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At least 163 records · Page 9
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Transient Mechanical Test Development for Accident Tolerant Cladding Candidates to Simulate PCMI-like Conditions

This study investigates the transient mechanical response of accident tolerant fuel (ATF) cladding candidates to provide information on potential PCMI-induced deformation and failure of candidate cladding materials. The mechanical testing is applied t to Gen-I FeCrAl and triaxially-braided SiC fiber SiC matrix composite materials under loading conditions like the PCMI during a postulated design basis reactivity insertion accident. To achieve the specific loading conditions, a modified burst test instrument with accurate mechanical pulse control has been designed and constructed. The mechanical response of FeCrAl, SiC-SiC composite tubes were compared with the pre-hydrided zirconium alloy cladding. The specimens subjected to mechanical pulses were investigated in a strain-controlled manner deformation.

Cinbiz, Nedim↗
DOE OSTI2023

AFC FY 2023 HFIR Irradiation Test Matrix – Supported by the Design of a Ring Specimen Irradiation Vehicle

The development and deployment of accident-tolerant cladding materials in light water nuclear reactors requires the collection of neutron-irradiated properties of these materials with specific consideration to the material properties orientation (i.e., tube geometry). To that end, a test plan to irradiate several cladding tube and ring specimens in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) was proposed, along with the development of a new capsule design that features ring specimens with a preexisting gauge section. This ring specimen geometry removes the need to machine a gauge section on the irradiated specimen and thus allows for direct application of a ring-pull test post-irradiation. The test matrix includes uncoated and chromium-coated Zircaloy-4 and advanced zirconium alloy cladding specimens, pre-characterized at ORNL. Plans are to irradiate the specimens at about 4 and 10 dpa—corresponding to two and five cycles at the HFIR centerline—with a target specimen gauge section temperature of 330°C.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Selection of Nuclear Fuel for TREAT: UO 2 vs U 3 O 8

The Transient Reactor Test (TREAT) that resides at the Materials and Fuels Complex (MFC) at Idaho National Laboratory (INL), first achieved criticality in 1959, and successfully performed many transient tests on nuclear fuel until 1994 when its operations were suspended. Resumption of operations at TREAT was approved in February 2014 to meet the U.S. Department of Energy (DOE) Office of Nuclear Energy’s objectives in transient testing of nuclear fuels. The National Nuclear Security Administration’s is converting TREAT from its existing highly enriched uranium (HEU) core to a new core containing low enriched uranium (LEU) (i.e., U-235< 20% by weight). The TREAT Conversion project is currently progressing with conceptual design phase activities. Dimensional stability of the fuel element assemblies, predictable fuel can oxidation and sufficient heat conductivity by the fuel blocks are some of the critical performance requirements of the new LEU fuel. Furthermore, to enable the design team to design fuel block and can specifications, it is amongst the objectives to evaluate TREAT LEU fuel and cladding material’s chemical interaction. This information is important to understand the viability of Zr-based alloys and fuel characteristics for the fabrication of the TREAT LEU fuel and cladding. Also, it is very important to make the right decision on what type of nuclear fuel will be used at TREAT. In particular, one has to consider different oxides of uranium, and most importantly, UO 2 vs U 3 O 8 . In this report, the results are documented pertaining to the choice mentioned above (UO 2 vs U 3 O 8 ). The conclusion in favor of using UO 2 was made based on the analysis of historical data, up-to-date literature, and self-consistent calculations of phase equilibria and thermodynamic properties in the U-O and U-O-C systems. The report is organized as follows. First, the criteria that were used to make the choice are analyzed. Secondly, existing historical data and current literature were reviewed. This analysis was supplemented by the construction and examination of the U-O and U-O-C phase diagrams at pressure close to negligent, thereby mimicking the conditions in which nuclear fuel is supposed to function inside the zirconium-based cladding in the reactor. Finally, our conclusion in favor of the UO 2 down selection was summarized and explained in the last Section of this document.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Study of hydride reorientation and ductility of SRA ZIRLO and PRXA ZIRLO using ring compression testing (RCT)

Before dry storage, spent nuclear fuel is dried at high temperatures, which induces hydride dissolution. Radial hydride particles precipitate when cooled under a stress above a threshold stress. This process can severely degrade the ductility of cladding tubes. Zirconium alloys with different microstructures, such as grain boundary density, were quantified and studied to examine their effect on hydride reorientation. The ring compression loading mode creates a range of stress states depending on the location within the ring with respect to the applied load, producing various hydride morphologies that can be used to study the effects of microstructures on hydride reorientation. Finite element modeling was performed to calculate the hoop stress, which is the dominant stress, during hydride reorientation. These results were coupled with hydride morphology to determine the threshold stress for hydride reorientation, depending on the microstructure. In conclusion, the samples were then mechanically tested using ring compression to evaluate the ductility of zirconium alloys as a function of different hydride morphologies.

Hydride embrittlement↗
DOE OSTI2022

A review of cladding failure thresholds in RIA conditions based on transient reactor test data and the need for continued testing

Transient reactor experiments on light water reactor (LWR) fuel pins had been conducted since the beginning of the nuclear era to help determine core coolability and cladding failure thresholds. During one such test in November of 1993 at the CABRI transient test reactor on a test involving a high burnup fuel rod with a corroded Zircaloy-4 cladding it was first observed that cladding failures could occur prior to a departure from nucleate boiling (pre-DNB) at lower-than-expected peak radial average enthalpies. This paper will present an independent review of the publicly available transient reactor test database on higher burnup LWR pins conducted at the CABRI and NSRR reactors as well as review of a selection of published out of pile mechanical testing methods. The purpose of the review is to determine how well the new regulatory limits are supported by experimental data. The review will identify if additional transient reactor tests could provide additional support for the NRC guidance or identify the need for revisions. The evaluation will consider how far the existing database can be extrapolated when considering low hydrogen zirconium alloy claddings (with and without protective coatings) containing very high burnup (> 70 MWd/kgU) UO2 fuel pellets. Finally, the authors will suggest how out of pile mechanical tests can be used in conjunction with a limited number of transient reactor tests to develop cladding specific failure thresholds in RIA type transients.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Development of an Electrochemical Impedance Instrument for Monitoring Corrosion Formation (Oxide and Hydride) on Cladding Materials

The objective of this work package is to develop electrochemical impedance spectroscopic (EIS) techniques for the in pile measurement of changes in cladding materials in coupling with model simulation and material characterization. This work package will involve the initial development of electrochemical sensing technologies for measuring spatial and time resolved changes in cladding chemistry. Specific attention will be paid to monitoring changes in monitor cladding corrosion including oxide and hydride formation and deformation. The planned work during this period was focused on cladding material zirconium alloy corrosion, oxide and hydride development, and to characterize samples using EIS under controlled gas atmosphere, atom probe tomography (APT), Raman Spectroscopy, and Synchrotron X-ray diffraction (XRD). The corrosion, including oxide and hydride of zirconium alloy samples, was tested in this study. The samples were treated at a gas environment containing 3% H2 (vol%) balanced with N2 gas at various temperatures (100 – 500°C) and different times (up to 60 hours) and were characterized by in situ EIS. After treatment, various characterization techniques were applied to measure the samples, including TGA, Ramen, XRD, and APT. This study found that impedance spectra can be affected significantly with (furnace/environment) temperature. Higher temperature resulted in higher overall impedance, and longer exposure time exerted similar effects. Proof of concept characterization of oxidized samples, including TGA, Raman, XRD, and APT, were accomplished on samples exposed to 700 oC gas environment containing 20% O2 and 80% N2. TGA provides observable breakaway and bulk oxide growth measurements. High-temperature, in-situ Raman provides observation of oxidation stress at the surface of the oxide. Additionally, a transition from tetragonal to monoclinic zirconia can be seen early in the oxidation experiment. Specifically, tetragonal-rich zirconia is seen at the metal/oxide interface, where a transition to monoclinic-rich zirconia is seen in the bulk of the oxide. It has also been confirmed that highly compressive residual stresses are seen near the metal/oxide interface. It is concluded that the transition from tetragonal to monoclinic zirconia is induced by massive shifts in stress as the oxide grows. APT has supplied evident differences in the distribution of zirconium and oxygen at the metal/oxide interface. On the pre-breakaway sample, the metastable Zr(O) region at the boundary is relatively thin, interfaced by Zr-rich (i.e., metal) and O-rich (i.e., ZrO2) on either side. In contrast, it has been seen on a post-breakaway sample that the Zr(O) metastable region is much greater in thickness, with evidence of four “zones” showing slightly contrasting composition between 49-54% Zr. In addition, it can be seen from the Zr-2.65Nb post-breakaway sample that secondary element concentrations are considerably different than bulk composition measurements show. Specifically, Nb concentration is much lower than expected, whereas Fe concentration is much higher than expected. These differences could play a role in each element’s contribution to oxide growth mechanisms, as well as causation for breakaway. Future work is focused on analyzing EIS spectra of hydride and its correlation with characterization features, finite element (FE) modeling of cladding materials, electrochemical sensor development, targeting cladding oxides/hydrides under thermodynamic equilibria, during transients and under irradiation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Transmission electron microscopy examinations of metal-oxide interface of zirconium-based alloys irradiated in Halden reactor-IFA-638 - 2016-0039

This paper provides the results of investigations by transmission electron microscopy (TEM) on the selected materials from in-reactor oxidation tests in the Halden test reactor (Reference No. IFA-638) from 1998 to 2006. The objective of the IFA-638 test was to study the corrosion behavior of modern zirconium-based claddings to high burnup in pressurized water reactor water chemistry and thermal hydraulic conditions. The aim of this paper is to report on the microstructure of selected materials (ZIRLO{sup R}, E635, and Alloy A) after the irradiation to different burnup levels to determine the modifications induced by irradiation and to correlate results to their oxidation behavior. The TEM examinations revealed the nature of secondary phase particles (SPPs) and their modification under irradiation. Four types of SPPs were observed, namely β-niobium precipitates, Zr{sub 0.5}Nb{sub 0.3}Fe{sub 0.2} (mainly in the ZIRLO alloy), Zr(Fe,Nb){sub 2} (in E635), and (Cr,Fe){sub 2}Zr,Nb with varying niobium content (present in Alloy A: Zr-0.58Sn-0.31Nb-0.36Fe-0.26Cr). TEM observations showed that all three materials contained still several precipitates after irradiation and in the case of the ZIRLO alloy even after high burnups. Furthermore, the analysis of the metal side of the interface and its comparison with the oxide side led to the conclusion that all types of precipitates dissolved to some extent under irradiation and that their alloying element content decreased. The dissolution was intensified in the oxide. However, a more detailed examination showed that the β-niobium precipitates dissolved at a slower rate, or knowing that their composition was much richer in niobium, the time needed for the precipitates to become fully depleted from niobium was longer. Regarding the amorphization under irradiation, the b-niobium- and chromium-containing precipitates did not amorphize in the metal part of the interface. This was not the case for the other types of precipitates. Furthermore, these two types of SPP both showed delayed oxidation and due to this behavior the typical crack above the SPP in the oxide was also observed. These results are discussed to gain an improved understanding of the oxidation behavior of materials studied as a function of irradiation and residence time. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2013

Characterization of Irradiated Metal Waste from the Pyrometallurgical Treatment of Used EBR-II Fuel

As part of the pyrometallurgical treatment of used Experimental Breeder Reactor-II fuel, a metal waste stream is generated consisting primarily of cladding hulls laden with fission products noble to the electrorefining process. Consolidation by melting at high temperature [1873 K (1600 degrees C)] has been developed to sequester the noble metal fission products (Zr, Mo, Tc, Ru, Rh, Te, and Pd) which remain in the iron-based cladding hulls. Zirconium from the uranium fuel alloy (U-10Zr) is also deposited on the hulls and forms Fe-Zr intermetallics which incorporate the noble metals as well as residual actinides during processing. Hence, Zr has been chosen as the primary indicator for consistency of the metal waste. Recently, the first production-scale metal waste ingot was generated and sampled to monitor Zr content for Fe-Zr intermetallic phase formation and validation of processing conditions. Chemical assay of the metal waste ingot revealed a homogeneous distribution of the noble metal fission products as well as the primary fuel constituents U and Zr. Microstructural characterization of the ingot confirmed the immobilization of the noble metals in the Fe-Zr intermetallic phase.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Innovative accident tolerant fuel concept enabled through direct manufacturing technology

Nuclear energy is one of the largest sources of carbon-free electricity in the world. Some countries are looking at new ways to support and revitalize the nuclear sector since the Fukushima disaster. The accident tolerant fuel program is geared toward improving the safety of nuclear energy by investigating materials that can replace or modify the current uranium-dioxide nuclear fuel and zirconium-based cladding. This research program is being supported by all major nuclear countries since 2011. The practical limitations on allowable uranium enrichment has taken the focus away from the most promising fuels with high radioactivity retention such as tristructural isotropic particle fuel. To overcome such enrichment limitation, a new fuel concept is proposed using advanced ceramic direct manufacturing with laser-induced chemical vapor deposition. The fuel-as-fiber concept is an accident-tolerant fuel design that features high thermal conductivity, strong capability of radioactivity retention and most importantly requires reasonable enrichment levels with uranium nitride as the fuel. In this work, the initial fabrication of uranium-based fuel with laser-induced chemical vapor deposition technology is demonstrated. Then an advanced multi-physics guided modeling approach based on finite element analysis codes and informed by the manufacturing capabilities is developed to accelerate the advancement of fuel-as-fiber concept for use in current light-water reactor technology. The detailed thermomechanical analysis showed promising results for viability of the innovative fuel-as-fiber concept. The predicted stresses in the fuel structural materials were similar to the case of tristructural isotropic particle fuel experience base that has shown excellent reliability in retention of fuel radioactivity at high temperatures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Technique development for modulus, microcracking, hermeticity, and coating evaluation capability characterization of SiC/SiC tubes

Driven by the need to enlarge the safety margins of nuclear fission reactors in accident scenarios, research and development of accident-tolerant fuel has become an important topic in the nuclear engineering and materials community. A continuous-fiber SiC/SiC composite is under consideration as a replacement for traditional zirconium alloy cladding owing to its high-temperature stability, chemical inertness, and exceptional irradiation resistance. An important task is the development of characterization techniques for SiC/SiC cladding, since traditional work using rectangular bars or disks cannot directly provide useful information on the properties of SiC/SiC composite tubes for fuel cladding applications. At Oak Ridge National Laboratory, experimental capabilities are under development to characterize the modulus, microcracking, and hermeticity of as-fabricated, as-irradiated SiC/SiC composite tubes. Resonant ultrasound spectroscopy has been validated as a promising technique to evaluate the elastic properties of SiC/SiC composite tubes and microcracking within the material. A similar technique, impulse excitation, is efficient in determining the basic mechanical properties of SiC bars prepared by chemical vapor deposition; it also has potential for application in studying the mechanical properties of SiC/SiC composite tubes. Complete evaluation of the quality of the developed coatings, a major mitigation strategy against gas permeation and hydrothermal corrosion, requires the deployment of various experimental techniques, such as scratch indentation, tensile pulling-off tests, and scanning electron microscopy. In addition, a comprehensive permeation test station is being established to assess the hermeticity of SiC/SiC composite tubes and to determine the H/D/He permeability of SiC/SiC composites. This report summarizes the current status of the development of these experimental capabilities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Determination of He and D permeability of neutron-irradiated SiC tubes to examine the potential for release due to micro-cracking

Driven by the need to enlarge the safety margins of light water reactors in both design-basis and beyond-design-basis accident scenarios, the research and development of accident-tolerant fuel (ATF) has become an importance topic in the nuclear engineering and materials community. Continuous SiC fiber-reinforced SiC matrix ceramic composites are under consideration as a replacement for traditional zirconium alloy cladding owing to their high-temperature stability, chemical inertness, and exceptional irradiation resistance. Among the key technical feasibility issues, potential failure of the fission product containment due to probabilistic penetrating cracking has been identified as one of the two most critical feasibility issues, together with the radiolysisassisted hydrothermal corrosion of SiC. The experimental capability to evaluate the hermeticity of SiC-based claddings is an urgent need. In this report, we present the development of a comprehensive permeation testing station established in the Low Activation Materials Development and Analysis laboratory at Oak Ridge National Laboratory. Preliminary results for the hermeticity evaluation of un-irradiated monolithic SiC tubes, uncoated and coated SiC/SiC composite tubes, and neutron-irradiated monolithic SiC tubes at room temperature are exhibited. The results indicate that this new permeation testing station is capable of evaluating the hermeticity of SiC-based tubes by determining the helium and deuterium permeation flux as a function of gas pressure at a high resolution of 8.07 x 10 -12 atm-cc/s for helium and 2.83 x 10 -12 atm-cc/s for deuterium, respectively. The detection limit of this system is sufficient to evaluate the maximum allowable helium leakage rate of lab-scale tubular samples, which is linearly extrapolated from the evaluation standard used for a commercial as-manufactured light water reactor fuel rod at room temperature. The un-irradiated monolithic SiC tube is hermetic, as is manifested by the un-detectable deuterium permeation flux at various feeding gas pressures. A large helium leakage rate was detected for the uncoated SiC/SiC composite tube exposed to atmosphere, indicating it is inherently not hermetic. The hermeticity of coated SiC/SiC composite tubes is strongly dependent on the coating materials and the preparation of the substrate SiC/SiC composite samples. To simulate the practical application environment, monolithic CVD SiC tubes were exposed to neutron irradiation at the High Flux Isotope Reactor under high heat flux from the internal surface to the external surface. Although finite element analysis and resonant ultrasound spectroscopy measurement indicated that the combined neutron irradiation and high heat flux gave rise to a high probability of cracking within the sample, the hermeticity evaluation of the tested sample still exhibited gas tightness, emphasizing that SiC cracking is inherently a statistical phenomenon. The developed permeation testing station is capable of measuring the gas permeation flux in the range of interest with full confidence based on the presented results. It is considered a critical pre- /post-irradiation examination technique to characterize SiC-based cladding materials in asreceived and irradiated states to aid the research and development of ATF.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Nuclear Energy Advanced Modeling and Simulation (NEAMS) Accident Tolerant Fuels High Impact Problem: FeCrAl Modeling Capabilities

Iron-chromium-aluminum alloys are a potential candidate replacement for zirconium-based claddings in light water reactor fuel rods due to their increased strength (lower creep rates and higher yield stress) and slower oxidation kinetics with steam. Their larger neutron absorption cross-section necessitates thinner claddings and larger fuel pellets to maintain the same as-fabricated fuel-to-clad gap. The aggressive development schedule of the United States Department of Energy's (DOE) Advanced Fuels Campaign (AFC) to have potential accident tolerant fuels (ATF) concepts in lead test rods or assemblies by 2022 has necessitated the need for advanced modeling and simulation techniques to investigate the behavior of these materials under normal operating and accident conditions. This report presents the modeling capabilities for FeCrAl claddings at the engineering (continuum) scale developed as part of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) high impact problem (HIP). Sensitivity studies of select material parameters have been completed to determine their significance in predictions of important fuel performance parameters such as fuel centerline and surface temperatures, cladding stresses and strains, fission gas released, and rod internal pressure. Comparative analyses of FeCrAl and Zircaloy-4 claddings for separate effects conditions are presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Radiation Tolerance of Controlled Fusion Welds in High Temperature Oxidation Resistant FeCrAl Alloys

High temperature oxidation resistant iron-chromium-aluminum (FeCrAl) alloys are candidate alloys for nuclear applications due to their exceptional performance during off-normal conditions such as a loss-of-coolant accident (LOCA) compared to currently deployed zirconium-based claddings [1]. A series of studies have been completed to determine the weldability of the FeCrAl alloy class and investigate the weldment performance in the as-received (non-irradiated) state [2,3]. These initial studies have shown the general effects of composition and microstructure on the weldability of FeCrAl alloys. Given this, limited details on the radiation tolerance of FeCrAl alloys and their weldments exist. Here, the highest priority candidate FeCrAl alloys and their weldments have been investigated after irradiation to enable a better understanding of FeCrAl alloy weldment performance within a high-intensity neutron field. The alloys examined include C35M (Fe-13%Cr-5% Al) and variants with aluminum (+2%) or titanium carbide (+1%) additions. Two different sub-sized tensile geometries, SS-J type and SS-2E (or SS-mini), were neutron irradiated in the High Flux Isotope Reactor to 1.8-1.9 displacements per atom (dpa) in the temperature range of 195°C to 559°C. Post irradiation examination of the candidate alloys was completed and included uniaxial tensile tests coupled with digital image correlation (DIC), scanning electron microscopy-electron back scattered diffraction analysis (SEM-EBSD), and SEM-based fractography. In addition to weldment testing, non-welded parent material was examined as a direct comparison between welded and non-welded specimen performance. Both welded and non-welded specimens showed a high degree of radiation-induced hardening near irradiation temperatures of 200°C, moderate radiation-induced hardening near temperatures of 360°C, and almost no radiation-induced hardening at elevated temperatures near 550°C. Additionally, low-temperature irradiations showed the non-welded specimens to exhibit strain-induced softening (decrease in the true stress level) with increasing plastic strain during tensile testing. Fracture for the weldments was found to occur exclusively within the fusion zone. The mechanical performance of the weldment was speculated to be directly linked to variances in the radiation-induced microstructure including the formation of dislocation loops and precipitation of the Cr-rich α' phase. The localized microstructural variation within the weldments, including grain size, was determined to play a significant role in the radiation-induced microstructure. The results summarized within highlight the need for additional data on the radiation tolerance of weldments as the mechanical performance of the fusion zone was shown to be the limiting factor in the overall performance of the weldments.

36 MATERIALS SCIENCE↗
DOE OSTI2019

TREAT Fuel Motion Summary Report. SETH A-E Experiments

In support of the Accident Tolerant Fuels program, experiment devices are being developed to enable transient testing of pressurized water reactor (PWR) type fuel specimens. In fiscal year 2019 several fueled tests were conducted in the Transient Reactor Test Facility (TREAT) using the Separate Effects Test Holder (SETH). Five different SETH capsules (SETH A-E) were assembled, each holding a 10-pellet rodlet with 4.9% enriched, commercially-produced fresh UO2 pellets in zirconium alloy cladding with pellet dimensions having typical PWR radial dimensions and a total pellet stack length measuring 10.16 cm. The rodlet’s length allows the specimen to be axially positioned in the center of the core with minimal flux variation over the height of the rodlet. The fuel motion monitoring system (FMMS) is located on the north side of the TREAT reactor. Having direct line-of-sight of the core enables the FMMS to collect spatially-resolved data for fast neutrons generated due to fission occurring in experiment specimens. The FMMS contains 360 channels that make up a 10 (horizontal) by 36 (vertical) array of viewing slots; for the SETH A-E test series 96 proton-recoil scintillator (PRS) detectors were installed in the system. The fast neutrons are detected in the PRS detectors from light produced by the recoil of protons interacting with ZnS(Ag) grains in each detector's scintillator button. This report focuses on the FMMS data captured in the SETH A-E experiments, with multiple transients varying in reactivity addition and length. The first three experiments had a total reactor energy of 101 MJ each and were critical in understanding the FMMS field of view, lead shielding, and other TREAT data acquisition systems. SETH-B2 and C had a reactivity addition of almost double the previous transients and served as a demonstration of the detector array behavior under increased reactor energy. The final two SETH experiments (D and E) were aimed at reaching cladding melting temperatures with a reactor energy over 500 MJ for each test. In preparation for the higher energy transients, the FMMS lead shielding was reduced to 0.5 inches from the 6 inches used in SETH A-C. This was expected to saturate the detectors at the peak of the transient in exchange for higher sensitivity at lower energies after the peak, when the motion was expected. SETH-D was clipped at approximately 3 seconds, which was too early for the FMMS to capture any rodlet motion. SETH-E had the same reactivity addition as SETH-D but the transient time was extended to 18 seconds. The FMMS captured the downward motion of the rodlet in SETH-E as the count rate significantly changed for the detectors viewing the specimen. Snapshots taken throughout the transient helped identify the approximate time at which the motion initiated. Further analysis of the individual detector data captured on a per-millisecond time basis resulted in a determination that the approximate time for the onset of the downward motion was 2.84 ± 0.02 seconds. Neutron radiographs confirmed that downward motion occurred for SETH-D and SETH-E. Motion in SETH-D could also have been captured by the FMMS if the transient time had been longer for SETH-D. Ultimately, these final two tests served to demonstrate the FMMS's ability to track fuel motion during transient experiments at TREAT.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Status Report on Development of TREAT Water Loop

The Transient Reactor Test (TREAT) facility is a graphite based transient test reactor at the Idaho National Laboratory (INL). TREAT started operation in 1959 and performed thousands of transient irradiations for the primary purpose of nuclear fuel safety research. TREAT operations were suspended in 1994, and then, several years later, resumed reactor operations in 2017 to support development of advanced nuclear fuels. TREAT’s more recent historic focus had been for transients with relevance to sodium cooled fast spectrum reactors. TREAT’s first year of modern experiment work performed in 2018, however, saw the demonstration of new kinds of transients focused on light water reactor (LWR) needs including reactivity initiated accidents (RIA) and loss of coolant accidents (LOCA). The successful outcome of these efforts confirmed TREAT’s viability as a transient neutron source for LWR safety research. Since TREAT’s core is air cooled and provides no “built-in” thermal hydraulic boundary conditions for tests specimens, efforts have been underway to establish water-environment capabilities for testing specimens. Small static water capsules have been established to support single rod RIA testing where transient heating is brief. Similarly, deployment of enhanced water capsules with blowdown capability is nearing completion to support single rod LOCA testing. With careful experiment design these capsules can provide representative temperature response histories and water phase transitions for some experiment parameters. The capsule approach is appropriate for cost-effective early phase phenomena identification studies on novel fuels, or when the important test variables are thought to be fairly well understood so that they can be precisely targeted (such as they are for current fleet fuels with moderate burnup UO2 in uncoated zirconium alloy cladding). Due to lack of forced convection, however, the capsule approach cannot achieve representative temperature histories in all regions of the test fuel simultaneously (e.g. cladding and fuel centerline temperatures) while providing representative timing in water phase transitions. Nor can the capsule approach provide appropriate hydraulic conditions for multi-rod bundle testing. For these reasons, design and development has been undertaken for the TREAT Water-Environment Recirculating Loop (TWERL).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Evaluation of Hydrogen Generation in High Burnup Demonstration Dry Storage Cask

This report provides a best-estimate evaluation of residual water content (post-dry out) in the High Burnup (HBU) LWR Spent Fuel Demonstration project TN-32 cask, and evaluates the radiolysis of the residual free water, and the physisorbed and chemisorbed waters on the surfaces of the fuel and cask internal contents. The evaluation of radiolytic breakdown of those waters with gamma radiation causing the generation of hydrogen gas (H 2 ) is made using available literature data and models. This evaluation is part of the overall materials performance evaluation of the SNF-in-canister system, and is part of the technical bases for their continued safe dry storage. The TN-32 cask contents included 32 HBU LWR spent fuel assemblies each with 264 fuel rods clad in zirconium alloys, aluminum neutron absorber components, and aluminum and stainless steel structural components. The residual free and surface (physisorbed/chemisorbed) waters are ascribed to water vapor in the free volume and to components’ surfaces, respectively. The total potential radiolytic hydrogen inventory from the water vapor and from waters ascribed to surfaces has been calculated assuming all the water produced molecular H 2 . The residual water that is chemically incorporated into the bulk of a hydrated oxide, i.e., chemisorbed water, and its total potential hydrogen inventory has been calculated. These calculations are at the physical limit of material available and are used for a bounding assessment purpose only.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗