surface after oxidation at 800 deg. C, however Fe(II) peak is noticed after 8 and 24 hours of steam exposure, which then disappears after 48 hours. This presence of FeCr{sub 2}O{sub 4} oxide phase is observed after 8 and 24 hours. After exposure to steam at 1000 deg. C, the Cr{sub 2}p{sub 3/2} doublet persists at all exposure times, indicating the presence of the FeCr{sub 2}O{sub 4} oxide. Nickel peak is found on the surface at all exposure times and is attributed to the presence of Ni(OH){sub 2} phase. This may be due to increased diffusion of Ni through chromic oxide scale at higher temperature as compared to 800 deg. C. This nickel reacts with steam at 1000 deg. C to form Ni hydroxide at the surface. For APMT, after exposure to steam at 800 and 1000 deg. C, no Al and Cr metal peaks are seen. Growth of oxide layer is seen with increase in exposure time. The oxide structure contains mainly Al{sub 2}O{sub 3} oxide with the addition of mixed oxide (Al{sub x}Cr{sub 1-x}){sub 2}O{sub 3}. These results demonstrate that Cr diffuses into the alumina layer. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

In Situ Sample Environment for Studying Corrosion of Nuclear Materials

Currently there is lack of fundamental understanding of kinetics and reaction mechanisms that occur at nuclear cladding-steam interface under extreme environments in the nuclear reactors. Such understanding is vital to design and develop advanced cladding materials for more economical and safer nuclear power. We present advanced in situ sample environment for probing solid-fluid interfaces using high resolution synchrotron methods for studying early stages of corrosion including development of oxide layer as corrosion proceeds. Both X-ray diffraction and X-ray fluorescence methods will be discussed to probe structural and elemental composition changes under steam environment using corrosion resistant cell at high temperatures. High intensity and brightness of synchrotron coupled with advanced in situ sample environment developed, presents an opportunity to understand the fundamental interfacial processes involved in degradation of materials. The in situ sample environment is optimized for XRD and XRF data collection at NSLS-II and will be available to all users through the proposal process. The in situ sample environment is built to study the interfacial interactions at temperatures and pressures up to 400 deg. C and 1500 psi. It consists of three main parts; the distribution manifold, the pressure vessel, and the reaction cell as shown in Figure 1. Water is allowed to heat up to 350 deg. C inside the pressure vessel to produce steam which is delivered to reaction cell using insulated delivery system. The cell design is adapted from J. Diefenbacher, et. al. However, it is modified for corrosion applications in steam environments. The reaction cell and sample holder are made from hastelloy due to its high corrosion resistance. The reaction cell is surrounded by copper heating block to control its temperature independently from the rest of the system. Lab view based platform is developed for remote monitoring of pressure and temperature. The sample environment was commissioned at 6-IDD beamline in the Advanced Photon Source, for grazing incidence XRD corrosion studies. Metallic Zr sample (200 μm thickness) was used due to its faster oxidation kinetics as compared to Zircaloy. Sample was exposed to steam at 350 deg. C and 50 psi for a total time of 64 hours. Results showed the presence of α-Zr as the dominant phase. After 32 hours of exposure to steam, three oxide peaks in the low diffraction angle side started to develop with time. These peaks are attributed to the growth of substoichiometric zirconium oxide phases (Zr{sub 3}O and Zr{sub 3}O{sub 1-x}). (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2022

Designing Cladded Components for High Temperature Nuclear Service: Part I—Analysis Methods

Use of corrosion-resistant cladding can greatly extend the design life of structural components in many advanced reactor systems. However, there are currently no ASME design rules for cladded components to guard against creep-fatigue failure and ratcheting strain accumulation in elevated temperature nuclear service. This paper, presented in two parts, addresses this gap by proposing a design strategy for cladded components that do not require long-term testing of clad materials. The proposed approach relies on approximate design analysis methods for two types of clad materials—a soft clad that creeps much faster than and has lower yield stress than the class A base material and a hard clad that creeps much slower than and has higher yield stress than the class A base material. Part I discusses the approximate analysis strategies for the clad materials—treat a soft clad as perfectly compliant and a hard clad as linear elastic—and Part II develops a complete set of design rules for each of the two types of cladded components. In conclusion, finite element analyses of representative high temperature reactor components show that the proposed design analysis methods can bound the design quantities in soft cladded components and approximate the design quantities in hard cladded components.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Characterization of LWRS Hybrid SiC-CMC-Zircaloy-4 Fuel Cladding after Gamma Irradiation

The purpose of the gamma irradiation tests conducted at the Idaho National Laboratory (INL) was to obtain a better understanding of chemical interactions and potential changes in microstructural properties of a mock-up hybrid nuclear fuel cladding rodlet design (unfueled) in a simulated PWR water environment under irradiation conditions. The hybrid fuel rodlet design is being investigated under the Light Water Reactor Sustainability (LWRS) program for further development and testing of one of the possible advanced LWR nuclear fuel cladding designs. The gamma irradiation tests were performed in preparation for neutron irradiation tests planned for a silicon carbide (SiC) ceramic matrix composite (CMC) zircaloy-4 (Zr-4) hybrid fuel rodlet that may be tested in the INL Advanced Test Reactor (ATR) if the design is selected for further development and testing

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Radiation tolerant joining for silicon carbide-based accident tolerant fuel cladding

Silicon carbide (SiC) and silicon carbide fiber reinforced, silicon carbide matrix (SiC-SiC) composites are candidate materials for nuclear fuel cladding in accident tolerant light water reactors and advanced high temperature reactors due to their chemical stability and retention of strength at elevated temperature and neutron flux. A reliable means to join silicon carbide components is imperative to the successful implementation of SiC-based structural nuclear materials as fuel cladding as well as more complex core internal structures. A comprehensive approach has been taken to address several of the significant challenges associated with SiC joining and fabrication of complex SiC structures. This work focuses on low activation joining materials that are expected to behave well under a variety of different reactor operating conditions. Pre- and post-irradiation mechanical testing and microstructural evaluation will be presented for specimens subjected to fast flux irradiation to 4.5 dpa at 730 C. degrees, as well as 8.7 dpa at 270 C. degrees. Joints did not exhibit any strength degradation under either irradiation condition. Additionally, the performance of joined cylindrical SiC-SiC composite specimens was evaluated following irradiation in pressurized water reactor (PWR) coolant chemistry to 0.5 dpa. Lastly, out-of-pile characterization of permeability has been conducted on joined cylindrical SiC-SiC specimens to assess performance of this joint material under additional conditions relevant to reactor operation. Details of the experimental set-up and relationships between microstructure and performance will be discussed. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Safely probing the chemistry of Chernobyl nuclear fuel using micro-focus X-ray analysis

Detailed chemical analysis of the solidified molten fuel still residing in the stricken Chernobyl reactor unit 4 are inferred using multi-modal micro-focus X-ray analysis of a low-radioactivity proxy. A fascinating mixture of molten UO 2 , nuclear fuel cladding, concrete, stainless steel and other nuclear reactor components, these materials behaved like lava, solidifying to form a complex, highly radioactive glass-ceramic. Using element-specific chemical probes (micro-X-ray fluorescence and X-ray absorption spectroscopy), coupled with micro-diffraction analysis, the crystalline phase assemblage of simulants of these heterogeneous materials was established, which included “chernobylite” and a range of compositions in the (U 1-x Zr x )O 2 solid solution. Novel insight to nuclear accident fuel chemistry was obtained by establishing the oxidation state and local coordination of uranium not only in these crystalline phases, but uniquely in the amorphous fraction of the material, which varied depending on the history of the nuclear lava as it flowed through the reactor. This study demonstrates that micro-focus X-ray analysis of very small fractions of material can yield rich chemical information, which can be applied to nuclear-melt down materials to aid decommissioning and nuclear fuel management at nuclear accident sites.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

DOE EM Sponsored Research on Long-Term Dry Storage of Aluminum Clad SNF

This PowerPoint presentation describes and discusses DOE EM and INL lead studies on addressing potential technical issues associated with long-term dry storage of aluminum clad spent nuclear fuel. This same presentation was made to the US Nuclear Waste Technical Review Board in Nov 2019.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

In-Cell Deployment and First Use of Digital Image Correlation for In-Situ Strain Analysis of Irradiated Nuclear Fuel Rods During LOCA Transient

Digital image correlation (DIC) is a noncontact, optical method increasingly used across industries and research environments for acquiring multidimensional strain data. At Oak Ridge National Laboratory’s (ORNL’s) Severe Accident Test Station (SATS), DIC has been extensively applied to study the thermomechanical response of nuclear fuel claddings, yielding fundamental insights into material behavior. However, these efforts have focused exclusively on unirradiated materials and relied on the SATS out-of-cell infrastructure. Efforts over the past two years have been made to extend these capabilities to ORNL’s Irradiated Fuels Examination Laboratory SATS system within the hot-cells to enable testing of irradiated fuel cladding materials. Integrating DIC into the hot-cell SATS infrastructure presents unique challenges, including enabling remote operation of optical equipment and adapting auxiliary hot-cell systems for DIC implementation. This report discusses design, stand-up testing, and application of DIC to an irradiated nuclear fuel cladding segment. A DIC testing rig was successfully built and validated out-of-cell through extensive surrogate tests and calibrations. The rig and specialized DIC furnace were installed in the hot-cell, and a DIC test was successfully conducted. Results from that test correspond to expectations for Zr-based alloys and showed similar uncertainties compared to out-of-cell tests.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2011

Transition from Irradiation-Induced Amorphization to Crystallization in Nanocrystalline Silicon Carbide

Response to irradiation of nanocrystalline 3C-SiC is studied using 2 MeV Au+ ions near the critical temperature for amorphization and is compared to the behavior of its monocrystalline counterpart under the identical irradiation conditions. The irradiated samples have been characterized using in-situ ion channeling, ex-situ x-ray diffraction, and helium ion microscopy. Compared to monocrystalline 3C-SiC, a faster amorphization process in the nanocrystalline material (average grain size = 3.3 nm) is observed at 500 K. However, the nanograin grows with increasing ion fluence at 550 K and the grain size tends to saturate at high fluences. The striking contrast demonstrates a sharp transition from irradiation-induced interface-driven amorphization at 500 K to crystallization at 550 K. The results could show potential impacts of nanocrystalline SiC on nuclear fuel cladding and structural components of next-generation nuclear energy systems.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

The mechanical interaction of clad and base metal for molten salt reactor structural components

Cladding structural components with a corrosion resistant material may greatly extend the design life of molten salt reactor concepts. A complete design methodology for such cladded, high temperature nuclear components will require addressing many issues: fabrication, corrosion resistance, metallurgical interaction, and the mechanical interaction of the clad and base materials under load. This work focuses on the final issue: the mechanical interaction of the base and clad under creep-fatigue conditions. Depending on the relative mechanical properties of the two materials the clad may substantially influence the longterm cyclic response of the structural system or its effect might be negligible. To quantify the effect of different clad material properties we develop an efficient method for simulating pressurized cladded components in the limiting case where the section of interest is far from structural discontinuities. Using this method we evaluate the mechanics of the clad/base system and identify different regimes of mechanical response. The focus is on situations relevant to high temperature nuclear components: thermal-cyclic Bree-type problems and similar axisymmetric structures. The insights gained from these structural studies will form the basis for developing design rules for high-temperature, nuclear, cladded components.

Messner, M. C.↗
DOE OSTI2017

A pulse-controlled modified-burst test instrument for accident-tolerant fuel cladding

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 prehydrided ZIRLO cladding tubes. A ZIRLO cladding sample with a hydrogen content of 168 wt. ppm showed ductile behavior and failed at the maximum limits of the test setup with hoop strain to failure greater than 9.2%. ZIRLO samples showed high resistance to failure even at very high hydrogen contents (1,466 wt. ppm). When the hydrogen content was increased to 1,554 wt. ppm, brittle-like behavior was observed at a hoop strain of 2.5%. Preliminary scoping tests at room temperature with FeCrAl tubes were conducted to imitate 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 integral tests for current and potentially accident tolerant cladding materials.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗