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At least 613 records · Page 34

Processing iridium foil targets for the production of radioplatinum isotopes

Dissolving noble metals remains challenging because of their high reduction potentials and tendency to passivate as oxides. This barrier is acute for iridium metal targets used to produce platinum radioisotopes, where post-irradiation processing is costly and hazardous with conventional routes. Here, we report a rapid alkali-fusion method that converts Ir metal directly into chloride media compatible with downstream separations. In the process, an Ir foil is fused with Na 2 O 2 and KCl at temperatures above the KCl melting point; the molten salt enhances heat transfer and facilitates O 2− /Na + transport to the Ir surface, overcoming Ir's oxidation resistance. The fused cake is then cooled and leached in 10 M HCl, yielding Ir in solution without specialized microwave digestion or HF. Pt separation from bulk Ir was evaluated using extraction-chromatography resins. A newly formulated ionic-liquid resin containing Cyphos IL 104 (E104) was compared with two commercial resins (TrisKem TK200 and TK201). All resins achieved >90 % Ir recovery. Pt co-recovery depended on resin chemistry: TK200 recovered 85.9 % Pt, TK201 recovered 54.1 % Pt, and E104 recovered 72.9 % Pt. This study introduces a fast, scalable dissolution route for solid Ir targets and demonstrates resin choices that tune Ir/Pt separation, lowering the barrier to routine production and purification of Pt radioisotopes.

07 ISOTOPE AND RADIATION SOURCES↗

Isotope harvesting from a NSCL beam blocker: Isolation of 173,174 lu

A decommissioned beam blocker from the National Superconducting Cyclotron Laboratory was processed for recovery of long-lived Lu radioisotopes at Lawrence Livermore National Laboratory. The beam blocker was a W alloy that had built up considerable levels of activation from high energy particle irradiation over the course of 20 years. The isolation of 173,174 Lu from the bulk material required a complex chemical separation scheme due to the large mass and the wide variety of elements present. The final yield of 173,174 Lu was 76.7 ± 4.8% with an extremely high radiopurity and chemical purity. The reduction in W mass from the initial material to the final sample was 8 × 10 10 . In conclusion, the results show great promise for extracting large quantities of Lu radioisotopes from these types of W beam blockers at the Facility for Rare Isotope Beams, which will be irradiated at levels significantly higher than the beam blocker investigated here.

and nuclear chemistry↗

Controls on the 36 Cl/Cl input ratio of paleo-groundwater in arid environments: New evidence from 81 Kr/Kr data

In this study, measurements of the long-lived 81 Kr and 36 Cl radioisotopes in groundwater from the Negev Desert (Israel) were used to assess the 36 Cl/Cl input ratios and Cl – contents for paleorecharge into the Nubian Sandstone Aquifer (NSA). The reconstructed Cl – content of the recharge flux was on the order of 300–400 mg/L. An initial 36 Cl/Cl ratio of 50 × 10 –15 was assessed for the groundwater replenishment in the Negev Desert since the late Pleistocene, in agreement with the 36 Cl/Cl ratios in recent local rainwater. This is despite possible changes in the climatic conditions and the 36 Cl production rates in the atmosphere over this timeframe. This similarity in values is explained by the major role played by the erosion and weathering of near-surface materials in the desert environment that dominate the hydrochemistry of rains, floods, and the consequent groundwater recharge. Spatial variation in the reconstructed initial 36 Cl/Cl ratio is accounted for by the differences in the mineral aerosol sources for specific recharge areas of the NSA. Accordingly, regional variations in the 36 Cl/Cl input in groundwater reservoirs surrounding the Mediterranean Sea indicate various processes that govern the 36 Cl/Cl system. Finally, the results of this study highlight the great advantage of integrating 81 Kr age information in evaluating the initial 36 Cl/Cl and Cl – input, which is essential for the calibration of 36 Cl radioisotope as an available long-term dating tool for a given basin.

54 ENVIRONMENTAL SCIENCES↗

Simultaneous Separation of Protactinium-230 and Uranium-230 Isotopes from a Proton-Irradiated Thorium Matrix

The use of radioisotopes in cancer therapy is becoming increasingly important. As a potential candidate for targeted alpha therapy, 230 U (t 1/2 = 20.8 d), the decay daughter of 230 Pa (t 1/2 = 17.4 d) is currently being investigated for cancer treatment. For radioisotopes to be used in biomedicine, they must be radiochemically pure and free from carrier interference. This can be challenging given their short half-life. Thus, in this study, radiological separation methods for harvesting isotopes for use in biomedicine must be simple, fast, and capable of obtaining the required purity levels. Herein, we demonstrate a new rapid method for the separation of 230 U and 230 Pa from a proton-irradiated 232 Th metal target and from coproduced fission products with high recovery and purity. A dual-column approach was used, in which the first column was an anion exchange (AX) column and the second column was a DGA column. The MP-1 AX resin was used for the primary separation of the three major components (U, Pa, and Th, as well as fission products), and the DGA column was used for further purification of the separated 230 U. The method is simple and allows for straightforward separation of U, Pa, Th, and fission products; also, the entire process from target dissolution to shipment of 230 U could be completed in 3 days with an overall 230 U recovery yield of 96 ± 3%. Additionally, separated 230 Pa fractions can be used to harvest ingrown 230 U after initial separation.

DGA resin↗

Harvesting krypton isotopes from the off-gas of an irradiated water target to generate 76Br and 77Br

Abstract A flowing-water target was irradiated with a 150 MeV/nucleon beam of 78 Kr at the National Superconducting Cyclotron Laboratory to produce 77 Kr and 76 Kr. Real-time gamma-imaging measurements revealed the mass transport of the krypton radioisotopes through the target-water processing, or “isotope harvesting”, system. The production rates were determined to be 2.7(1) × 10 –4 nuclei of 76 Kr and 1.18(6) × 10 –2 nuclei of 77 Kr formed per incident 78 Kr ion. Utilizing an off-gas processing line as part of the isotope harvesting system, a total of 7.2(1) MBq of 76 Kr and 19.1(6) MBq of 77 Kr were collected in cold traps. Through the decay, the daughter radionuclides 76 Br and 77 Br were generated and removed from the traps with an average efficiency of 77 ± 12%. Due to the differences in half-lives of 76 Kr and 77 Kr, it was possible to isolate a pure sample of 76 Br with 99.9% radionuclidic purity. The successful collection of krypton radioisotopes to generate 76 Br and 77 Br demonstrates the feasibility of gas-phase isotope harvesting from irradiated accelerator cooling-water. Larger-scale collections are planned for collecting by-product radionuclides from the Facility for Rare Isotope Beams.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

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↗

Detection and quantification of trace technetium in the presence of molybdenum using laser-induced breakdown spectroscopy

Technetium (Tc) is a very important element that is encountered in many aspects, from its presence in radioactive waste and its potential environmental impact to its use as a medical radioisotope. Its detection and quantification in liquid samples is traditionally cumbersome, involving detailed sample preparation and analysis by mass spectrometry or scintillation. This article demonstrates the first comprehensive emission spectral analysis of Tc from a liquid sample by immobilization in a polymer and analysis by laser-induced breakdown spectroscopy (LIBS). A survey of LIBS spectra was completed to identify the strongest analytical lines for quantification of trace Tc in the presence of Mo. The quantification of Tc in a Mo-containing matrix was selected because Tc radioisotopes are the daughter products of Mo isotope decay. The first reported calibration curves by LIBS are provided with limits of detection and quantification down to 0.710 µg mL −1 and 1.39 µg mL −1 , respectively. Ultimately, this study demonstrated the feasibility of trace Tc quantification using LIBS and will serve as a reference for future research related to monitoring this radioactive species.

Andrews, Hunter B. [Oak Ridge National Laboratory ↗

Thermal decomposition of neptunyl ammonium nitrate: mechanistic insights and structural characterization of the Np 2 O 5 intermediate phase

Neptunium (Np) possesses a rich and unique chemistry that often diverges from other actinide elements yet remains relatively underexplored compared with the other light actinides. A resurgence of interest in Np has been spurred by the application of 237 Np for plutonium-238 ( 238 Pu) production for use in radioisotope thermoelectric generators (RTGs), necessitating evaluation of Np chemical reactions and materials. The work presented here studied the thermal decomposition of neptunyl ammonium nitrate (NH 4 Np VI O 2 (NO 3 ) 3 ) for synthesis of neptunium dioxide (NpO 2 ), which is the target material used for production of 238 Pu. Additionally, structural characterization of the intermediate solid Np pentoxide (Np 2 O 5 ) was performed. Advanced solid-state characterization techniques, including simultaneous thermal analysis (STA), powder X-ray diffraction (pXRD), Raman spectroscopy, and density functional theory (DFT) modeling have been combined to study the reaction pathways. Analysis revealed that NH 4 Np VI O 2 (NO 3 ) 3 thermally decomposes to a proposed neptunyl nitrate intermediate, followed by Np 2 O 5 and finally NpO 2 , all within the temperature range of 150 °C–600 °C. Further characterization of the pentoxide intermediate provided the first Raman spectra of pure-phase Np 2 O 5 and associated DFT modeling confirmed Raman peak assignments for this phase. These findings provide mechanistic information to advance production of the critical radioisotope 238 Pu and advance the state of knowledge on Np materials chemistry using modern characterization techniques.

Lawson, Kathryn M. [Oak Ridge National Laboratory ↗

The scoping, design, and plasma physics optimization of the Eos neutron source stellarator

On the path to a fusion pilot plant, Thea Energy plans to build Eos, a sub-breakeven, deuterium-deuterium, beam-target fusion, stellarator neutron source facility for producing tritium and other valuable radioisotopes. In this paper, a set of 1D plasma physics models are coupled and used to design the operating point of the facility and predict performance. At this foundational stage of the design, analytic and approximate models are sufficient to capture the leading-order effects, and fast enough to run in the inner loop of an optimizer. Higher-fidelity analyses will follow. Models of 1D profile-dependent neutral beam stopping, ion beam slowing down, beam-target fusion, electron-ion classical heat transfer, energy confinement (ISS04), beam pressure, beam heating of ions and electrons, beam-beam fusion fraction, and neutral beam injection and gyrotron heating electrical efficiencies are included. A numerical optimizer is used to determine the minimum required facility electric power to generate tritium at a given rate. A potentially advantageous regime is described in which modern precisely-quasisymmetric stellarators, new high-temperature superconductors, ITER-derived neutral beam injection, and new high-frequency gyrotrons enable a suitible target plasma with hot electrons, cold ions, peaked density and temperature profiles, and high beam-injected ion density. It appears possible at this time for a facility with a medium-scale and medium-strength stellarator whose required facility electric power is less than 40 MW to produce $2.5\times 10^{17}$ neutrons s -1 for the production of radioisotopes. With the addition of a tritium breeding blanket, such a facility could produce 0.2 grams d -1 or 70 grams yr -1 of tritium.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Methods to Account for CAP-88 PC-Omitted Nuclides in Radioactive Air Emissions From DOE Facilities

Routine research and development activities at US Department of Energy facilities can result in the release of radioactive emissions, potentially exposing the public and the environment. Such emissions are subject to certain Clean Air Act regulations, specifically those outlined in 40 CFR Part 61, Subpart H, which are enforced by the US Environmental Protection Agency. Compliance is determined in part with a dose standard. Doses to members of the public from most US Department of Energy facility airborne emissions are modeled using CAP-88 PC software, the latest version of which contains more than 1,200 radionuclides. Radioactive effluents from US Department of Energy facilities may contain radioisotopes that are not available in CAP-88 PC, and those radioisotopes must still be evaluated to determine whether doses to members of the public are below established limits. CAP-88 PC-omitted radionuclides can be accounted for using alternate methods, including the use of surrogate radionuclides. Here, this paper elucidates the process of accounting for CAP-88 PC-omitted radionuclides by providing criteria and rationale for surrogate selection, a compilation of surrogate radionuclides used in the past by US Department of Energy facilities, a discussion of alternate methods used to account for CAP-88 PC-omitted radionuclides and a comparison of methods and impacts on receptor doses. Ultimately, this paper aims to aid in the process of surrogate selection and consequently to simplify and expedite compliance with Clean Air Act regulations.

61 RADIATION PROTECTION AND DOSIMETRY↗

Methods to Track Effective Doses from Airborne Radioactive Emissions for Compliance with 40 CFR 61, SUBPART H

US Department of Energy national laboratories can play an integral role in not only the advancement of science but also in the treatment of various medical conditions through research and development activities conducted at radioisotope production facilities. Here, a project has been underway at Oak Ridge National Laboratory since 2016 whose mission is to produce and supply the radioisotope 227 Ac, which is used in a radiopharmaceutical developed to treat certain types of prostate cancer and bone metastases. Production activities result in the environmental release of airborne radioactive emissions, which are governed by Clean Air Act regulations described in 40 CFR Part 61, Subpart H. Stack 3039, the source that emits radioactive effluents from 227 Ac production, is subject to additional requirements outlined in American National Standards Institute (ANSI) N13.1-1969 due to its grandfathered status. Radioactive emissions are limited to levels below those that would cause annual compliance dose standards for members of the public to be exceeded and stack 3039 to lose its grandfathered status. To allow for maximum production of 227 Ac without exceeding relevant dose limits, monthly tracking of project emissions and resulting CAP88-PC modeled effective doses to a maximally exposed individual have been implemented. Four years of tracking data were compiled and analyzed to identify additional methods that could be used to estimate project doses more frequently, potentially further optimizing 227 Ac production while maintaining compliance with applicable regulations.

atmospheric emissions↗

Photonuclear cross sections for the 197 Au ⁢(𝛾, 𝑝⁢𝑛)⁢ 195⁢𝑚 Pt reaction near threshold

Platinum radioisotopes are of growing interest for targeted cancer therapy and diagnostic imaging because their decay delivers highly localized radiation doses in tissue, herewith enabling precise DNA damage through Auger-electron emission. Developing production technologies that provide platinum isotopes with high specific activity is essential in radioisotope therapy. Photonuclear reactions on stable nuclei offer a viable accelerator-based route for isotope production when supported by reliable cross-section data. We report photonuclear cross-section measurements for the 197 Au(γ, pn) 195m Pt reaction at incident γ-ray energies of 27, 29, and 31 MeV using the activation method. The measurements were performed by irradiating a stack of concentric-ring gold targets with a quasi-monoenergetic γ-ray beam provided by the High Intensity Gamma-ray Source (HI γS). The induced 195m Pt activity was quantified using off-line γ-ray spectroscopy. These data provide the first experimental constraints on the 197 Au(γ, pn) 195m Pt cross section in the near-threshold region. Furthermore, the comparison of the measured excitation function to PHITS and TALYS calculations indicates that the reaction becomes measurable only near 30 MeV and that substantially higher bremsstrahlung end-point energies are required for practically meaningful production.

190 ≤ A ≤ 219↗

Sustainable production of radionuclidically pure antimony-119

Background Radiopharmaceutical therapy (RPT) uses radionuclides that decay via one of three therapeutically relevant decay modes (alpha, beta, and internal conversion (IC) / Auger electron (AE) emission) to deliver short range, highly damaging radiation inside of diseased cells, maintaining localized dose distribution and sparing healthy cells. Antimony-119 ( 119 Sb, t 1/2 = 38.19 h, EC = 100%) is one such IC/AE emitting radionuclide, previously limited to in silico computational investigation due to barriers in production, chemical separation, and chelation. A theranostic (therapeutic/diagnostic) pair can be formed with 119 Sb’s radioisotopic imaging analogue 117 Sb (t 1/2 = 2.80 h, E γ = 158.6 keV, I γ = 85.9%, β + = 262.4 keV, I β+ = 1.81%). Results Within, we report techniques for sustainable and cost-effective production of pre-clinical quality and quantity, radionuclidically pure 119 Sb and 117 Sb, novel low energy photon measurement techniques for 119 Sb activity determination, and physical yields for various tin target isotopic enrichments and thicknesses using (p, n) and (d, n) nuclear reactions. Additionally, we present a two-column separation providing a radioantimony yield of 73.1% ± 6.9% (N = 3) and tin separation factor of (6.8 ± 5.5) x 10 5 (N = 3). Apparent molar activity measurements for deuteron produced 117 Sb using the chelator TREN-CAM were measured at 42.4 ± 25 MBq 117 Sb/µmol (1.14 ± 0.68 mCi/µmol), and we recovered enriched 119 Sn target material at a recycling efficiency of 80.2% ± 5.5% (N = 6) with losses of 11.6 mg ± 0.8 mg (N = 6) per production. Conclusion We report significant steps in overcoming barriers in 119 Sb production, chemical isolation and purification, enriched target material recycling, and chelation, helping promote accessibility and application of this promising therapeutic radionuclide. We describe a method for 119 Sb activity measurement using its low energy gamma (23.87 keV), negating the need for attenuation correction. Finally, we report the largest yet-measured 119 Sb production yields using proton and deuteron irradiation of natural and enriched targets and radioisotopic purity > 99.8% at end of purification.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Establishing Pb-203 production from electrodeposited Tl targets at Brookhaven National Laboratory

Background: Promising developments in Pb-212 radiopharmaceutical therapies have increased demand for Pb-203 diagnostic agents. Building on previous work from various isotope production facilities, this study optimized Pb-203 production from electrodeposited Tl targets at Brookhaven National Laboratory (BNL). The additional supply of Pb-203 may help meet growing preclinical and clinical demands. Results: Two Tl targets were irradiated at the Brookhaven Linac Isotope Producer facility with 30 ± 1 MeV protons, measured using previously published cross section data. Distribution coefficients for Pb Resin in acetate media were investigated for both Na + and K + cations, where potassium acetate was ~ 4 times more effective at stripping Pb from the Pb Resin. The Tl electrodeposition was optimized to deposit 350 mg of Tl (~ 60 mg/cm 2 ) on Au backing in under 6 h. The proposed separation process was completed in < 1.5 h and achieved > 98% and 92 ± 3% recovery of Tl and Pb, respectively, with an overall Tl-Pb separation factor of 6 × 10 5 . The experimentally measured half-life of Pb-203 was 52.4 ± 0.7 h, agreeing with 51.93 ± 0.02 h reported by the National Nuclear Data Center. The radioisotopic purity of the Pb fraction at 24 h post end of bombardment (EOB) from a 24 h irradiation was 66% Pb-203, 28% Pb-201, and 6% Pb-200. Following chemical separation, the Pb-203 produced in this work (21 MBq Pb-203 EOB) achieved apparent molar activities of 10 ± 5 and 0.9 ± 0.5 GBq/µmol for [ 203 Pb]Pb-DOTAM and [ 203 Pb]Pb-DO3A, respectively, decay corrected to EOB. Data derived from this work suggests BNL can produce > 10’s GBq Pb-203 with > 99% radiochemical and radioisotopic purity from Tl-205 for worldwide distribution. Conclusions: The production and separation of Pb-203 from natural Tl target material was successfully demonstrated at BNL. Existing methods were adapted and optimized for the facilities at BNL. Results from this work will guide future large-scale Pb-203 production opportunities at BNL for clinical applications.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Search for New Isotope Production Pathways

The isotope group at Jefferson Lab is carrying out R&D for producing medically interesting radioisotopes, especially those with theranostic (therapeutic and diagnostic) attributes. Here the search for viable production mechanisms has been expanded to multi-step reactions where a daughter is produced from the target and decays into a medically interesting granddaughter radioisotope. It is difficult to find efficient production routes when investigating both the initial excitation reaction as well as the decay routes leading to medically interesting isotopes. The overall goal of this project is to create a structured code in Python to find these decay routes by automatically exploring the large number of isotopes and their possible decay modes. The program structure is described, and preliminary results are presented.

Dabill, L.↗

Recovery and Recycle of Irradiated Low-Enriched Uranium from the Production of 99 Mo

Technetium-99m ( 99m Tc), the daughter (decay product) of molybdenum-99 ( 99 Mo), is the most commonly used medical radioisotope in the world. 99 Mo is produced primarily from the fissioning of 235 U. A potential future producer, Eden Radioisotopes, is planning to irradiate low-enriched-uranium metal-foil targets for production of 99 Mo. Their plan is to recover, purify, and recycle the irradiated LEU in a new set of targets inside their facility. This study assesses processes to (1) purify and recycle uranium following 99 Mo recovery and (2) convert the uranium to metal for subsequent foil production and target fabrication. A UREX (uranium recovery by extraction) liquid–liquid extraction flowsheet that utilizes centrifugal contactors was developed using the Argonne model for universal solvent extraction (AMUSE) to recover and purify the uranium. The calculated flowsheet predicted high decontamination from Pu and most fission products and >99.99% recovery of uranium. Suggestions for conversion of the UREX U-product (uranyl nitrate in dilute nitric acid) to U metal were provided based on literature studies and earlier laboratory studies performed at Argonne National Laboratory.

07 ISOTOPE AND RADIATION SOURCES↗

Radiological Consequence Evaluation for Dragonfly Mission

The scoping calculations presented in this report were developed to provide a conservative estimate of radiological dose consequences associated with launch accidents involving one multi-mission radioisotope thermoelectric generator (MMRTG) and possibly up to 43 light-weight radioisotope heater units (LWRHUs) for the upcoming National Aeronautics and Space Administration (NASA) Dragonfly mission. This report summarizes dose consequence analysis from SAND2019-11148, “Nuclear Risk Assessment 2019 Update for the Mars 2020 Mission Environmental Impact Statement,” referred to in this document as the Mars 2020 nuclear risk analysis (NRA) and compares those results to a parametric analysis used to analyze potential dose consequences for the Dragonfly mission.

07 ISOTOPE AND RADIATION SOURCES↗

Testing of Helium Cooled Metal Molybdenum Disk Target System

NorthStar Medical Radioisotopes LLC is planning to produce the important medical radioisotope molybdenum-99 (Mo-99) through a photonuclear reaction on molybdenum-100 (Mo-100). In this approach, multiple thin disks of enriched molybdenum metal will be bombarded with a 40-MeV electron beam. Because enriched Mo-100 is expensive, we intend to use as much beam power as possible to achieve maximum production yield and minimize the size of the target. This requirement leads to very high beam power density (heat deposition in the target), which sets challenging requirements for cooling. Together with scientists at Los Alamos National Laboratory, a team at Argonne National Laboratory has developed and demonstrated a cooling approach using pressurized helium, which allows for efficient heat removal. One of the challenges in this approach is the management of the heat load on the target window. The target window separates the high-pressure helium atmosphere inside the target from the vacuum in the beamline, so it is constantly under stress from differential pressure. Also, the window is cooled only by the helium gas flow from one side, making the window design challenging. High heat deposition in the target disks also imposes a strict requirement on performance of the helium cooling system and thickness of the target disks. The target disks are produced from metal powder via a press-and-sinter process. The resulting disks do not possess as high a tensile strength as solid molybdenum and might not survive the vibration from the high-velocity helium coolant and the high thermal stress from beam heating.

42 ENGINEERING↗