Neutron Flux Research Articles

Numerical approximation of a two-dimensional fractional neutron diffusion model describing dynamics of neutron flux in a nuclear reactor

The goal of this work is to develop a high-order numerical technique to solve a two-dimensional fractional neutron diffusion (FND) model describing dynamical behavior of a lead-cooled fast reactor. This method is based on L1 technique for temporal discretization and a high-order compact Alternating Direction Implicit (ADI) finite difference scheme for the spatial discretization. Numerical simulations are conducted to demonstrate the applicability of the method. We calculate the neutronic flux in the core and examine how the order of the temporal-fractional derivative influences the neutronic flux. The results obtained with FND model for different values of anomalous diffusion coefficient are compared against those obtained with P1 model and classical neutron diffusion equation. In addition, the computational time (CPU time) is provided in order to justify the computational efficiency of the proposed method.

Thermoplastic and thermoset polymer matrix composites reinforced with bismuth oxide as radiation shielding materials

Polymer composites reinforced with heavy metals have wide applications in the nuclear industry as radiation shielding materials against emitted radiation from nuclear equipment. In this study, considering bismuth oxide as reinforcement, various polymers including thermoplastic polymers, such as high-density polyethylene, low-density polyethylene, polypropylene, polyvinyl acetate, polymethyl methacrylate, and thermosetting epoxy resin have been considered as matrices. After preparing various samples of polymer composites loaded with 60 % by weight of bismuth oxide, some physical, mechanical, and radiation shielding properties (neutron and gamma attenuation coefficients) of these polymer composites have been measured and compared with each other. First, the density of the samples was measured, and the response of these samples to varying compressive force was investigated. The stress-strain diagrams of composites with different matrices were compared. Subsequently, the linear and mass attenuation coefficients of the composites were measured over a wide range of energies (32–1480 keV) against direct and collimated beams from various standard gamma-ray sources (134Cs, 60Co, 154Eu, 133Ba, and 22Na) using an HPGe detector. A collimated beam from an Am–Be source was utilized to measure and compare the attenuation properties of the composite samples against neutron flux. Furthermore, the properties of the composites against neutron and gamma radiation were calculated using the MCNP Monte Carlo code over a broad energy range, and the results were compared and validated with the experimental data. The comparison of the calculated and measured results shows reasonable agreement. The results indicate that in situations where high attenuation properties coupled with high strength are desired, polymeric composites with an epoxy matrix can be considered as radiation shielding materials. If the use of flexible and malleable material is required, such as for reactor piping shielding or the fabrication of protective gloves and clothing in a radiation environment, among the polymers studied, polyvinyl acetate is a more suitable matrix.

Numerical study on the characterization of NdFeB permanent magnets with fast-neutrons induced (n, n′γ) reactions

The potential of prompt gamma analysis based on inelastic scattering of 2.5 MeV neutrons for a rapid characterization of NdFeB permanent magnets is investigated by means of numerical simulations using an HPGe detector and a CZT detector-array. The results show that rapid assay of a 42 g magnet can be achieved in some minutes when the neutron flux at sample position is about 1.6 × 109 cm−2 s−1 and the detector count rate limited to 500 kcps. Such a high neutron flux could be delivered by a compact 5 MeV proton accelerator with a thick beryllium target for neutron production through the 9Be(p,xn)9Be.

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A Monte Carlo simulation study for designing collimators for a CZT-based spent nuclear fuel characterization system

As an attempt to develop a spent nuclear fuel characterization system that can acquire both high-resolution gamma-ray energy spectrum and neutron flux simultaneously at multiple positions around the spent nuclear fuel assembly and provide more information compared to conventional equipment, simulation study was conducted to design a collimator around a gamma-ray detector to obtain high-resolution gamma-ray spectra in a high-flux gamma-ray environment. The ORIGEN depletion code was used to evaluate the source terms of spent nuclear fuel, and MCNP 6.2, the Monte Carlo simulation code, was used to simulate the detector response. Considering the realistic operating conditions of Korea's pressurized light water reactor, the source term for the ACE7 17×17 nuclear fuel assembly was derived under the conditions of initial enrichment ranging from 3 to 5%, burnup ranging from 30 to 60 GWd/MTU, and maximum cooling time of 30 years. MCNP 6.2, the Monte Carlo simulation code, was used to design a spent nuclear fuel characterization system using a CZT detector as a gamma-ray detector to obtain a high-resolution gamma-ray energy spectrum. In particular, the research focused on designing collimators to be used in high-flux radiation environments, such as spent nuclear fuel inspection, with CZT detectors. A simulation model of the CZT detector to be installed internally was made and the validity of the detector simulation model was confirmed by comparing simulation results with the measurement experiment results. Finally, the response of the CZT detector was calculated by changing various setting parameters of the collimator. By comparing the calculated detector response with the design standard count rate, the design of a collimator that can utilize the CZT detector in a spent nuclear fuel inspection environment was derived.

Evaluation of neutron radiation damage in the VVER-1200 reactor pressure vessel

In nuclear reactors, the structural materials of reactor pressure vessel (RPV) are damaged by the radiation that produced by fission reactions. Neutron irradiation damage is one of the most critical factors for the degradation of the reactor pressure vessel (RPV). A displacement per atom (dpa) is a standard measurement for calculating neutron-and gamma-induced radiation damage in materials. The aim of this paper was to evaluate the neutrons radiation damage in the reactor pressure vessel of VVER-1200. In this paper, neutron damage factors, including fast neutron fluxes above 1 MeV, above 0.5 MeV, above 0.1 MeV, the damage rate (dpa/s), and Integrated damage [DPA for one effective full power year (EFPY)] were investigated for VVER-1200/AES-2006 reactor pressure vessel (RPV). The calculations are performed by using NJOY code together with the MCNP6 code at beginning of Cycle (BOC). The cross section library used in present calculations is based on ENDF/B-VII.1 library (ENDF71x). The NJOY code was used for generating DPA cross sections in an ACE formatted file that suitable for using in the MCNP6 code. The MCNP6 code was used for simulating the VVER-1200 reactor core and to calculate neutron DPA rates. MCNP6 simulation of the VVER-1200 reactor core identified the area where the RPV neutron radiation is maximized. The integrated damage for a full effective year was evaluated to be 2.128E-04 DPA/EFPD for the VVER-1200 RPV.

Pulse operation mode of inertial electrostatic plasma confinement devices

The paper discusses the operation features of the inertial electrostatic plasma confinement (IEC) devices in the pulsed high voltage mode. The pulse operation mode offers significant advantages in IEC devices, which are associated with high discharge currents that are inaccessible for continuous operating modes. However, the implementation of such operating modes is associated with a number of physical limitations and technical difficulties.A stand with an IEC chamber powered by a high-voltage pulse transformer and a unit for gas preliminary ionization in the IEC chamber has been developed. In the chamber, ions are accelerated from a background glow discharge. A study of the burning background discharge in the IEC chamber, background discharge current Ipre influence on the duration of current and voltage pulses of the main discharge and the delay between them was carried out. The results of the neutron yield measuring in an IEC chamber operating in pulsed mode are presented at a charging voltage up to 105 kV, a discharge current amplitude of several tens of amperes, a preionization current of 8 mA, and a pulse repetition rate from single to 300 Hz. A stable neutron flux with an energy of 2.5 MeV at a level of 106 neutrons/s was experimentally obtained.

Investigation of VVER-1000 fuel assembly bowing effect on power distribution during cycle using neutron noise adiabatic approximation

Thermomechanical and radiation loads in the reactor core cause lateral changes in the shape of Fuel Assemblies (FAs). This phenomenon, known as FA bowing, has significant implications for reactor safety and operation. Bowing results in alterations to the spacing between FAs, which in turn leads to neutron flux perturbations affecting power distribution. These perturbations are time-dependent due to variations in boric acid concentration during the cycle. Neutron noise theory is employed to analyze these time-dependent neutron perturbations. In this research, it is investigated the bowing effect during the cycle in a VVER-1000 reactor using the neutron noise adiabatic approximation. It was initially considered a VVER-1000 reactor core to verify the calculations algorithm without bowed FAs. Results had acceptable accuracy in power, initial boric acid, and cycle length. Subsequently, it was explored various modes of FA bowing as neutron noise sources the results revealed that the power noise caused by bowing varies throughout the cycle, as expected. The noise amplitude during the cycle depends on changes in boric acid concentration, burnup, size, and the direction of bowing.

Fission counter array for pulse-mode measurements of high-flux and high-energy neutrons

This manuscript describes a neutron counting system based on cylindrical fission counters that can monitor neutron activity for high-energy neutron flux above 10 MeV under electrically noisy environments with intense gamma rays. Miniature fission counters with depleted uranium as sensitive material and modular electronics were built for digital signal processing and high-countrate operation. The counters are 9.5 mm in diameter and 71.1 mm in active length. The author presents the results of Monte Carlo simulations of the fission-counter response for selected neutron sources and energies based on ENDF7.1, JENDL-5, and TENDL-2021 nuclear data libraries from 1 meV to 200 MeV. For a white neutron beam (E‾ = 16.36 MeV) that irradiates the front face of a counter, the intrinsic efficiency is evaluated to be (2.24 ± 0.02) × 10−5 counts/n, while the efficiency of the counter in the array appears to increase by at most 6.7%.

Возможности создания источников излучения для персонализированной брахитерапии на основе 3D-каркасов из сплава титана

Purpose: The study explores the possibility of manufacturing radiation sources for personalized brachytherapy using titanium alloys, activated in a neutron flux reactor, by measuring the radiation composition of applicator implants and their dosimetric characteristics. Material and methods: A 3D implant of a brachytherapy source was made from a titanium alloy using an additive selective laser melting setup. The titanium 3D prototype was irradiated for three days in the horizontal experimental channel of the IR-8 reactor. Subsequently, measurements of the gamma-ray spectrum from the irradiated implant were carried out on a spectrometer, and dose characteristics of the 3D implant were measured using a dosimeter-radiometer. Results: In the experimental 3D implant obtained by us, the radionuclide 47Sc exhibits the highest activity. Currently, 47Sc is considered a promising candidate for brachytherapy. It possesses attractive nuclear and physical properties as a β-emitter, decaying into the ground state (27 %) of 47Ti (Eβmax = 600 keV) and the excited state of 47Ti (Eβmax = 439 keV) with a half-life of 3.4 days. Additionally, 47Sc emits γ-radiation at an energy of 159 keV (68 %), which is suitable for imaging, allowing for SPECT or planar scintigraphy and obtaining a picture of the drug’s distribution in the body. In the experimental implant, small amounts of scandium radionuclides – 46Sc and 48Sc, were also detected, emitting sufficiently hard gamma radiation, which can pose a problem for patient dosage determination. The advantages of using titanium-47 with an enrichment of over 95 %, economically available, have been demonstrated, allowing for high radiochemical yields of 47Sc, sufficient for therapy. Conclusion: The 3D printing technology allows the production of a customized applicator for brachytherapy of specific dimensions and the delivery of arbitrarily-shaped sources to the tumor area for personalized therapy of oncological diseases. When implanting sources based on titanium alloys activated in a neutron flux of a research nuclear reactor, the radionuclide scandium-47 exhibits the highest activity.

Increase in the count rates of ground-based cosmic-ray detectors caused by the heliomagnetic disturbance on 5 November 2023

This letter presents a rare physical phenomenon associated with solar activity, manifesting in anomalies within neutron, electron, and gamma-ray fluxes in the atmosphere. Conventionally, the Earth's magnetic-field disturbances reduce cosmic-ray intensity reaching the surface. However, a temporary surge in cosmic-ray flux occurs intermittently known as the magnetospheric effect (ME). Our observations reveal that this effect predominantly induces a count rate increase in particle detectors positioned at middle latitudes on mountaintops. On November 5, 2023, a 2–3% increase in neutron monitors at mountain altitudes and up to 5% increase in thin plastic scintillators registering electrons and gamma rays was observed. This flux escalation coincided with a southward orientation of the interplanetary magnetic field. Importantly, we present, for the first time, the energy spectrum of the Magnetospheric Effect observed at two mountaintops: Aragats and Zugspitze. Simulations of low-energy proton interactions in the terrestrial atmosphere affirm the augmentation of low-energy cosmic rays. Protons, typically restricted by the geomagnetic cutoff, reached the Earth's atmosphere, generating detectable particle showers on the Earth's surface. To sum up, 1) we measure an increase in the count rate of magnetospheric origin using particle detectors located at mountain altitudes and middle latitudes; 2) for the first time, we measured the energy spectra of the particle fluxes during the magnetospheric effect with spectrometers located on Mount Aragats and Zugspitze; 3) particle flux enhancement coincides with the depletion of the horizontal component of the geomagnetic field; 4) we explain why the magnetospheric effect was observed at mountain altitudes and not at sea level.

A re-activation model for 169Yb intensity modulated brachytherapy sources accounting for spatiotemporal isotopic composition.

Intensity modulated brachytherapy based on partially shielded intracavitary and interstitial applicators is possible with a cost-effective 169Yb production method. 169Yb is a traditionally expensive isotope suitable for this purpose, with an average γ-ray energy of 93keV. Re-activating a single 169Yb source multiple times in a nuclear reactor between clinical uses was shown to theoretically reduce cost by approximately 75% relative to conventional single-activation sources. With re-activation, substantial spatiotemporal variation in isotopic source composition is expected between activations via 168Yb burnup and 169Yb decay, resulting in time dependent neutron transmission, precursor usage, and reactor time needed per re-activation. To introduce a generalized model of radioactive source production that accounts for spatiotemporal variation in isotopic source composition to improve the efficiency estimate of the 169Yb production process, with and without re-activation. A time-dependent thermal neutron transport, isotope transmutation, and decay model was developed. Thermal neutron flux within partitioned sub-volumes of a cylindrical active source was calculated by raytracing through the spatiotemporal dependent isotopic composition throughout the source, accounting for thermal neutron attenuation along each ray. The model was benchmarked, generalized, and applied to a variety of active source dimensions with radii ranging from 0.4 to 1.0mm, lengths from 2.5 to 10.5mm, and volumes from 0.31 to 7.85 mm3, at thermal neutron fluxes from 1×1014to1×1015 n cm-2 s-1. The 168Yb-Yb2O3 density was 8.5g cm-3 with 82% 168Yb-enrichment. As an example, a reference re-activatable 169Yb active source (RRS) constructed of 82%-enriched 168Yb-Yb2O3 precursor was modeled, with 0.6mm diameter, 10.5mm length, 3 mm3 volume, 8.5g cm-3 density, and a thermal neutron activation flux of 4 × 1014 neutronscm-2s-1. The average clinical 169Yb activity for a 0.99 versus 0.31 mm3 source dropped from 20.1 to 7.5Ci for a 4×1014ncm-2s-1 activation flux and from 20.9 to 8.7Ci for a 1×1015ncm-2s-1 activation flux. For thermal neutron fluxes ≥2×1014ncm-2s-1, total precursor and reactor time per clinic-year were maximized at a source volume of 0.99 mm3 and reached a near minimum at 3mm3. When the spatiotemporal isotopic composition effect was accounted for, average thermal neutron transmission increased over RRS lifetime from 23.6% to 55.9%. A 28% reduction (42.5 days to 30.6 days) in the reactor time needed per clinic-year for the RRS is predicted relative to a model that does not account for spatiotemporal isotopic composition effects. Accounting for spatiotemporal isotopic composition effects within the RRS results in a 28% reduction in the reactor time per clinic-year relative to the case in which such changes are not accounted for. Smaller volume sources had a disadvantage in that average clinical 169Yb activity decreased substantially below 20Ci for source volumes under 1mm3. Increasing source volume above 3 mm3 adds little value in precursor and reactor time savings and has a geometric disadvantage.

Characterisation of the neutron field for streaming analyses in TT operations at JET

Assessing radiation fields in the biological shield penetrations of fusion reactors is a challenging task. At the Joint European Torus (JET) the neutron field at larger distances from the torus has been calculated and measured. JET operated in 2021–22 with a tritium-tritium plasma and neutronics experiments were performed for validating in a real fusion environment the neutronics codes and nuclear data applied in ITER nuclear analyses. In particular, the fluence of neutrons passing through the penetrations of the JET vacuum vessel and the torus hall was measured and compared with calculations in order to assess the capability of state-of-the-art numerical tools to correctly predict the radiation streaming in large and complex geometries. The neutron fluence was monitored at several locations inside the torus hall at larger distances from the tokamak with activation foils and thermo-luminescent detectors. The calculations have been performed in a two-step process using the deterministic code ADVANTG to determine the variance reduction parameters and with MCNP for subsequent calculation of the neutron field with the Monte Carlo method. The paper presents results of calculations and the first comparison to experimentally obtained values.

Induction brazing of DEMO's large bore cooling pipes

The DEMO first wall blanket components will be exposed to high neutron fluence and high levels of thermal energy, the latter being removed by a coolant fluid transmitted out of the reactor via service piping. Due to the thermal and nuclear loading on the first wall, the components will need to be periodically removed and replaced. Installation of new blankets requires the joining of their service pipes by remote handling.Brazing is one of the alternatives that are considered for joining the service pipes of the DEMO breeding blankets. This technique has great potential and has several major advantages over welding.The Fusion Technology Department of HUN-REN Centre for Energy Research (CER) has been working on the development of a viable brazing concept of thick, large bore pipes inside the ports of DEMO. Induction has been chosen as the heat source for the brazing operation, which is a well-studied technology and has many uses in the industry already. However, brazing in this size is a completely new horizon for induction heating so CER's work includes a significant amount of development. Firstly, the coil is naturally a bespoke component, which has been optimized by electromagnetic analysis. Secondly, a test rig has been designed and manufactured in-house by CER, to test the performance of the induction heater in a real scenario.This paper outlines the development strategy of the induction brazing of DEMO's large bore pipes.

Simulation of Neutron Leakage Variations at Fuel Substitution in a Small Modular Reactor and Implications for Unattended Safeguards Monitoring

According to a recently proposed nuclear safeguards technique, monitoring the power-normalized, ex-core neutron detection rate over time could be used to detect undeclared changes to the fissile composition of a reactor core. In this study, Monte Carlo simulations have been used to verify some of the underlying assumptions of this technique and the possibilities of using it to detect undeclared fuel substitutions during the first 2-year cycle of a light water small modular reactor. Depletion calculations and neutron transport simulations were used to study the changes in the power-normalized neutron leakage rate J b / P core through the core barrel upon fuel substitutions and whether these changes are fully explained by changes in the core fissile composition. Several substitution scenarios have been studied, where partially depleted fuel assemblies were substituted with fresh fuel assemblies after 1 year of irradiation. The modeled substitution scenarios, which included substituting up to 4 out of 37 fuel assemblies in the core at a time, resulted in changes in J b / P core of up to 3.5% depending on which fuel assemblies were substituted. The results indicate that the ex-core neutron signatures are not only sensitive to core-averaged nuclide densities, fission cross sections, and neutron flux, but also the spatial distributions of these and other parameters throughout the core. Effects such as these could mean that monitoring the core fissile composition with the proposed technique might be more complex than previously suggested.

Production and transfer of cosmogenous tritium in the Earth atmosphere in the «PARMA» model

The paper analyses contribution of reactions (n,X) and (p,X) on nitrogen and oxygen nuclei to the production and transfer of cosmogenic tritium in the Earth atmosphere. To calculate the fluxes of secondary protons and neutrons in the nuclear-electromagnetic cascade, the analytical model «PARMA» was used. The model is based on analytical approximations of both numerous experimental data and computer simulation data, and allows one to calculate the fluxes of various particles of secondary cosmic radiation (nucleons, muons, a- and b-particles) with the choice of a given energy range, atmospheric height, geomagnetic cutoff rigidity, solar activity. The problem of vertical turbulent diffusion of tritium in the Earth's atmosphere was solved and the dependence of its concentration on altitude was obtained. Numerical integration of the transport equations was carried out using the integro-interpolation method. An estimate of the total reserve of cosmogenic tritium in the Earth's atmosphere, balanced by turbulent diffusion, decay and cosmogenic formation, was given, which amounted to ~1.9∙1017 Bq. The calculation results show that the earth's atmosphere contains no more than 10% of all cosmogenic tritium. The results are generally consistent with earlier empirical and semi-empirical models confirming that man-made tritium currently accounts for the majority of the total tritium supply

Maxwell Spectrum as a Parameter to Verify the Dose in Brain Cancer (Glioblastoma) by Boron Neutron Capture Therapy (BNCT) using Monte Carlo Method

To evaluate the efficiency of neutron capture therapy (BNCT) treatment in glioblastoma multiforme, it is necessary to evaluate the impact of the neutron beam on the tumor cell and find better results so that BNCT treatment is viable. Glioblastoma multiforme is one of the most lethal cancers and conventional radiotherapy is almost ineffective for this type of tumor. Among several approaches to describe the procedure and the neutron spectrum, the Maxwell spectrum in the epithermal neutron range was used. For this, T=0.0025 MeV was used to describe this spectrum. MCNP software was used to simulate a BNCT treatment using the Maxwell spectrum to describe the neutron source. The user provided the quantities of interest, such as fluence and dose. These are extremely important quantities to describe a BNCT planning protocol. A concentration of 30 ppm of Boron-10 was simulated in the tumor. Output data provides normalized values. It was necessary to carry out some mathematical operations to obtain values closer to reality. Thus, a dose of 32 Gy was obtained for the Maxwell spectrum described with T=0.0025 MeV and a neutron fluence of 1.5 x 1012 n/cm². The values calculated based on the simulation in MCNP5 described by an epithermal neutron source obeying a Maxwellian function, were in agreement with the reference values in the literature.

Development of radiation sources based on CAD models for the nuclear analysis of IFMIF-DONES lithium loop

IFMIF-DONES is a neutron source designed to irradiate materials to be used in future fusion power reactors such as DEMO. The facility is based on a deuteron beam impinging onto a liquid lithium jet to generate the neutron flux. Lithium and Corrosion Products will get activated and produce Be-7 and Activated Corrosion Products (ACP). These products will be distributed along the lithium loop, both dissolved in lithium and deposited locally. This complex gamma source should be properly represented to perform radiological safety studies. A new tool, called CAD2CDGS, was created to represent these sources. This tool creates decay gamma sources based on CAD geometries and codifies them into CDGS format. Sources are specified in the CAD model, allowing a user-friendly approach. The tool is based on Open-Source tools (FreeCAD) and EUROfusion codes (cR2S European SDDR tool). In this article the CAD2CDGS tool methodology, inputs and workflow are explained. Two verification tests have been done to check the correctness. Finally, it is demonstrated the applicability of the tool to IFMIF-DONES, with the radiological zoning analysis of the rooms surrounding the Lithium Loop Cell impact by the Be-7 and ACP.

Verification of the McStas code using two double axis neutron diffractometers

The Monte Carlo ray-tracing simulation package McStas is verified by modeling two double axis neutron diffractometers, KARL and KANDI-II, located at the Israel Research reactors IRR-1 and IRR-2, respectively. The advantage of simulating these instruments is their simple design and small number of components. We focus on the simulation of a Soller slit collimator as well as Pyrolytic Graphite and Copper single crystal monochromators. The simulation is compared to neutron flux measurements, performed via activation analysis along the neutron beam as well as observed diffraction patterns from different powder samples. Calculated diffraction patterns are found to agree with observations to within 20%–50% in reflection peak width, while calculated attenuation of the neutron beam agrees to within 20% when the entire instrument is considered, and can show a discrepancy of up to 60% for individual components. Finally, a possible optimization of the neutron flux in the KARL diffractometer is presented.

Effects of Gamma-ray and neutron irradiation on infrared optical properties of ZnSe and ZnS for ITER divertor thermography

Gamma-ray and neutron irradiation tests were performed by using antireflective-coated and non-antireflective-coated samples of ZnSe and ZnS to examine the impact on their infrared optical properties. In neutron irradiation tests, ZnSe and ZnS samples showed minimal transmittance reduction even up to 1.44 × 1016 n/cm2 (evaluated by dpa value, the values are 5.77 × 10−6, 2.07 × 10−5 for ZnSe and ZnS, respectively), about 5.8 times higher than the expected neutron fluence in ITER. Gamma irradiation at 100 kGy did not affect any of the samples, however, at 5.6 MGy (2.6x the expected dose of ITER), non-antireflective-coated ZnSe and ZnS exhibited induced absorption bands. Antireflective coating did partially mitigate this effect, but ZnS, even with antireflective coating, showed up to 2.5 % degradation over the wide wavelength range of 1.5–4.5 µm, whereas ZnSe was less affected. Both ZnSe and ZnS exhibited a modest 2 % absorption around 3 µm. Surface analysis by SEM, laser microscope, and XPS linked the 3-µm degradation to gamma-irradiation-induced Zn(OH)2 formation that was confined to the surface. Thus, transmittance degradation of ZnSe over the lifespan of ITER was estimated to be a maximum of approximately 7.8 %. The practical use of ZnSe in the IRTh system requires further evaluation with Si and CaF2 lenses and for the optical design to be finalized. Nevertheless, these findings suggest that ZnSe is a promising candidate for the IRTh system's lenses.