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At least 37 records · Page 2

Benchmarking MCNP models to support gamma irradiation experiments

Seismic isolators and damping devices have been used for 30+ years to reduce seismic forces and deployed in 10,000+ structures, including bridges, buildings, and mission-critical infrastructure. Studies have shown that seismic isolation of nuclear power plants can reduce the seismic risk by several orders of magnitude and substantially reduce the overnight capital cost. Advanced reactor developers are considering seismic isolation as an integral design feature in their reactor design. It can either be deployed at the building level, isolating the entire reactor building, or at the component level, isolating individual pieces of safety-class equipment, such as a reactor vessel or a steam generator. For equipment isolation, located close to the reactor vessel, the seismic isolators and/or damping devices may be exposed to gamma and neutron radiation, with a possible effect on their mechanical properties which would affect the response of the isolated equipment under earthquake shaking. The Department of Energy, as part of its Nuclear Energy University Program (NEUP), in collaboration with the Idaho National Laboratory (INL), is funding a project at the University at Buffalo (UB) to investigate the effect of gamma radiation on the mechanical properties of various seismic protective devices. (When isolating equipment, the seismic protective devices will be located outside the reactor vessel where the neutron exposure will be insignificant.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗

Interinstitutional Study of the New EURO-GANEX Process Resistance by Gamma Irradiation Test Loops

As part of the homogeneous actinides recycling strategy, the EURO-GANEX process is one of the most promising options to achieve the goal of minor actinides recovery. However, EURO-GANEX also has various drawbacks that are being addressed. Improvements made to EURO-GANEX system have resulted in the emergence of the so-called New EURO-GANEX process, where the composition of the solvent has been modified by replacing TODGA and DMDOHEMA with cis-mTDDGA in the organic phase and SO3-Ph-BTP with PyTri-Diol in the aqueous phase in order to resolve important issues. The objective of this work is twofold: evaluate the gamma radiolysis resistance of the new EURO-GANEX process by dynamic irradiation conditions simulating the three main steps of the process and validate the design of CIEMAT Náyade, CEA Marcel and INL irradiation loop devices since each of them mimics different aspects of the real process. Náyade and the INL loops could irradiate together the organic and aqueous phases, whereas in the CEA loop, the irradiated solvent is recycled continuously inside a platform with several stages of mixer-settlers containing aqueous flows simulating the three main steps of the process. The extraction performances and changes in the composition of the solvent have been analysed during the irradiation experiment by different techniques: gamma spectrometry and ICP-MS/OES for cations or tracers extraction, and HPLC-MS to quantify the degradation and evaluate the degradation compounds. Additionally, in spite of some differences between the three-irradiation facilities, this inter-institutional study shows that they are three comparative tools, which provide results on the stability towards radiolysis of a liquid-liquid extraction system. Favourable extraction results for the different steps are obtained according to the static irradiation studies found in literature. However, the degradation of cis-mTDDGA is appreciable leading to degradation compounds, some of which form precipitates and produce important changes in viscosity, important aspects that must be addressed prior to the successful industrial application of the new EURO-GANEX process.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Examination of Gamma-irradiated Calcium Silicate Hydrates. Part II: Mechanical Properties

Mechanical properties of calcium silicate hydrates (C-S-H) with C/S ratios of 0.75, 1, and 1.33 were examined with nanoindentation after gamma-adsorbed doses of 0.145, 0.280, 0.500, and 0.784 MGy, and were compared with control samples. Young' s modulus and stress relaxation tests showed no apparent trend with irradiation dose. Qualitatively, most of the irradiated samples were found to relax more than their respective controls, but not always in a statistically significant manner. Most of the Young' s modulus irradiated–control pairs showed marginally higher stiffness in the irradiated samples, but overall trends with irradiation dose were not obvious. Creep compliance was obtained for the samples irradiated at the highest dose and their respective controls. Two of the three irradiated samples exhibited less creep than their respective controls, but only one of which was statistically significant. The lack of clear changes in mechanical properties for these samples correlates with separate chemical analyses that showed no loss of interlayer water by exposure to irradiation or changes in the mean silicate chain length. Further research evaluating higher doses (25 and 200 MGy) representative of those received by concrete structures in nuclear power plants at prolonged operation is being carried out to complement the present study.

36 MATERIALS SCIENCE↗

Investigation of $\gamma$-irradiated polyvinylidene fluoride and its acute toxicity

Polyvinylidene fluoride (PVDF) is an industrial polymer with many applications. The EPR spectrum of polyvinylidene fluoride (PVDF) irradiated by γ-rays at low temperature (-196°C) provides evidence for the presence of radicals formed by loss of a fluorine atom from PVDF, and polyenyl radicals. Polyenyl radicals dominate in the polymer irradiated at room temperature. Peroxide radicals formed as a result of oxidation of primary free radicals by atmospheric oxygen are observed in the EPR spectrum of a polymer irradiated in air, or when air is introduced to a polymer irradiated in a vacuum. Carbonization and oxidation of γ-irradiated PVDF were observed in the XPS spectrum. A model for polymer degradation based on correlated G3(MP2) molecular orbital theory electronic structure calculations is described in this report. The presence of HF in the products and a low yield of products with the composition CxFyHz indicate the absence of noticeable thermal destruction of the main polymer chain upon heating to 220°C of both the initial and γ-irradiated PVDF. Although any HF released during the PVDF dehydrofluorination would be toxic, a study of the acute toxicity of γ-irradiated PVDF showed that the polymer does not exhibit toxicity after a single intragastric route of administration to mice.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Evaluation of Radiolysis Data for Hydrogen Gas Generation During Gamma Irradiation of Pre-Corroded and Pristine Aluminum Samples - An Aluminum SNF Dry Storage Study Interim Report

Information and data from radiolysis testing to measure hydrogen (H₂) generated from hydrated oxides on aluminum exposed to ⁶⁰Co radiation were reviewed to evaluate hydrogen generation with radiation dose. Hydrogen generation rate is a primary input to the coupled thermal-chemical system model of the performance of aluminum-clad spent nuclear fuel in a dry storage canister (ASNF-in-canister) over its storage life. Hydrogen generation data and hydrogen generation rate (hydrogen generated per absorbed radiation dose, $\Delta$(H₂)/$\Delta$(dose)) were reported in two reports for radiolysis testing [1, 2]. The testing involved first-time data¹ for radiolytic yield of hydrogen from hydrated-oxides-on-aluminum substrates, and from pristine (non-corroded) aluminum substrates. The laboratory-grown hydrated-oxides-on-aluminum substrates contained a ~ 5 μm film consisting of the trihydroxide bayerite (Al(OH)₃) with amounts of the oxyhydroxide boehmite (AlOOH). These specimens were assumed to also contain physisorbed water (unquantified). The pristine substrates were flat coupons of aluminum that were ground to a 600-grit finish but were not immersed in water to grow a hydrated oxide. The pristine substrates were tested as companion specimens to provide information on hydrogen generation from material without chemisorbed water and were assumed to contain physisorbed water (unquantified).

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Lithium indium diselenide — An advanced material for neutron detection

Here, this paper describes the synthesis, crystal growth, detector fabrication, radiation hardening studies, MCNP modeling, and characterization of lithium indium diselenide or LiInSe 2 . This newly-developed room temperature thermal neutron detector has semiconducting and scintillating properties and it is suitable for neutron detection application. LiInSe 2 was synthesized starting from elemental Li, In, Se in two steps due to high reactivity of Li. A single crystal of LiInSe 2 was grown using the Vertical Bridgman method. The room temperature band gap was found to be 2.8 eV using optical absorption measurements. Bulk resistivity was measured at ∼5 × 10 11 Ωcm. Photoconductivity measurements of LiInSe 2 wafers identified a peak in the photocurrent around 445 nm. Nuclear radiation detectors were fabricated from single crystal wafer and the responses to alpha particles at various biases were measured. The mobility-lifetime product was estimated. Gamma irradiation studies were performed with calculated absorbed doses ranging from 0.2126 to 21,262 Gy. The characterization of the two wafers for their scintillator performance was conducted after each irradiation. The gamma irradiation produced a reduction of the light yield that translated to a lower channel number for the centroid of alpha detection spectra. It also showed a considerable reduction of the decay time after the first irradiation. These are the first studies on gamma radiation hardening with this material.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Influence of metal ion complexation on the radiolytic longevity of butyramide extractants under direct dissolution conditions

The direct dissolution of volox-treated used nuclear fuel (UNF) into an organic solution—comprised of diluent and specialized extractants—poses a promising alternative to the traditional liquid-liquid solvent extraction approach to reprocessing UNF. However, moving to direct dissolution removes the presence of a concentrated nitric acid aqueous phase, which has been shown to significantly influence the radiolytic longevity of extractants in liquid-liquid solvent extraction flowsheets. With this in mind, and given the limited knowledge of radiation effects under direct dissolution conditions, we present a time-resolved and dose accumulation study on the impact of direct dissolution conditions on the radiolytic longevity of two candidate butyramide extractants—N,N-di-(2-ethylhexyl) butyramide (DEHBA) and N,N-di-(2-ethylhexyl)isobutyramide (DEHiBA)—in pre-equilibrated n-dodecane solvent in the presence and absence of process relevant metal ions, uranium and rhenium. Rhenium, and by extension technetium, extraction had little impact (=10%) on the overall radiolytic stability of these ligands, despite observed increases in chemical kinetic reactivity (>2×) of the corresponding complexes with the n-dodecane radical cation. Uranium-loading on the other hand, significantly improved the lifetime of both ligands (>30%) under gamma irradiation, with a greater stabilization observed for DEHBA over DEHiBA. This draft manuscript has been prepared in fulfillment of NTRD-MRWFD-2024 M3FT-24IN030101115.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Initial Radiation of Nuclear Power Plant Components (CRADA NFE-17-06892 Final Report)

As advanced reactor concepts are developed, instrumentation and control (I&C) cables and sensors will play a significant role in assuring their operation is well understood, and safety-related process information can be collected reliably over the course of their design life. As a first step to develop acceptance criteria and define the limits of currently available nuclear I&C cable and sensor technology, a collaboration between Oak Ridge National Laboratory and Analysis and Measurement Services (AMS) Corporation was initiated under the Gateway for Accelerated Innovation in Nuclear (GAIN) Program. This collaborative effort focused on providing technical information related to the combined thermal and radiation aging of nuclear power plant (NPP) cables, companion NPP cable insulation samples, and two prototype temperature and two pressure sensors. This work was carried out using the instrumented gamma irradiation system that is part of the High Flux Isotope Reactor Gamma Irradiation Facility (HFIR GIF). This system provided the opportunity for the condition of the prototype sensors to be tracked during the irradiation in order to determine their functional limits with greater granularity when compared to a simple irradiation insertion at the HFIRGIF. Experimental assemblies for sensor irradiation were successfully completed, and sensor irradiation to accumulated dose of 3000 kGy at a rate of approximately 6 kGy/hr was carried out. Two prototype temperature assemblies were continuously monitored during exposure to 3000 kGy and found to have no observable degradation from the irradiation. Two pressure sensors were first irradiated to 1000 kGy. At which point, one of the sensors was found to operate normally; however, a second sensor failed the operability test. The operable pressure sensor output was then continuously monitored from 1000 kGy to 3000 kGy. The sensor output exhibited steady trendable drift versus the accumulated radiation dose. At the conclusion of the irradiation, the four test sensors were released for laboratory calibration checks. The calibration check of the temperature sensors found no significant change after irradiation compared to baseline measurements before irradiation; however, both pressure sensors failed the post-test calibration checks and both sensors were not functioning properly. These results will help inform sensor manufacturers and I&C design engineers of potential limitations of these types of sensors for use in high radiation environments associated with the compact configuration of new Small Modular Reactors (SMRs) such as NuScale’s reactor power module. Irradiations on NPP cables and companion samples were not completed due to programmatic delays, however a path forward that was based on lessons learned from this project was defined to accommodate future, comparable instrumented gamma irradiations at the HFIR GIF.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Sterilization of N95 Respirators via Gamma Radiation: Comparison of Post-sterilization Efficacy

This study evaluated gamma irradiation for sterilization and reuse of two models of N95 respirators after gamma radiation sterilization as a method to increase availability of N95 respirators during a shortage. The Sandia National Laboratories Gamma Irradiation Facility was used to irradiate two different models of N95 filtering facepiece respirators at doses ranging from 0 kGy(tissue) to 50 kGy(tissue). The following tests were used to determine the efficacy of the respirator after irradiation sterilization: Ambient Aerosol Condensation Nuclei Counter Quantitative Fit Test, tensile test, strain cycling, oscillatory dynamic mechanical analysis, microscopic image analysis of fiber layers, and electrostatic field measurements. Both of the respirator models exhibited statistically significant changes after gamma irradiation as shown by the Quantitative Fit Test, electrostatic testing and the aerosol testing. The change in electrostatic capability of the filter reduced the efficiency of challenging particles near the 200 nm size by approximately 40-50%. Both tested respirators showed statistically significant changes associated with gamma sterilization. However, our results indicate that choices in materials and manufacturing methods to achieve N95 filtration lead to different magnitudes of damage when exposed to gamma radiation at sterilization relevant doses. This damage results in lower filtration performance. While our sample size (2 different types of respirators) was small, we did observe a change in electrostatic properties on a filter layer that coincided with the failure on the Quantitative Fit Test.

61 RADIATION PROTECTION AND DOSIMETRY↗

Sterilization of N95 Respirators via Gamma Radiation: Comparison of Post-sterilization Efficacy

This study evaluated gamma irradiation for sterilization and reuse of two models of N95 respirators after gamma radiation sterilization as a method to increase availability of N95 respirators during a shortage. The Sandia National Laboratories Gamma Irradiation Facility was used to irradiate two different models of N95 filtering facepiece respirators at doses ranging from 0 kGy(tissue) to 50 kGy(tissue). The following tests were used to determine the efficacy of the respirator after irradiation sterilization: Ambient Aerosol Condensation Nuclei Counter Quantitative Fit Test, tensile test, strain cycling, oscillatory dynamic mechanical analysis, microscopic image analysis of fiber layers, and electrostatic field measurements. Both of the respirator models exhibited statistically significant changes after gamma irradiation as shown by the Quantitative Fit Test, electrostatic testing and the aerosol testing. The change in electrostatic charge of the filter was correlated with a reduction in capturing particles near the 200 nm size by approximately 40-50%. Both tested respirators showed statistically significant changes associated with gamma sterilization. However, our results indicate that choices in materials and manufacturing methods to achieve N95 filtration lead to different magnitudes of damage when exposed to gamma radiation at sterilization relevant doses. This damage results in lower filtration performance. While our sample size (2 different types of respirators) was small, we did observe a change in electrostatic properties on a filter layer that coincided with the failure on the Quantitative Fit Test and reduction in aerosol filtering efficiency. Key Words: N95 respirators, respirators, airborne transmission, pandemic prevention, COVID-19, gamma sterilization

COVID-19↗

Radiolytic evaluation of a new technetium redox control reagent for advanced used nuclear fuel separations

Technetium is a problematic radioisotope for used nuclear fuel (UNF) and subsequent waste management owing to its high environmental mobility and coextraction in reprocessing technologies as the pertechnetate anion (TcO 4 - ). Consequently, several strategies are under development to control the transport of this radioisotope. A proposed approach is to use diaminoguanidine (DAG) for TcO 4 - and transuranic ion redox control. Although the initial DAG molecule is ultimately consumed in the redox process, its susceptibility to radiolysis is currently unknown under envisioned UNF reprocessing conditions, which is a critical knowledge gap for evaluating its overall suitability for this role. To this end, we report the impacts of steady-state gamma irradiation on the rate of DAG radiolysis in water, aqueous 2.0 M nitric acid (HNO 3 ), and in a biphasic solvent system composed of aqueous 2.0 M HNO 3 in contact with 1.5 M N,N-di-(2-ethylhexyl)isobutyramide (DEHiBA) dissolved in n-dodecane. In addition, we report chemical kinetics for the reaction of DAG with key transients arising from electron pulse radiolysis, specifically the hydrated electron (e aq - ), hydrogen atom (H ˙ ), and hydroxyl ( ˙ OH) and nitrate (NO 3 ˙ ) radicals. The DAG molecule exhibited significant reactivity with the ˙ OH and NO 3 ˙ radicals, indicating that oxidation would be the predominant degradation pathway in radiation environments. This is consistent with its role as a reducing agent. Steady-state gamma irradiations demonstrated that DAG is readily degraded within a few hundred kilogray, the rate of which was found to increase upon going from water to HNO 3 containing solutions and solvents systems. This was attributed to a thermal reaction between DAG and the predominant HNO 3 radiolysis product, nitrous acid (HNO 2 ), k(DAG + HNO 2 ) = 5480 ± 85 M -1 s -1 . Although no evidence was found for the radiolysis of DAG altering the radiation chemistry of the contacted DEHiBA/n-dodecane phase in the investigated biphasic system, the utility of DAG as a redox control reagent will likely be limited by significant competition with its degradation by HNO 2 .

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗