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At least 433 records · Page 24
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Behavior of fuel with zirconium alloy cladding in reactivity-initiated accident and loss-of-coolant accident - 2016-0090

Along with our colleagues at the Japan Atomic Energy Research Institute (JAERI) and Japan Atomic Energy Agency (JAEA), we have performed extensive research programs for more than 2 decades and have developed a better understanding of fuel behavior under accident conditions. Data and findings from these research programs have provided the technical basis directly and indirectly for regulatory criteria in Japan and other countries. This paper reviews and summarizes the major outcomes from the research programs performed at JAERI and JAEA facilities and identifies further research needs. The programs aim to generate data and models to (1) evaluate the adequacy of present safety criteria and safety margins; (2) create a database for the future regulation of higher-burnup UO{sub 2} and mixed-oxide (MOX) fuels, new cladding, and pellets; (3) provide adequately mechanistic computer codes for regulatory application; (4) promote a better understanding of phenomena that appear in high-burnup regions and MOX fuels, such as rim effect and an effect of plutonium agglomerates in MOX fuels, and to evaluate those effects on fuel behavior in accident conditions; and (5) to assess fuel behaviors with higher fuel duty, such as plant power uprates, longer operating cycles, and water chemistry changes. The programs comprise reactivity-initiated accident studies, including pulse-irradiation experiments in the Nuclear Safety Research Reactor, cladding mechanical tests, and the development and verification of RANNS; loss-of-coolant accident tests, including integral thermal shock tests, oxidation rate measurements, and cladding mechanical tests; development and verification of a computer code FEMAXI-6 that simulates high-burnup fuel behavior under normal operating and abnormal transient conditions; and studies on phenomena specific to high-burnup regions, including thermal property measurements, experimental and analytical simulation of rim structure formation, cladding mechanical tests on effects of hydrides, etc. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Small-angle neutron scattering measurements of δ-phase deuteride (hydride) precipitates in Zircaloy 4

Small-angle neutron scattering (SANS) measurements have been performed under ambient conditions to characterize deuteride (hydride) particles in Zircaloy 4, a fuel cladding material used in pressurized light-water nuclear reactors. Hydrogen pickup by the cladding leads to a rim structure in which large circumferential hydride plate-like particles preferentially form on the cooler water-side region of the cladding. Deuterium substitution has been used to increase the coherent response and decrease the incoherent background of the SANS measurements. Four bulk deuterium concentrations were investigated, approximately 100, 400, 500 and 1000 parts per million by weight (w.p.p.m.) deuterium, as well as a zero-deuterium-concentration reference sample. The net SANS response from the deuteride phase was determined at all concentration values after subtraction of the reference SANS response, which effectively subtracted the strong scattering from second-phase particles in as-received Zircaloy. The net SANS response consisted of strong Porod scattering from deuteride particles over the entire measured Q range (0.005–0.4 Å -1 ). The net SANS response was anisotropic at concentrations greater than 100 w.p.p.m. and required elliptical averaging analysis. A significant sample orientation effect on the intensity of the SANS response was observed, due to preferential alignment of deuteride particles. The effect of ex situ applied stress at elevated temperature on deuteride phase dissolution and reprecipitation was investigated; a weak effect was observed with SANS that could not be confirmed by optical microscopy.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Integrating commercially available remote systems to address current D and D challenges - 15399

Many current D and D challenges across the DOE complex require some form of remote handling expertise and equipment. For example, high visibility projects such as Building 324 D and D and the recovery of wastes from the Alpha-Caissons and the 618 10-11 caissons at Hanford all require the development and application of advanced remote systems. Additionally although alternate solutions and options are being sought, robotic/remote systems will likely be required for completion of D and D at Savannah River Site's Building 235-F, at the West Valley Demonstration Project and many other DOE facilities that are scheduled for D and D. Similar systems are also needed to support the UK cleanup program on projects such as the decommissioning of the Pile Fuel Cladding Silo. As a result, there are a significant number of R and D and Technology Development projects underway in order to address the growing need for remote systems in D and D. In addition to these significant R and D efforts, there are also numerous remote systems in use in multiple applications in both nuclear and non-nuclear industries and the use of remote systems at the Fukushima Daiichi plant has been extensive in the past 12-24 months. In many cases, the systems developed are customized for their given application but there are also multiple systems in use which could be applied to the worldwide D and D challenge either directly or with some minor adaptation. During FY14, the Office of D and D and Facility Engineering (EM13) funded a review/assessment of the current state-of-the-art of robotics and remote technologies in order to generate a baseline of what is available, what has been used where and for what applications and what has been successful and what hasn't. Where possible, the assessment was extended to identify remote systems that have been used in non-nuclear industries but which, with limited modification, may have application in the nuclear industry. The work also included a review of the most recent publicly available TBuRD (Technical Baseline and Underpinning R and D) activities being undertaken by the UK NDA and their Site Licensed Companies (SLCs) in the area of remote systems in order to compile a searchable database of both available, commercial-off-the-shelf (COTS) remote systems and developing technologies. Other suitably qualified organizations in the US and the UK (such as UK National Nuclear Laboratory, SRNL, PNNL ORNL, OC Robotics and others) were also engaged to identify other ongoing technology development efforts which may be of interest and benefit to the decommissioning community. In all, of the order of 500 remote systems were identified and cataloged in both hard copy and as a searchable database. These systems were categorized in terms of their main application e.g. manipulator, wheeled land-based platform, track land-based platform, water-based platform, air-based platform, end effectors, sensor, crawler etc to clearly demonstrate the depth and breadth of available systems in each category. In addition, single page summaries of each system were produced and used, in a workshop setting, to identify technologies which, when integrated with other currently-available technologies, could provide a workable system to address existing needs. An initial workshop was held with CH2 Washington Idaho (CWI) and DOE Idaho in August 2014 to explore how the available systems could be integrated to address current and future D and D needs at that site. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Survivability of TREAT Zircaloy 3 Cladding

This chapter analyzes the effects of oxidation, growth, thermal stress, and fatigue on the Zircaloy-3 alloy used in TREAT fuel assembly cladding. The cladding surrounding the fuel is a protective shell which is used to prevent damage to the enclosed fuel. Therefore, its integrity is importance to guarantee this protection. All of the major factors which can affect the Zircaloy-3 cladding are considered in this chapter and investigated. Limits to operation are determined. The oxidation of Zircaloy-3 in air is of interest to air cooled reactors. The temperature range of interest is from room temperature where the fuel is stored for long periods of time, through the temperature range encountered in normal operation, to the highest temperatures which are possible in extreme accident situations. This temperature range is considered in this chapter to be from room temperature to 1100 C. Growth of the cladding can occur due to the buildup of the oxide layer which is less dense than the original metal and causes the metal to stretch. Thermal stress of the cladding can cause rupture of the cladding and therefore loss of protection to the fuel. Fatigue of the metal cladding occurs due to fuel handling.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Pool Boiling CHF Experiment for Cr-alloy-coated Accident Tolerant Fuel Cladding

The severe accident at the Fukushima nuclear power plants forced the nuclear communities to revise and reinforce their safety technology, especially accident tolerance. The main problem of Zr-based alloys is hydrogen generation owing to steam oxidation in a very high temperature environment. Currently, several types of Accident Tolerant Fuel (ATF) claddings that eliminate the steam oxidation problem have been presented as alternatives. Although the ATFs have shown appropriate performances in terms of material engineering, thorough tests including thermal hydraulic concerns are still required to ensure the design compatibilities. In this study, we conducted pool boiling critical heat flux (CHF) experiments using Cr-alloy-coated ATF claddings to verify basic thermal hydraulic performances. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2013

GeoMelt{sup R} ICV{sup TM} Treatment of Sellafield Pond Solids Waste - 13414

Kurion, Inc., in partnership with AMEC Ltd., is demonstrating its GeoMelt{sup R} In-Container Vitrification (ICV){sup TM} Technology to Sellafield Ltd. (SL). SL is evaluating the proposition of directly converting a container (skip/box/drum) of raw solid ILW into an immobilized waste form using thermal treatment, such that the resulting product is suitable for interim storage at Sellafield and subsequent disposal at a future Geological Disposal Facility. Potential SL feed streams include sludges, ion-exchange media, sand, plutonium contaminated material, concrete, uranium, fuel cladding, soils, metals, and decommissioning wastes. The solid wastes have significant proportions of metallic constituents in the form of containers, plant equipment, structural material and swarf arising from the nuclear operations at Sellafield. GeoMelt's proprietary ICV process was selected for demonstration, with the focus being high and reactive metal wastes arising from solid ILW material. A composite surrogate recipe was used to demonstrate the technology towards treating waste forms of diverse types and shapes, as well as those considered difficult to process; all the while requiring few (if any) pre-treatment activities. Key strategic objectives, along with their success criterion, were established by SL for this testing, namely: 1. Passivate and stabilize the raw waste simulant, as demonstrated by the entire quantity of material being vitrified, 2. Immobilize the radiological and chemo-toxic species, as demonstrated via indicative mass balance using elemental analyses from an array of samples, 3. Production of an inert and durable product as evidenced by transformation of reactive metals to their inert oxide forms and satisfactory leachability results using PCT testing. Two tests were performed using the GeoMelt Demonstration Unit located at AMEC's Birchwood Park Facilities in the UK. Post-melt examination of the first test indicated some of the waste simulant had not fully processed, due to insufficient processing time and melt temperature. A second test, incorporating operational experience from the first test, was performed and resulted in all of the 138 kg of feed material being treated. The waste simulant portion, at 41 kg, constituted 30 wt% of the total feed mass, with over 90% of this being made up of various reactive and non-reactive metals. The 95 liters of staged material was volume reduced to 41 liters, providing a 57% overall feed to product volume reduction in a fully passivated two-phase glass/metal product. The GeoMelt equipment operated as designed, vitrifying the entire batch of waste simulant. Post-melt analytical testing verified that 91-99+% of the radiological tracer metals were uniformly distributed within the glass/cast refractory/metal product, and the remaining fraction was captured in the offgas filtration systems. PCT testing of the glass and inner refractory liner showed leachability results that outperform the DOE regulatory limit of 2 g/m{sup 2} for the radiological species of interest (Sr, Ru, Cs, Eu, Re), and by more than an order of magnitude better for standard reference analytes (B, Na, Si). (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Corrosion resistant coatings for zirconium-alloy cladding with improved accident tolerance

The Westinghouse-led accident tolerant fuel (ATF) program is evaluating Accident Tolerant Fuel (ATF) coatings of MAX phase (Ti{sub 2}AlC) and multi-layer Ti{sub x}Al{sub 1-x}N/TiN to enhance the oxidation resistance of zirconium alloy fuel cladding in both normal operation and beyond design-based conditions of light water reactors. A variety of coating technologies have been applied to deposit these coating materials onto Zircaloy-4 or Zirlo sheets and tubes, including cold spray and cathodic arc-physical vapor deposition (CA-PVD). Neutronic analysis using the elemental compositions of the coating materials showed that the coating thicknesses should be less than 30 μm for most coating materials to avoid non-negligible economic penalty. The coated samples were autoclave tested in water at 360 Celsius degrees and 150 bar or in steam at 427 Celsius and 103 bar for corrosion resistance evaluation. The initial Ti{sub 2}AlC coating made by using large particle size MAX phase powder and cold spray deposition did not provide the desired corrosion resistance due to un-optimized microstructure and porosity. Subsequent optimization including reduction in particle size and tuning of cold spray deposition parameters resulted in denser coatings less than 30 microns in thickness. Optimization of process conditions resulted in reduced oxidation kinetics and showed that the coating acts as an effective barrier to prevent the zirconium substrate from oxidizing in accident condition temperatures. High purity Ti{sub 2}AlC powder is needed to further improve oxidation resistance of the Ti{sub 2}AlC coating. The oxidation resistance of pure monolithic MAX phase compound was also found to be remarkable, and the weight gain was only 1.85 mg/cm{sup 2} for 96 hours exposure in 1200 C. degrees steam, which is consistent with the literature. The corrosion resistance of the TiAlN/TiN multilayer coating was excellent. The weight gain was negligible and no oxidation was observed on the surface of zirconium-alloy at the coating-substrate interface. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Summary Report on Ion Irradiation Study of Ceramic Coating on Suppressing FCCI

Advanced cladding is critical for advanced nuclear reactors with an enhanced performance in radiation tolerance and neutron transparency. Using advanced cladding will ensure the adequate thermal conductivity and mechanical stability of the cladding base material, corrosion resistance, high-temperature coolant compatibility of the cladding surface, and chemical stability in the cladding inner wall against fuel cladding chemical interaction (FCCI). An innovative cladding with a three-layer structure (i.e., a modified surface, a clad base material, and a modified inner wall) promises to meet all these requirements. Initial research and development (R&D) regarding this innovative cladding seek to demonstrate the effectiveness of a thin ceramic coating in suppressing FCCI under ion irradiation to high dose. This United States (U.S.) Department of Energy (DOE)–Office of Nuclear Energy (NE) project report summarizes the results of recent ion irradiation studies of diffusion-couple samples with details of an ion irradiation experiment, the characterization of the interface microstructure of cerium (Ce)/titanium nitride (TiN)/oxide-dispersion strengthened (ODS), Ce/TiN/iron (Fe), Ce/ODS, and Ce/Fe samples irradiated with 80 million electron-volts (MeV) of xenon (Xe) ions to 100 displacements per atom (dpa) at 500?C using the Argonne Tandem Linac Accelerator System (ATLAS) facility at Argonne National Laboratory (ANL). The results of the ODS-substrate sample are encouraging and demonstrate the effectiveness of a thin ceramic coating on FCCI mitigation under irradiation. The results from an Fe-substrate sample revealed a complex microstructure, and the root cause is discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2014

Dose Rate Analysis Capability for Actual Spent Fuel Transportation Cask Contents

The approved contents for a U.S. Nuclear Regulatory Commission (NRC) licensed spent nuclear fuel casks are typically based on bounding used nuclear fuel (UNF) characteristics. However, the contents of the UNF canisters currently in storage at independent spent fuel storage installations are considerably heterogeneous in terms of fuel assembly burnup, initial enrichment, decay time, cladding integrity, etc. Used Nuclear Fuel Storage, Transportation & Disposal Analysis Resource and Data System (UNF ST&DARDS) is an integrated data and analysis system that facilitates automated cask-specific safety analyses based on actual characteristics of the as-loaded UNF. The UNF-ST&DARDS analysis capabilities have been recently expanded to include dose rate analysis of as-loaded transportation packages. Realistic dose rate values based on actual canister contents may be used in place of bounding dose rate values to support development of repackaging operations procedures, evaluation of radiation-related transportation risks, and communication with stakeholders. This paper describes the UNF-ST&DARDS dose rate analysis methodology based on actual UNF canister contents and presents sample dose rate calculation results.

Radulescu, Georgeta↗
DOE OSTI2016

Neutronics analysis of fuel-in-fiber concept

One of the consequences of Fukushima was the increase in interest in development of safer, more accident tolerant fuels and claddings. Accident tolerant fuels (ATF) are defined by the Nuclear Regulatory Committee (NRC) as fuels that, in comparison with the standard UO{sub 2}- Zircaloy system, can tolerate loss of active cooling in the core for a considerably longer time period while maintaining or improving the fuel performance during normal operations. One of the ideas being considered is tri-isotropic (TRISO) fuel particles. TRISO fuel pebbles consist of several layers - the outermost being graphite. TRISO pebbles are considered to be accident tolerant, because of their ability to retain fission products at high temperatures. One of the ideas implementing TRISO pebbles is Fully Ceramic Microencapsulated fuel where the pellet is comprised of TRISO pebbles in a matrix. TRISO particles though have a low maximum packing fraction when put in the spherical or cylindrical configuration with the current fabrication techniques (hot and cold press). This leads to considerable increase in enrichment to maintain cycle length for today's Light Water Reactors (LWR). The Fuel-in-Fiber concept, developed by FreeForm-Fiber (FFF), has the advantages of TRISO particles while utilizing a novel additive direct manufacturing technique for rapid fabrication. Fuel-in-Fiber are small fibers made of layers just like TRISO particles but because of their cylindrical nature, they can be packed with a tighter packing fraction than the TRISO particles. The fibers are woven within a SiC matrix and used to replace a typical pellet in a fuel pin. This means that they can easily be placed into LWR's without having to change the assembly or core structures. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Post Irradiation Examination of SiC Tube Subjected to Simultaneous Irradiation and Radial High Heat Flux

Silicon carbide (SiC) based composite is considered among leading candidate materials for accident tolerant fuel (ATF) cladding in light water reactors (LWRs) because of its outstanding irradiation and oxidation resistance, as well as low neutron absorption. Under normal operation conditions, the synergism between neutron irradiation and a high radial heat flux though the cladding thickness is expected to produce significant stress. This stress is attributed to a significant through-thickness temperature gradient and the resulting differential swelling strain due the highly temperature-dependent point-defect swelling of SiC. Although modeling of the in-pile stress state of SiC cladding has been conducted using finite element analysis, the analysis is challenging because of the complex stress state that depends upon swelling, irradiation creep, and thermal conductivity, all of which are both temperature and dose-dependent. These integral models have not yet been verified by experiments. In order to experimentally validate the multi-physics thermo-mechanical models of ATF SiC cladding during operation, SiC tube specimens have been irradiated under a high radial heat flux. This paper reports preliminary results from the post-irradiation examination (PIE) in on-going project supported by the U.S. Department of Energy's Nuclear Science User Facility program. The objectives of the PIE are (1) determination of the irradiation temperature distribution within the tube specimens, and (2) evaluation of the state and magnitude of the stress within the tube specimens arising from the temperature gradient. This paper mainly presents findings for objective (1)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Overview of lower length scale model development for accident tolerant fuels regarding U 3 Si 2 fuel and FeCrAl cladding

U 3 Si 2 and FeCrAl have been proposed as fuel and cladding concepts, respectively, for accident tolerance fuels with higher tolerance to accident scenarios compared to UO 2 . However, a lot of key physics and material properties regarding their in-pile performance are yet to be explored. To accelerate the understanding and reduce the cost of experimental studies, multiscale modeling and simulation are used to develop physics-based materials models to assist engineering scale fuel performance modeling. In this report, the lower-length-scale efforts in method and material model development supported by the Accident Tolerance Fuel (ATF) high-impact-problem (HIP) under the NEAMS program are summarized. Significant progresses have been made regarding interatomic potential, phase field models for phase decomposition and gas bubble formation, and thermal conductivity for U 3 Si 2 fuel, and precipitation in FeCrAl cladding. The accomplishments are very useful by providing atomistic and mesoscale tools, improving the current understanding, and delivering engineering scale models for these two ATF concepts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

A Multi-Stage Wear Model for Grid-to-Rod Fretting of Nuclear Fuel Rods

The wear of fuel rod cladding against the supporting structures in the cores of pressurized water nuclear reactors (PWRs) is an important and potentially costly tribological issue. Grid-to-rod fretting (GTRF), as it is known, involves not only time-varying contact conditions, but also elevated temperatures, flowing hot water, aqueous tribo-corrosion, and the embrittling effects of neutron fluences. The multi-stage, closed-form analytical model described in this paper relies on published out-of-reactor wear and corrosion data and a set of simplifying assumptions to portray the conversion of frictional work into wear depth. The cladding material of interest is a zirconium-based alloy called Zircaloy-4, and the grid support is made of a harder and more wear-resistant material. Focus is on the wear of the cladding. The model involves an incubation stage, a surface oxide wear stage, and a base alloy wear stage. The wear coefficient, which is a measure of the efficiency of conversion of frictional work into wear damage, can change to reflect the evolving metallurgical condition of the alloy. Wear coefficients for Zircaloy-4 and for a polyphase zirconia layer were back-calculated for a range of times required to wear to a critical depth. Inputs for the model, like the friction coefficient, are taken from the tribology literature in lieu of in-reactor tribological data. Concepts of classical fretting were used as a basis, but are modified to enable the model to accommodate the complexities of the PWR environment. Factors like grid spring relaxation, pre-oxidation of the cladding, multiple oxide phases, gap formation, impact, and hydrogen embrittlement are part of the problem definition but uncertainties in their relative roles limits the ability to validate the model. Sample calculations of wear depth versus time in the cladding illustrate how GTRF wear might occur in a discontinuous fashion during months-long reactor operating cycles. A means to account for grid/rod gaps and repetitive impact effects on GTRF wear is proposed

fretting↗
DOE OSTI2013

Sensitivity and uncertainty analyses for thermo-hydraulic calculation of research reactor

The sensitivity and uncertainty analysis of input parameters on thermohydraulic calculations for a research reactor has successfully done in this research. The uncertainty analysis was carried out on input parameters for thermohydraulic calculation of sub-channel analysis using Code COOLOD-N. The input parameters include radial peaking factor, the increase bulk coolant temperature, heat flux factor and the increase temperature cladding and fuel meat at research reactor utilizing plate fuel element. The input uncertainty of 1% - 4% were used in nominal power calculation. The bubble detachment parameters were computed for S ratio (the safety margin against the onset of flow instability ratio) which were used to determine safety level in line with the design of 'Reactor Serba Guna-G. A. Siwabessy' (RSG-GA Siwabessy). It was concluded from the calculation results that using the uncertainty input more than 3% was beyond the safety margin of reactor operation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Load following capability of CANDLE reactor by adjusting coolant operation condition

The load following capability of CANDLE reactor is investigated in the condition that the control rods are unavailable. Both sodium cooled metallic fuel fast reactor (SFR) and {sup 208}Pb cooled metallic fuel fast reactor (LFR) are investigated for their performance in power rate changing by changing its coolant operation condition; either coolant flow rate or coolant inlet temperature. The change by coolant flow rate is difficult especially for SFR because the maximum temperature criteria on cladding material may be violated. The power rate can be changed for its full range easily by changing the coolant temperature at the core inlet. LFR can reduce the same amount of power rate by smaller change of temperature than SFR. However, the coolant output temperature is generally decreased for this method and the thermal efficiency becomes worse.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Metal Fuel Implications on Severe Accident Scenarios

Superior steady-state irradiation performance of metal fuel has been demonstrated, and metal fuel has been the choice for recent sodium-cooled fast reactor designs, e.g. PRISM, SAFR, 4S, TWR, PGSFR, and so on. Metal fuel also has excellent transient capabilities. The metal fuel itself does not impose any restrictions on transient operations or load-following capabilities. Metal fuel has benign run beyond cladding breach (RBCB) performance characteristics. Because metal fuel is compatible with sodium, there is no reaction product and the fuel loss is practically zero. Metal fuel is expected to be very reliable. However, even if unforeseen fuel pin failure occurs, it will simply release gases and does not cause failure propagation to neighboring pins. The eutectic formation temperature between the fuel and the cladding has been considered a critical parameter for the metal fuel pin design. The onset of fuel-cladding eutectic formation starts in the 700-725 deg. C temperature range, depending on the fuel alloy and cladding types. However, at this onset temperature, not much interaction occurs. In fact, even at 100 deg. C above the eutectic temperature, the eutectic penetration into the cladding is minimal in one hour. Only at much higher temperatures, close to the fuel melting point itself, the eutectic penetration into cladding becomes rapid. Therefore, the eutectic formation is not a primary safety concern during transient overpower conditions. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

ORNL Interim Progress Report on Hydride Reorientation CIRFT Tests

A systematic study of H. B. Robinson (HBR) high burnup spent nuclear fuel (SNF) vibration integrity was performed in Phase I project under simulated transportation environments, using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) hot cell testing technology developed at Oak Ridge National Laboratory in 2013–14. The data analysis on the as-irradiated HBR SNF rods demonstrated that the load amplitude is the dominant factor that controls the fatigue life of bending rods. However, previous studies have shown that the hydrogen content and hydride morphology has an important effect on zirconium alloy mechanical properties. To address the effect of radial hydrides in SNF rods, in Phase II a test procedure was developed to simulate the effects of elevated temperatures, pressures, and stresses during transfer-drying operations. Pressurized and sealed fuel segments were heated to the target temperature for a preset hold time and slow-cooled at a controlled rate. The procedure was applied to both non-irradiated/prehydrided and high-burnup Zircaloy-4 fueled cladding segments using the Nuclear Regulatory Commission-recommended 400°C maximum temperature limit at various cooling rates. Before testing high-burnup cladding, four out-of-cell tests were conducted to optimize the hydride reorientation (R) test condition with pre-hydride Zircaloy-4 cladding, which has the same geometry as the high burnup fuel samples. Test HR-HBR#1 was conducted at the maximum hoop stress of 145 MPa, at a 400°C maximum temperature and a 5°C/h cooling rate. On the other hand, thermal cycling was performed for tests HR-HBR#2, HR-HBR#3, and HR-HBR#4 to generate more radial hydrides. It is clear that thermal cycling increases the ratio of the radial hydride to circumferential hydrides. The internal pressure also has a significant effect on the radial hydride morphology. This report describes a procedure and experimental results of the four out-of-cell hydride reorientation tests of hydrided Zircaloy-4 cladding, which served as a guideline to prepare in-cell hydride reorientation samples with high burnup HBR fuel segments. This report also provides the Phase II CIRFT test data for the hydride reorientation irradiated samples. The variations in fatigue life are provided in terms of moment, equivalent stress, curvature, and equivalent strain for the tested SNFs. The CIRFT results appear to indicate that hydride reoriented treatment (HRT) have a negative effect on fatigue life, in addition to hydride reorientation effect. For HR4 specimen that had no pressurization procedure applied, the thermal annealing treatment alone showed a negative impact on the fatigue life compared to the HBR rod.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗