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At least 883 records · Page 49
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Development and Integration of a Stochastic Clad Damage Propagation Model into PRONGHORN-SC Subchannel Analysis Code

The failure of fuel pins in nuclear reactors is intrinsically stochastic. Typically, a combination of variation in manufacturing that affects the material characteristics and the fuel assembly dimensions, variation in operating conditions, such as local power, coolant flow rate, and irradiation induced changes in material properties lead to a large uncertainty in failure margin of the fuel pins. Failure, therefore, may occur in exceptional pins with adverse combinations of these variations. Upon a metal fuel pin (U-Pu-Zr/HT9) failure, depressurization of the fuel pin takes place by release of fission gas, liquid sodium bond, and potentially solid fuel particles or molten/eutectic fuel droplets through the hole in cladding. The effect of a fission gas jet on neighbor fuel pins and possible propagation of a clad damage during normal operation was studied experimentally in 1970s and it was found that the post-failure fission gas jet insulates the jet impingement area of the target fuel pin surface and could increase the target pin’s surface temperature by as much as 100 – 200 K during the failed pin depressurization. It was concluded that the effect should not lead to fuel pin failure propagation during normal operation. In accident scenarios of sodium and lead fast reactors such as Unprotected Loss-Of-Flow (ULOF) or Unprotected Transient Over Power (UTOP), the fuel pins can be subjected to higher clad temperatures and fuel pin pressures or fuel clad mechanical/chemical interaction where thermal creep margin becomes significantly lower compared to the normal operation conditions. Therefore, possible stochastic failure and the post-failure fission gas/fuel jet impingement could be critical in order to predict fuel pin failure propagation. Pin depressurization due to fission gas release may degrade the heat transfer by formation of a gas blanket on a neighboring pin surface, which is a local phenomenon, and by causing coolant flow deceleration and starvation, which could affect a surrounding region as well. Furthermore, the potential presence of solid fuel particles or molten fuel at the time of clad failure could boost post-failure jet induced degradation even further. The present study models the U-Pu-Zr/HT9 metal fuel pin failure and stochastic clad damage propagation by biased sampling based on a Cumulative Damage Fraction (CDF) type clad failure criterion and the normal distribution of fuel failure probability density as a function of logarithm of Cumulative Damage Fraction. In addition, the effect of post-failure fission gas jet on heat transfer degradation is modeled for the target pins. This model is called stochastic Clad Damage Propagation (CDAP). The CDAP model is now fully integrated into developmental version of PRONGHORN-SC subchannel analysis code, allowing for modeling local failures and its propagation potential. Section 2 describes the components of the CDAP models. Section 3 describes the model implementation to PRONGHORN-SC and input specifications. Section 4 describes the CDAP model validation coupled to PRONGHORN-SC.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Fuel hardware considerations for BWR PCI mitigation

Pellet-clad interaction (PCI) remains a potential cause of fuel failures and is a major consideration in the design of new BWR fuel assemblies. The design and material properties of both the pellet and the cladding have been investigated and optimized to mitigate the impact of PCI. The creep behavior of standard and ADOPT (Advanced Doped Pellet Technology) fuel pellets has been measured across a wide range of temperatures. Dopants have been found to increase the high temperature creep properties of the fuel and can explain the improved PCI resistance of ∼5 kW/m in ramp tests. Other effects, such as chemical trapping of fission products and modified crack patterns cannot be ruled out, but these are at an earlier stage of investigation. It is shown that the enhanced creep effect occurs at the high centerline fuel temperatures during a ramp test, but is not present during the much lower temperatures of commercial fuel operation. Thus low temperature PCI benefits in routine commercial operation are not expected. The development of cladding materials with liner provided a great improvement for PCI mitigation in BWRs. With the softer Westinghouse liner, stepwise ramp tests have shown a benefit of ca 15-20 kW/m. Differences between stepwise and single step ramp tests are discussed. Notably the benefit of the liner remains under all types of fuel operation and temperatures. It is also thought to provide significant benefit in the case of fuel relocation or pellet defects such as missing pellet chips. The results of post irradiation examination (PIE) of a commercial fuel rod that experienced fuel fragment relocation are presented in the paper. To provide the maximum PCI benefit in the next generation BWR fuel design, TRITON11, the proven benefits of both the liner and ADOPT pellets will be included. To reduce the hydrogen pickup, below that of Zircaloy-2, the outer cladding component will be made from HiFi. TRITON11, ADOPT and HiFi are registered trademarks of Westinghouse Electric Company LLC in the United States. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Assessing Manufacturability of the Oxide Dispersion Strengthened (ODS) 14YWT Alloy Fuel Cladding Tube Using SolidStir™ Technology

The 14YWT (Fe–14Cr–3W–0.4Ti–0.3Y2O3 (wt.%)) alloy is a potential candidatematerial for fabricating fuel cladding tubes and other structural components to be used in next-generation advanced nuclear reactors. Due to inherent limitations of current processing routes, this research explored a new variant of friction stir processing, commercially referred to as SolidStir™ Extrusion (patent pending) technology from Enabled Engineering to consolidate and extrude 14YWT powder into a fuel cladding tube. Several single-layer and multiple-layer consolidation experiments were carried out in an Inconel 625 die to determine the suitable processing parameters for SolidStir™ extrusion of the 14YWT alloy cladding tube. With these processing parameters, a 16–18 mm long tube of the 14YWT alloy with no visible defects was successfully extruded using this novel technology. Microstructural characterization results obtained using optical microscopy, scanning electron microscopy, energy dispersive spectroscopy, electron backscatter diffraction, and transmission electron microscopy will be presented.

Advanced Materials and Manufacturing Technology↗
DOE OSTI2016

The Role of Electron and Atom Probe Tomography in Characterization of Nuclear Fuels

Dramatic advances in materials characterization capabilities over the last few years enabled microstructural and chemical 3D characterization of specimens at the atomic-scale with unprecedented quantitative precision and high data collection rates. These improvements allow for the transition of electron tomography (ET) and atom probe tomography (APT) techniques to a much broader research field and their implementation in the characterization of nuclear fuels and highly radioactive materials. Nanoscale characterization has been widely developed for non-nuclear applications, yet in the nuclear arena, particularly for fuels, these modern characterization techniques have not been applied in concert to elucidate fundamental material behavior in metal fuel. In this work, we present cross-correlated APT and ET data and provide atomic-scale representations of microstructure and chemistry of nuclear fuels, which is essential in the validation of current and future atomistic and meso-scale modeling efforts. A firm understanding of microstructural evolution in reactor materials is imperative and will lead to deeper understanding of in-reactor degradation behavior, creating a new generation of radiation-resistant materials and fuels with the potential for significant increases in burn-up. In summary, two complementary techniques were used to characterize phases formed as a result of interdiffusion between U-Pu-Zr metal fuel and cladding and fission gas bubbles produced in U-Mo fuel upon irradiation. Understanding in-reactor degradation behavior of these materials is critical to advancing nuclear energy technology. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Pellet cladding mechanical interaction modeling using the extended finite element method

Fracturing of light water reactor (LWR) UO{sub 2} pellet has multiple important effects on fuel performance. One particularly important concern is that cracks in the fuel cause elevated stresses in the cladding when pellet cladding mechanical interaction (PCMI) occurs. The extended finite element method (XFEM) is a powerful method to represent arbitrary propagating discrete cracks in finite element models. The use of XFEM has been previously demonstrated for modeling propagating discrete cracks in the BISON fuel performance code. This paper demonstrates an initial application of XFEM to model stress concentrations induced by fuel fractures at the fuel/cladding interface during PCMI. This is demonstrated on a study of a pre-defined stationary crack in a 2D cross-section model of a LWR fuel rod. The results from a model with a discrete crack defined with XFEM compare favorably with the results from a model with the same crack geometry defined in the finite element mesh. This study focuses on benchmarking the use of XFEM for PCMI modeling with a stationary crack, but this technique will be readily extended in the future to consider PCMI in conjunction with arbitrary, physics-driven crack propagation. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Determine Operating Reactor to Use for the 2016 PCI Level 1 Milestone

The Consortium for Advanced Simulation of Light Water Reactors (LWRs) (CASL) Level 1 milestone to “Assess the analysis capability for core-wide [pressurized water reactor] PWR Pellet- Clad Interaction (PCI) screening and demonstrate detailed 3-D analysis on selected sub-region” (L1:CASL.P13.03) requires a particular type of nuclear power plant for the assessment. This report documents the operating reactor and cycles chosen for this assessment in completion of the physics integration (PHI) milestone to “Determine Operating Reactor to use for PCI L1 Milestone” (L3:PHI.CMD.P12.02). Watts Bar Unit 1 experienced (at least) one fuel rod failure in each of cycles 6 and 7, and at least one was deemed to be duty related rather than being primarily related to a manufacturing defect or grid effects. This brief report documents that the data required to model cycles 1–12 of Watts Bar Unit 1 using VERA-CS contains sufficient data to model the PHI portion of the PCI challenge problem. A list of additional data needs is also provided that will be important for verification and validation of the BISON results.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Impacts of Fast Critical Assembly Fuel Discards on Liquid Waste Processes

The Savannah River Nuclear Solutions (SRNS) Fast Critical Assembly (FCA) mission is reestablishing the electrolytic dissolver for processing of Pu and Pu-U materials clad in stainless steel (SS). H-Canyon is planning to dissolve and neutralize FCA fuel without recovering the special nuclear material (i.e., Pu) prior to discarding to the Concentration, Storage, and Transfer Facilities (CSTF) operated by the Savannah River Mission Completion (SRMC) Liquid Waste (LW) Organization. The FCA discards will be combined with sludge in the CSTF after Low Temperature Aluminum Dissolution (LTAD) if needed. The combined waste will be washed, concentrated, and vitrified at the Defense Waste Processing Facility (DWPF). The high level waste canisters produced will be temporarily stored in the Glass Waste Storage Buildings prior to transfer to a future federal repository. Decants from LTAD and sludge washing will be combined with DWPF recycle and dissolved salt cake to prepare salt batches for processing in the Salt Waste Processing Facility. The resulting decontaminated salt solution will be processed in the Saltstone Production Facility.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Pilger Mill Processing

The Nuclear Technology Research & Development program of the Office of Nuclear Energy has implemented a program to develop an advanced reactor cladding fabrication capability by using the extrusion and pilgering processes. Oxide-dispersion strengthened (ODS) steels are a promising class of advanced materials, but processes need to be developed to produce tubing for these advanced alloys. The Pacific Northwest National Laboratory (PNNL) has converted its rolling mill into a lab-scale pilger mill, which required designing pilger tooling and rolls that contained multiple pilgering grooves and designing a servo-driven feed mechanism. Upon assembling and checking out the system, pilgering runs on stainless steel, MA956, and 14YWT iron (Fe)-based alloys were completed. The feed rates and rotation rates were established, the reduction schedules were determined. This report will provide a description of the lab-scale pilger mill performance and processing results for the stainless steel, MA956, and 14YWT alloys.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Microstructural and chemical characterization of medium burnup fast reactor MOX using electron microscopy

A thorough post-irradiation examination campaign of medium burnup Sodium Fast Reactor (SFR) mixed oxide (MOX) fuel has been undertaken over the past two years to better understand the performance at intermediate burnup of this annular fuel design. Non-destructive examinations as well as destructive examinations of two pins have been recently published by the authors, providing data regarding the performance at macro-scale. The analyses of selected samples have continued to provide more details regarding the overall microstructure and, particularly, the fuel-cladding interaction.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Effect of Natural Organic Matter on Radionuclide Partitioning in Nuclear Fuel Storage Ponds - 17136

One of the nuclear fuel storage ponds at Sellafield is open to the air, and has contained a significant inventory of corroded Magnox fuel and sludge for several decades, with some fission products having been released into solution. The sludge is predominantly composed of brucite (Mg(OH){sub 2}), a corrosion product from the magnesium-based fuel cladding, amongst other corrosion products resulting from the uranium metal fuel. Sr-90 is known to constitute a small mass of the radionuclides present in the pond, but due to its solubility and activity, it is a radionuclide that is at risk of challenging effluent discharge limits. Organic molecules are known to be present also due to the ponds being open to the air, with occasional algal blooms restricting visibility. Understanding the chemical interactions of these components is important to inform ongoing sludge retrievals and effluent management. Batch sorption-desorption experiments have been performed with brucite, Sr-90 and natural organic matter (NOM) (namely humic acid (HA) and Pseudanabaena catenata cyano-bacterial growth supernatant) in both binary and ternary systems at high pH. Ionic strength, pH and order of addition of components were varied. Sr-90 was shown not to interact strongly with the bulk brucite surface in binary or ternary systems under pH conditions relevant to the pond. Humic acid in both binary and ternary systems demonstrated a strong affinity for the brucite surface. In ternary systems HA appeared to enhance sorption of Sr-90, and vice versa, regardless of the order of addition of components. There was no distinct effect of varying ionic strength or pH. (authors)

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

Pellet Cladding Mechanical Interaction Modeling Using the Extended Finite Element Method

As a brittle material, the ceramic UO2 used as light water reactor fuel experiences significant fracturing throughout its life, beginning with the first rise to power of fresh fuel. This has multiple effects on the thermal and mechanical response of the fuel/cladding system. One such effect that is particularly important is that when there is mechanical contact between the fuel and cladding, cracks that extending from the outer surface of the fuel into the volume of the fuel cause elevated stresses in the adjacent cladding, which can potentially lead to cladding failure. Modeling the thermal and mechanical response of the cladding in the vicinity of these surface-breaking cracks in the fuel can provide important insights into this behavior to help avoid operating conditions that could lead to cladding failure. Such modeling has traditionally been done in the context of finite-element-based fuel performance analysis by modifying the fuel mesh to introduce discrete cracks. While this approach is effective in capturing the important behavior at the fuel/cladding interface, there are multiple drawbacks to explicitly incorporating the cracks in the finite element mesh. Because the cracks are incorporated in the original mesh, the mesh must be modified for cracks of specified location and depth, so it is difficult to account for crack propagation and the formation of new cracks at other locations. The extended finite element method (XFEM) has emerged in recent years as a powerful method to represent arbitrary, evolving, discrete discontinuities within the context of the finite element method. Development work is underway by the authors to implement XFEM in the BISON fuel performance code, and this capability has previously been demonstrated in simulations of fracture propagation in ceramic nuclear fuel. These preliminary demonstrations have included only the fuel, and excluded the cladding for simplicity. This paper presents initial results of efforts to apply XFEM to model stress concentrations induced by fuel fractures at the fuel/cladding interface during pellet cladding mechanical interaction (PCMI). This is accomplished by enhancing the thermal and mechanical contact enforcement algorithms employed by BISON to permit their use in conjunction with XFEM. The results from this methodology are demonstrated to be equivalent to those from using meshed discrete cracks. While the results of the two methods are equivalent for the case of a stationary crack, it is demonstrated that XFEM provides the additional flexibility of allowing arbitrary crack initiation and propagation during the analysis, and minimizes model setup effort for cases with stationary cracks.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Processing and Characterization of Pilgered 14YWT Tubing

The Nuclear Technology Research & Development program of the Office of Nuclear Energy has implemented a program to develop an advanced reactor cladding fabrication capability by using the extrusion and pilgering processes. Oxide-dispersion strengthened (ODS) steels are a promising class of advanced materials, but processes need to be developed to produce tubing for these advanced alloys. The Pacific Northwest National Laboratory (PNNL) has converted its rolling mill into a lab-scale pilger mill, which required designing pilger tooling and rolls that contained multiple pilgering grooves and designing a servo-driven feed mechanism. Upon assembling and checking out the system, pilgering runs on stainless steel, MA956, and 14YWT iron (Fe)-based alloys were completed. Once the feed rates and rotation rates were established, the reduction schedules were determined. This report will focus on the processing and characterization of pilgered 14YWT thin-wall tubing that was processed from an extruded thick-wall tube using the PNNL laboratory pilger rolling mill. Characterization included optical metallography of the pilgered tube and microhardness of the processed material. The 14YWT characterization is being done in cooperation with researchers at Los Alamos National Laboratory (LANL), with LANL performing mechanical property testing of pilgered 14YWT and 14YWT tubes produced by hydrostatic extrusion at Case Western Reserve University (CWRU).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

A non-fueled nuclear-heated rod for in-pile transient boiling studies

Separate-effects boiling experiments have recently been conducted in the Transient Reactor Test Facility at Idaho National Laboratory to investigate transient heating and irradiation effects on cladding-to-coolant heat transfer. Specifically, transient critical heat flux (CHF) remains an important area of research, and better understanding of this phenomenon has potential for improving predictive models related to operational and safety limits. Consequently, this knowledge is expected to improve efficiency of light-water reactor operations. A novel borated nuclear-heated rodlet (BNHR) was designed to enable observation of transient cladding-to-coolant heat transfer phenomena. The final BNHR design takes a surrogate approach, wherein nuclear heating is induced by 10 B(n, α) reactions rather than derived from fissions in a fueled specimen. The structure of the BNHR consists of a hollowed out borated (B nat ~ 2.05 wt %) stainless steel tube with an hourglass-shaped outer surface, capped at both ends with non-borated stainless steel. This geometry allows for inner-rodlet instrumentation and generation of the highest nuclear heating rates near the center of the rodlet to ensure onset of boiling near instrumentation for real-time observation. A novel approach to measuring the nuclear energy deposition rate in the BNHR separate and apart from the influence of the coolant, termed the n-a thermometer, is also detailed in this paper. This device has demonstrated excellent repeatability, and measurements indicate predictive modeling results for energy deposition in the BNHR rod agree within a 10% margin of the experiment measurements. In conclusion, these results give confidence that the BNHR design has successfully met experiment objectives.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Raman spectroscopy of zirconium hydride

The characterization of zirconium hydride is important in the nuclear industry because of the hydrogen-induced embrittlement of Zircaloy cladding and its use as a neutron moderator. This paper introduces the use of Raman spectroscopy for the characterization of zirconium hydride. First-principles density functional theory (DFT) calculations were used to predict the Raman spectra of ζ-ZrH 0.5 , γ-ZrH, δ-ZrH 1.5 , δ-ZrH 1.66 , and ε-ZrH 2 with all their predicted symmetries; ζ-ZrH 0.5 (P3m1, R$\overline{3}$m, C2/m, Cm, Cmmm, and Pn$\overline{3}$m); γ-ZrH (P222, Ccce, and P4 2 /mmc); δ-ZrH 1.5 (P$\overline{4}$m2, P4 2 /mcm, Fmmm, Pn$\overline{3}$m, Ibam, P2/c, PI, and P4 2 /nnm); δ-ZrH 1.66 (Fm$\overline{3}$m); and ε-ZrH 2 (Fm$\overline{3}$m, R$\overline{3}$m, and I 4 /mmm). Two samples of Zircaloy-4 containing 133 wt ppm and 360 wt ppm hydrogen were characterized by Raman spectroscopy, showing two signal lines at 215 cm -1 and 1,187 cm -1 , which were assigned to the presence of δ-ZrH 1.66 . These signals had a good spatial correlation with visible hydride precipitates in Raman spectroscopy images. In conclusion, this work provides the basis for the characterization of all possible zirconium hydride compositions and structures using Raman spectroscopy.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Transmission electron microscopy characterization of the fuel-cladding chemical interactions in HT9 cladded U-10Zr fuel

Fuel cladding chemical interaction (FCCI) is a key phenomenon needs to be better understood to establish the design basis for U-10Zr metallic fuel performance. Characterizing the microstructure and chemical composition of FCCI at micron and sub-micron scale is critically important toward a more mechanistic understanding of FCCI phenomenon and its potential effects on cladding integrity and metallic fuel performance. This paper, by using transmission electron microscopy, investigated the FCCI region in HT9 cladded U-10Zr fuel irradiated to 5.7% FIMA burnup at a peak inner cladding temperature of 615 °C in Fast Flux Test Facility (FFTF). Four distinct layers are identified in the FCCI region. The migration of Fe into the fuel side leads to the formation of several U-Zr-Fe ternary phases, including ?-Fe0.5Zr0.32U0.18, e-Fe0.3Zr0.4U0.3, and ?-Fe0.06Zr0.23U0.71, mingled with UFe2, U6Fe, and U phase at various Fe penetration depth up to ~ 150 µm. On the cladding side, grain coarsening and significant lanthanides infiltration along grain boundaries are observed. Laves phase, (Fe,Cr)2(Mo,W), which typically does not exist in fresh HT9, is identified in a wide radial range in the cladding. The typical HT9 martensitic lath structure and pre-existing M23C6 precipitates disappear, partially or completely, depending on the radial distance from the fuel-cladding interface. Those microstructural and compositional changes could cause mechanical degradation in the HT9 cladding. The present characterization results will improve the understanding of FCCI phenomenon and facilitate the development of microstructure-informed FCCI modeling for metallic fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Enhanced Thermal Conductivity UO{sub 2}-BeO and UO{sub 2}-SiC Fuels Behavior

Commercial light-water reactors currently use UO{sub 2} as a nuclear fuel, which has demonstrated several desirable characteristics. However, UO{sub 2} has a low thermal conductivity that leads to the development of a large temperature gradient across the fuel pellet. This temperature gradient contributes to limits in the operational performance of the reactor due to effects that include thermal stresses causing pellet cladding interaction and the release of fission product gases as demonstrated. These high fuel temperatures can be decreased and reactor performance improved by developing an enhanced thermal conductivity nuclear fuel. A high thermal conductivity nuclear fuel would decrease fuel temperatures and facilitate a reduction in pellet cladding interaction through lessening thermal stresses that result in fuel cracking, relocation, and swelling as described. Additionally, fission gas release would be decreased allowing for higher fuel burn-up, and the safety of the reactor would be improved with a faster thermal response and less stored energy in the fuel pins. A ceramic-ceramic composite nuclear fuel has an increased effective thermal conductivity due to adding a second compatible and high thermal conductivity ceramic to UO{sub 2}. SiC and BeO are two materials that demonstrate compatibility with UO{sub 2} and have a high thermal conductivity. In summary, the enhanced thermal conductivity UO{sub 2}-BeO and UO{sub 2}-SiC fuel performance has been modelled in a light water reactor, and a more complete and detailed understanding was achieved for enhanced thermal conductivity UO{sub 2}-BeO and UO{sub 2}-SiC fuels. BeO and SiC have high melting point, low neutron absorption, low thermal expansion coefficient and chemical compatibility with current UO{sub 2} fuels which makes them a suitable material for thermal conductivity enhancement. In our modeling and simulation work, the enhanced thermal conductivity nuclear fuel showed the capability of decreasing the fuel temperatures, increasing the gap total conductance, reducing the fission gas release, decreasing the gap/plenum pressure, lessening oxygen redistribution, and facilitating a reduction in pellet cladding interaction through lessening thermal stresses that result in fuel cracking, relocation, and swelling. The safety of the reactor operation would be then improved. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Interaction of Polymethyl Methacrylate with Boehmite-Filmed Aluminum Cladding Under Gamma Irradiation

This paper presents an overview of ongoing work to qualify the Advanced Test Reactor (ATR) driver fuel elements that have been affected by irradiation-degraded polymethyl methacrylate (PMMA) flux wands. Irradiation testing was performed on PMMA material in contact with aluminum clad material. The cladding was prefilmed with a boehmite oxide layer, an important feature of the ATR driver fuel. The effects on the boehmite layer due to gamma irradiation of the PMMA-aluminum clad system were investigated. PMMA embrittlement, followed by softening and degradation, occurred at high radiation levels. Adhesion between the cladding and irradiated PMMA was observed. Flow testing at prototypic ATR flow rates demonstrated the effective removal of the adhered material. Measurements of the boehmite layer thickness were performed, and Raman spectroscopy was utilized to detect the presence of boehmite in the irradiated PMMA material.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗