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At least 541 records · Page 30
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Effects of Neutron Irradiation on the Bond Strength of Steel Embedded in Concrete

The structural integrity of the concrete biological shield (CBS) in light water reactors (LWRs) exposed to long-term radiation raises the question of the effects of fast-neutron and gamma irradiation on the bond properties of embedded carbon-steel reinforcement in concrete, which is critical to the transfer of in-service and accidental loads. The design and execution of a first-of-a-kind irradiation experiment conducted in the LVR-15 test reactor (Czech Republic) are presented along with preliminary post-irradiation examination and testing results. Preliminary analysis shows that the fast neutron flux affects the radiation-induced volumetric of concrete aggregate, that the gamma irradiation induces shrinkage of the chemically bonded water in cementitious hydrates and that the bond strength of steel in concrete at the tested irradiation conditions decreases moderately.

Le Pape, Yann↗
DOE OSTI2024

Quantitative Insight to Fission Gas Pores Distribution in Irradiated Annular U-10Zr Metallic Fuel Using Machine Learning

Metallic fuels, particularly U-10Zr and its performance in reactor irradiation conditions, have been thoroughly investigated and are a promising candidate for next-generation sodium-cooled fast spectrum nuclear reactors. Irradiation in reactors can lead to the formation of fission gas and increased pore formation which can significantly impact fuel performance. Due to the large number of pores and various phases formed in metallic fuel during irradiation, a quantitative description of fission gas pores as a function of irradiation conditions is not yet available, undermining the fidelity of fuel performance modeling to support fuel qualification. It has been difficult to clearly detect pore boundaries and distinguish matrix phases from fission gas pores using optical microscopy by using simple threshold methods working with low magnification images. The pre-trained deep learning model for fission gas pore detection was applied to ~10,260 high magnification scanning electron microscopy images. The model increased the accuracy of fission gas pore segmentation to obtain statistical features, which cannot be processed manually. A pre-trained decision tree model was used to classify pores as isolated or connected pores, providing new insight into the correlation between the movement of lanthanides, solid fission products, and the radial temperature gradient developed in fuel irradiation conditions. This paper emphasizes the potential that artificial intelligence-based machine learning models have to accelerate qualification and support nuclear fuel development.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Impact of Irradiation on Microstructure and Mechanical Properties of Materials Produced by Advanced Manufacturing

Advanced non-light water reactor designs, known as Generation IV (Gen IV) reactors, typically operate at higher temperatures and under more extreme radiation conditions than conventional light water reactors. A critical aspect of the successful deployment and advancement of Gen IV reactor designs is the selection of appropriate structural materials for specific applications, which necessitates the timely development of new materials and manufacturing processes. Advanced manufacturing (AM) offers numerous opportunities for innovative designs, enabling the production of high-performance components with potentially shorter development cycles compared to traditional manufacturing methods. However, a significant challenge in deploying AM technologies in the nuclear energy sector is the current lack of data on the irradiation performance of AM-produced components. This presentation will discuss neutron and ion irradiation results of AM materials, including stainless steel 316L, Grade 91, SA508, Inconel 718, and Inconel 625 materials. The materials were manufactured by AM, such as Powder Metallurgy Hot Isostatic Pressing (PM HIP), Laser Powder Bed Fusion (LPBF) and Directed Energy Deposition (DED). The effects of neutron irradiation on microstructure (e.g., dislocations, loops, and nanoclusters), tensile properties (e.g., strength and ductility), and ion irradiation on Irradiation-Assisted Stress Corrosion Cracking (IASCC) will be explored. Results are collected from several projects supported by the Nuclear Science User Facilities (NSUF) program.

36 - MATERIALS SCIENCE↗
DOE OSTI2025

Future Opportunities for LWR Irradiations in US Test Reactors

After several years of relatively low activity in the field of Light Water Reactor (LWR) fuel development, the Department of Energy again began to engage in developing new fuel technologies and irradiation performance data prompted by the Fukushima Daichi nuclear accidents. New competencies for irradiation testing in material test reactors in the United States began to be developed at this time using the Advanced Test Reactor (ATR), High Flux Isotope Reactor (HFIR), Massachusetts Institute of Technology Reactor (MITR), and the Transient Reactor Test Facility (TREAT). Capsules for testing fuel and cladding materials in ATR and HFIR were deployed, a Pressurized Water Reactor (PWR) condition loop for testing fuel rods was established in ATR, cladding corrosion studies were performed using a water loop in MITR, and TREAT pulse testing capabilities were commissioned for fuel rods in water capsules. The more recent and unexpected closure of the Halden Boiling Water Reactor (HBWR) also prompted further investments in Loss of Coolant Accident (LOCA) testing capabilities at TREAT. New configurations of these test devices show further potential in enhanced steam condition control and other investigations are building toward a flowing water loop for testing transient to dryout conditions. The closure of HBWR also prompted a major project currently underway to construct additional water loops in ATR where a novel approach is being pursued to enable Boiling Water Reactor (BWR) conditions. A meaningful collaborative project was awarded to MITR which, amidst an unexpected major overhaul of the reactor, has expanded cladding corrosion test capabilities at MITR. New explorations have led to methods for unique experiments at HFIR including channel box irradiations. New device developments are also bridging toward future potential for instrumented capsule irradiation tests in ATR and HFIR. Finally, a new project referred to as the System Physics Advanced Reactor Critical facility (SPARC) is gaining traction towards a large zero-power reactor able to produce physics validation data for LWR fuel bundle designs with increased enrichment and enhanced absorbers for 24-month operation cycles. This paper provides a brief summary of the status of these irradiation testbed capabilities with an emphasis on current efforts toward future capabilities to obtain new data and maximize the performance potential of LWR fuel technologies.

Woolstenhulme, Nicolas [Idaho National Laboratory ↗
DOE OSTI2025

Deconsolidation and Leach Burn Leach of Seven As Irradiated AGR 5/6/7 TRISO Fuel Compacts from Capsules 2, 3, 4, and 5

Seven as-irradiated Advanced Gas Reactor (AGR) 5/6/7 compacts underwent destructive post-irradiation examination via deconsolidation-leach-burn leach at Idaho National Laboratory (INL). The selection of the compacts extended the upper and lower limits of time-average volume-average (TAVA) temperature for compacts that had gone through deconsolidation-leach-burn leach so far. The measured inventories of fission products and actinides in the compact matrix and outer pyrolytic carbon were reported. Results indicated unexpectedly higher rates of fuel kernel leaching compared to compacts from AGR-1 and AGR-2. These failure rates were attributed to damage during post-irradiation sample handling, rather than irradiation itself. AGR-5/6/7 compacts have little or no matrix coverage for some particles at the top and bottom ends of cylindrical fuel compacts, making them more fragile. Sixty particles were randomly sampled from each compact, and the gamma results were reported. Three SiC shells from were identified, one of which showed signs of chemical attack in the high-irradiation temperature compact.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

USABILITY EXTENSION OF THE ŠKODA VPVR/M CASK FOR TRANSPORTING IRRADIATED FUEL ASSEMBLIES

New abstract from the final version being submitted now: This paper discusses the evolution and enhanced usability of the ŠKODA VPVR/M cask for the transport of irradiated fuel assemblies, particularly within the context of the demand for the delivery of newly appearing irradiated HEU fuel types for which the cask did not yet have a license to transport. Over time, the cask’s internal basket construction has demonstrated notable adaptability to accommodate various exotic HEU fuel types from research reactors of differing origins. The paper outlines sever-al custom internal baskets developed for specific fuel types, including those from Belarus and Serbia, as well as from Georgia, Uzbekistan, and MNSR cores, as well as a recently designed basket for MTR and TRIGA assemblies. The findings high-light the high flexibility and adaptability of the cask, supported by successful rede-signs and licensing efforts, underscoring its value for the safe and secure transport of nuclear material. Old abstract from the draft version that was already approved: The Russian Research Reactor Fuel Return (RRRFR) program, since its inception, has continuously used the ŠKODA VPVR/M Cask fleet designed for the repatriation of irradiated highly enriched uranium (HEU) fuel. As the program progressed (from shut-down and a quasi-abandoned reactor, and/or as it began to include fuels of Chinese and US origin), new challenges emerged for the transport Cask. These were fuel types that had not yet been licensed for the Cask. Although these requirements did not arise during the design of the basic ŠKODA VPVR/M Cask, as revealed by the retrospective analyses, the Internal Basket of the ŠKODA VPVR/M Cask gives a high degree of flexibility to accommodate additional fuel types. This paper provides a brief overview of the ŠKODA VPVR/M Cask, which holds a B(U) type license, and introduces the different types of Internal Baskets that have already been licensed to transport so-called exotic irradiated HEU fuel types, in addition to the original license. The paper presents a new Internal Basket design for accommodating MTR-type and TRIGA-type irradiated HEU fuel assemblies. This includes a detailed presentation of the design basis and the new MTR-TRIGA Internal Basket, as well as the licensing matters of the package under the name ŠKODA MTR-TRIGA Cask, and the conformity test (dry- and wet-run) operations made to verify compliance with the new Internal Basket. Then, as a summary, the usage record for the Cask fleet is presented, and finally, the paper concludes with the consolidated experiences gained during the utilization of the ŠKODA VPVR/M Cask fleet, emphasizing the high degree of Cask flexibility ensured by the Internal Basket’s construction.

42 - ENGINEERING↗
DOE OSTI2026

Effects of ZrN Coating and Heat Treatment on U-Mo Dispersion Fuel under Irradiation

The stability of U-Mo fuel particles in Al matrix can be improved using ZrN coating and heat treatment. This study investigated irradiation behaviors of fuel plates containing U-Mo fuel particles fabricated with different conditions including heat treatment and the thickness of ZrN coating. Additionally, different fission densities were applied to each fuel plate to understand how varying fission densities affect its irradiation behaviors. This study showed that more recrystallization of grains and development of high burnup structure (HBS) occurred in irradiated fuel particles with higher fission densities. Heat treatment reduced the accumulation of fission gas bubbles in the fuel particles by coarsening their grains. The fuel particles with a ZrN coating thickness of 1.2-1.8 µm showed less significant development of U-Mo/Al interaction layers than the particles with a coating thickness of 0.5 µm after irradiation. This might indicate the existence of a critical thickness of ZrN coating to minimize the formation of the interaction layers. Fission gas bubbles were observed at grain boundaries of irradiated U-Mo fuel particles at low fission densities, and the formation of dendritic features with a darker contrast was identified in backscattered electron images. These features stemmed from grain boundaries with fission gas bubbles or HBS, and some of them contained the bubbles, indicating that they might represent an early stage of HBS propagation in the fuel particles.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Stacked target irradiations and radiochemical separations for simultaneous production of short-lived Ni and Ir isotopes

A method was developed for simultaneous production of neutron-deficient Ni and Ir isotopes by irradiation of stacked targets with a 50.5 MeV proton beam, followed by radiochemical separations to remove excess target material. The target, irradiation configuration, and post-irradiation radiochemical separations were developed and tested in a proof-of-principle irradiation. Radiochemical yields and purity of the products are discussed. Finally, the stacked target method, developed herein, can be used to produce other isotope pairs with appropriate beam tailoring.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2020

In situ microstructural evolution in face-centered and body-centered cubic complex concentrated solid-solution alloys under heavy ion irradiation

This study characterizes the microstructural evolution of single-phase complex concentrated solid-solution alloy (CSA) compositions under heavy ion irradiation with the goal of evaluating mechanisms for CSA radiation tolerance in advanced fission systems. Three such alloys, Cr 18 Fe 27 Mn 27 Ni 28 , Cr 15 Fe 35 Mn 15 Ni 35 , and equimolar NbTaTiV, along with reference materials (pure Ni and E90 for the Cr-FeMnNi family and pure V for NbTaTiV) were irradiated at 50 K and 773 K with 1 MeV Kr ++ ions to various levels of displacements per atom (dpa) using in-situ transmission electron microscopy. Cryogenic irradiation resulted in small defect clusters and faulted dislocation loops as large as 12 nm in face-centered cubic (FCC) CSAs. With thermal diffusion suppressed at cryogenic temperatures, defect densities were lower in all CSAs than in their less compositionally complex reference materials indicating that point defect production is reduced during the displacement cascade stage. High temperature irradiation of the two FCC CSA resulted in the formation of interstitial dislocation loops which by 2 dpa grew to an average size of 27 nm in Cr 18 Fe 27 Mn 27 Ni 28 and 10 nm in Cr 15 Fe 35 Mn 15 Ni 35 . This difference in loop growth kinetics was attributed to the difference in Mn-content due to its effect on the nucleation rate by increasing vacancy mobility or reducing the stacking-fault energy.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Characterization of boundary precipitation in a heavy ion irradiated tungsten heavy alloy under the simulated fusion environment

In the concerted effort to identify materials capable of surviving the adverse environment of a fusion reactor interior, tungsten heavy alloys have been put forth as leading candidates. Experimental trials and behavioral studies have yielded positive results for their adoption by taking advantage of the alloy’s unique balance of inherently high fracture toughness and low sputter yields; yet due to their relative novelty in the fusion community, there remains a lack of understanding on the response of these materials to the extended high temperature irradiation environment of the reactor interior. Further, to alleviate this issue and provide the necessary data on the behavior of tungsten heavy alloys to the simulated fusion environment, a 90W-7Ni-3Fe alloy has been subjected to elevated temperature sequential Ni + and He + ion irradiations to mimic the expected displacement damage and gas production expected after five years of service as a plasma facing material component. Atomic-scale structural analyses and nanoscale chemical mapping have identified the formation of two distinct precipitation structures, a surface localized η-carbide and an a-W 2 C type tungsten carbide, both of which appear to originate at the bi-phase interface between W and the ductile phase. This irradiation assisted and induced precipitate formation respectively is anticipated to adversely affect these materials by selective embrittlement at bi-phase interfaces leading to a reduction in the material’s overall fracture toughness during prolonged high temperature irradiation. It is asserted that any potential exposure to C during fusion reactor operational service should be minimized as to prevent the formation of these phases.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Exploring constituent redistribution in irradiated U-19Pu-14Zr fuel via electron probe microanalysis

Here, the phenomena of constituent redistribution, wherein a previously homogeneous metallic fuel forms discrete, radially concentric compositional zones upon irradiation was investigated by examining an irradiated U-19Pu-14Zr fuel (where numbers represent wt. %) with a burnup of 11.5 at.% with electron probe microanalysis (EPMA) and quadruple inductively coupled plasma mass spectroscopy (Q-ICP-MS). EPMA-generated U, Pu, and Zr compositional data obtained from a diameter traverse of the sample was converted to mass and was used to: 1) compare the overall fuel element analysis results between the two methods, 2) determine the number of compositionally distinct zones forming as a result of constituent redistribution; and 3) quantify the post-irradiation loss or gain of U, Pu, and Zr atoms in each distinct compositional zone. Weight percent concentrations of U, Pu, and Zr for the overall cross section compare favorably between the two analytical methods, suggesting that the spatially resolved EPMA analysis complements bulk chemical analysis. Among the four identified compositional zones, post-irradiation quantification of U, Pu, and Zr elemental atom content changes shows that the quantity of U atoms lost from the innermost zone is slightly less than the quantity of U atoms gained by the middle two zones, and the quantity of Zr atoms lost from the high-U third zone is slightly less than is gained by the two innermost zones. Pu is lost from all four zones, although the innermost zone and the high-U third zone lose a significantly higher percentage (> 22 %) of their initial Pu atoms than the other two zones. For all three elements, EPMA cannot distinguish between atoms lost due to transport to a different zone from atoms lost due to nuclear processes; however, the insight gained from using this process can be used to experiment with new modeling techniques to predict constituent redistribution in U-Pu-Zr fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Irradiation-induced segregation at dislocation loops in CoCrFeMnNi high entropy alloy

To understand the redistribution of alloying elements in high entropy alloys under irradiation, we irradiated a CoCrFeMnNi alloy with 1 MeV Kr ions at room temperature and at 500°C, and characterized with atom probe tomography and transmission electron microscopy. At 500°C, Co and Ni were enriched around the interstitial, faulted and perfect, dislocation loops resulted from the ion irradiation. In contrast, no segregation was observed at room temperature. The inverse Kirkendall effect through vacancy flux, as opposed to the interstitial binding mechanism, was the primary underlying process attributing to the observed segregation. In addition, a ring-shaped segregation pattern was observed at the faulted dislocation loops, indicating a non-equilibrium nature of the defect clustering and solute segregation process in CoCrFeMnNi under irradiation at high temperature.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Quantum Emitters Induced by High Pressure and UV Laser Irradiation in Multilayer GaSe

In this work, we report on defect generation in multilayer GaSe through hydrostatic pressure quenching and UV laser irradiation. The Raman line width from the UV 266 nm irradiated sample is much wider than that in pressure-quenched GaSe, corresponding to a wider defect energy distribution range in the former sample than the latter. After quenching from 11.2 GPa, three photoluminescence (PL) peaks from defect states are observed at 657, 681, and 695 nm at a low temperature of 93 K. Defect-related peaks at 649, 694, 750, and 774 nm also appear in low-temperature PL spectra after UV laser irradiation, with a nonmonotonous intensity dependence on irradiation duration. There are common features in defects produced by these two methods: the PL peaks with the lowest energy are sharp, and their PL intensities increase linearly with the excitation laser power and saturate above a certain excitation laser power. These two features are similar to those in defects for single-photon emission (SPE) in other 2D materials at even lower temperatures. Fluorescence lifetime imaging shows distinguished short (2.3 ns) and long (75.6 nm) lifetimes of the 695 nm PL line in pressure-quenched GaSe. The density functional theory predicts defect energy levels related to Se vacancy.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2021

Conceptual design of HFIR irradiation experiment for material compatibility study on liquid Sn divertor

Liquid Sn is one of the promising coolants for liquid surface divertor concept of fusion reactors. However, the compatibility between liquid Sn and structural materials is an important issue that has to be addressed, because liquid Sn is extremely corrosive to steels at high temperatures. The corrosion may be mitigated when a protective Al2O 3 layer is formed on the surface of alumina forming steels. However, the effect of neutron irradiation on the integrity of protective layer is not made clear so far. Japan and US joint research project “FRONTIER” started in 2019 to investigate the material compatibility under neutron irradiation. The purpose of the present study is to develop the conceptual design of the irradiation test capsule which enables material compatibility tests for the alumina forming steels - liquid metal systems under neutron irradiation in the High Flux Isotope Reactor at Oak Ridge National Laboratory, TN, USA. The three dimensional drawing of capsule structure was then developed. The validity of the material selections for the capsule design was investigated by means of corrosion tests of SiC, Si3N4, Ti, and Mo in liquid Sn at 773 K for 262 hr.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2026

Downwelling Shortwave and Longwave Irradiance from CC-RIDER on Mount Soledad

The Clouds and Climate - Remote Integrated Deployment of Radiometers (CC-RIDER) is a suite of five Eppley Laboratory (Inc.) instruments that was deployed during EPCAPE at the secondary Mount Soledad site in La Jolla, CA. A primary and backup Precision Spectral Pyranometer (PSP Primary, PSP Backup) measured broadband downwelling shortwave irradiance in the spectral interval 280-2800 nm. A third Precision Spectral Pyranometer (PSP NIR) was fitted with a near-infrared long pass filter and measured downwelling shortwave irradiance in the spectral interval 780-2800 nm. A Total Ultraviolet Radiometer (TUVR) measured broadband downwelling broadband ultraviolet irradiance in the spectral interval 295-385 nm. A Precision Infrared Radiometer (PIR) Pyrgeometer measured downwelling broadband longwave irradiance in the spectral interval 3.5 - 50 microns. Data collection began on 18 April 2023 at 21:31 UTC and ended on 20 February 2024 at 22:34 UTC. Data were recorded by a Campbell Scientific (Inc.) CR1000X datalogger in one-minute intervals, for a total of 443584 data records. The datalogger was solar powered enabling data collection to proceed without interruption from start to finish.

Clouds and Climate – Remote Integrated DEployement↗
DOE OSTI2023

ASME Irradiation Model

ASME Irradiation Model: How to deal with irradiation data in ASME code rules. Includes discussion on challenges to nuclear graphite, such as lack of sufficient irradiation data to qualify graphite for nuclear application for all grades and temperatures, variety of grades, and time/room in available MTRs to get all the required data. Discussion of Irradiation behavior, Leveraging data generated by GIF countries the past 20 years, Dimensional Change Theory, behavior, analysis, turnaround, and data. Future discussions include Arrhenius approach to predict turnaround behavior for the ASME MDS requirements for all nuclear graphite.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Irradiation-Induced Extremes Create Hierarchical Face-/Body-Centered-Cubic Phases in Nanostructured High Entropy Alloys

A nanoscale hierarchical dual-phase structure is reported to form in a nanocrystalline NiFeCoCrCu high-entropy-alloy (HEA) film via ion irradiation. Under the extreme energy deposition and consequent thermal energy dissipation induced by energetic particles, a fundamentally new phenomenon is revealed, in which the original single-phase face-centered-cubic (FCC) structure partially transforms into alternating nanometer layers of a body-centered-cubic (BCC) structure. The orientation relationship follows the Nishiyama–Wasser-man relationship, that is, (011) BCC || (1¯1¯1) FCC and [100] BCC || [11¯0] FCC . Simulation results indicate that Cr, as a BCC stabilizing element, exhibits a tendency to segregate to the stacking faults (SFs). Furthermore, the high densities of SFs and twin boundaries in each nanocrystalline grain serve to accelerate the nucleation and growth of the BCC phase during irradiation. By adjusting the irradiation parameters, desired thicknesses of the FCC and BCC phases in the laminates can be achieved. Finally, this work demonstrates the controlled formation of an attractive dual-phase nanolaminate structure under ion irradiation and provides a strategy for designing new derivate structures of HEAs.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Application of radiochromic gel dosimetry to commissioning of a megavoltage research linear accelerator for small–field animal irradiation studies

To develop and implement an efficient and accurate commissioning procedure for small-field static beam animal irradiation studies on an MV research linear accelerator (Linatron-M9) using radiochromic gel dosimetry. The research linear accelerator (Linatron-M9) is a 9 MV linac with a static fixed collimator opening of 5.08 cm diameter. Lead collimators were manually placed to create smaller fields of 2 × 2 cm 2 , 1 × 1 cm 2 , and 0.5 × 0.5 cm 2 . Relative dosimetry measurements were performed, including profiles, percent depth dose (PDD) curves, beam divergence, and relative output factors using various dosimetry tools, including a small volume ionization chamber (A14), GAFCHROMIC™ EBT3 film, and Clearview gel dosimeters. The gel dosimeter was used to provide a 3D volumetric reference of the irradiated fields. The Linatron profiles and relative output factors were extracted at a reference depth of 2 cm with the output factor measured relative to the 2 × 2 cm 2 reference field. Absolute dosimetry was performed using A14 ionization chamber measurements, which were verified using a national standards laboratory remote dosimetry service. Absolute dosimetry measurements were confirmed within 1.4% (k = 2, 95% confidence = 5%). The relative output factor of the small fields measured with films and gels agreed with a maximum relative percent error difference between the two methods of 1.1 % for the 1 × 1 cm 2 field and 4.3 % for the 0.5 × 0.5 cm 2 field. These relative errors were primarily due to the variability in the collimator positioning. The measured beam profiles demonstrated excellent agreement for beam size (measured as FWHM), within approximately 0.8 mm (or less). Film measurements were more accurate in the penumbra region due to the film’s finer resolution compared with the gel dosimeter. Following the van Dyk criteria, the PDD values of the film and gel measurements agree within 11% in the buildup region starting from 0.5 cm depth and within 2.6 % beyond maximum dose and into the fall-off region for depths up to 5 cm. The 2D beam profile isodose lines agree within 0.5 mm in all regions for the 0.5 × 0.5 cm 2 and the 1 × 1 cm 2 fields and within 1 mm for the larger field of 2 × 2 cm 2 . The 2D PDD curves agree within approximately 2% of the maximum in the typical therapy region (1–4 cm) for the 1 × 1 cm 2 and 2 × 2 cm 2 and within 5% for the 0.5 × 0.5 cm 2 field. This work provides a commissioning process to measure the beam characteristics of a fixed beam MV accelerator with detailed dosimetric evaluation for its implementation in megavoltage small animal irradiation studies. Radiochromic gel dosimeters are efficient small-field relative dosimetry tools providing 3D dose measurements allowing for full representation of dose, dosimeter misalignment corrections and high reproducibility with low inter-dosimeter variability. Altogether, radiochromic gels are valuable for fast, full relative dosimetry commissioning in comparison to films for application in high-energy small-field animal irradiation studies.

62 RADIOLOGY AND NUCLEAR MEDICINE↗