Thermal Neutron Absorption Cross Section Research Articles

Zircoborate glass-ceramic composite reinforced with aluminum oxynitride: Microstructural evolution, physical properties, and charged radiation attenuation analysis

This study used the powder metallurgy method to create aluminum oxynitride (AlON)-reinforced zircoborate glass-ceramic composites. The solid-phase reaction method was used to synthesize AlON powder, which was then added to Zr-based glass-ceramic in different amounts (0, 4, and 8 % by weight) to make it stronger and harder. The prepared materials were characterized for their structural, physical, and radiation shielding properties. The density of GZr8 (AlON-free) had a value of 2.9 g/cm3, and the sintered glass-ceramic density with 4 % wt. AlON reinforcement (GZr8Al4) reached 3.05 g/cm3. The highest density of 3.11 g/cm3 was obtained for the GZr8Al8 sample with 8 % AlON wt. The Vickers hardness for GZr8, GZr8Al4, and GZr8Al8 was measured to be 4.89, 5.94, and 6.33 GPa, respectively. The AlON-reinforced GZr8 samples had fast neutron (FN) removal cross-sections of 0.0796 cm−1 and 0.0824 cm−1 for AlON content of 4 and 8 wt%, respectively. Generally, the coherent scattering cross-section (σcoh), incoherent scattering cross-section (σinc), and absorption cross-section (σabs) of thermal neutrons (TNs) increase from 0.047856 cm−1, 0.02072 cm−1, and 0.01498 cm−1, respectively for GZr8 to 0.53124 cm−1, 0.02074 cm−1, and 0.01795 cm−1, for GZr8Al8. AlON was proven to improve the ability of the Zr-based glass-ceramics to attenuate FNs and TNs. Stopping powers of electrons, protons, alpha particles and carbon ions increase with the AlON content of the materials. Comparatively, the range of charged radiation in AlON-rich samples is lower. The inclusion of AlON in the matrix of the Zr-based glass-ceramics makes it more efficient at absorbing charged radiation and can be used in ion beam or hadron therapy, ion beam analysis, or other charged radiation environments as an absorber.

Evaluation of shielding performance and light emission characteristics of luminescent radiation shields using Gd2O2S

ABSTRACT Gadolinium oxysulfide (Gd2O2S: GOS) has a large absorption cross-section for thermal neutrons and a mass density similar to that of iron used for shielding against X and γ-ray radiation. During GOS scintillator production process, its waste is also generated. If GOS waste can be used as an effective shielding material, it is worthwhile as beneficial use of the waste. We developed a novel radiation-shielding material using GOS waste. The developed shielding material can shield X, γ-rays and thermal neutrons more efficiently than conventional shielding materials and emit visible light due to leaked radiation when the shielding material is damaged, which subsequently be used to check the integrity of shielding using visible light. In this study, two types of shields, concrete and mortar with GOS as the aggregate and epoxy resin with GOS were prototyped and tested for radiation transmission. Results indicate that GOS shielding has a 10% higher shielding ability for X and γ-rays than heavy concrete with iron as aggregate. Additionally, the same shielding effect for thermal neutrons was attained at 1/10 the thickness of heavy concrete. Thus, using GOS waste is not only cost effective but also contributes to a circular economy.

Accident tolerant fuel design optimization for SMART reactor

Fully Ceramic Microencapsulated (FCM) fuel is one of the most advancedfuel types for the next nuclear reactor generations. The use of uranium carbide fuel in FCM offers multiple benefits such as high fuel density, improved thermal conductivity and enhanced mechanical stability. Both UC and conventional fuel (UO2) have the same melting point. As a result, under reactor operating conditions, UC has better safety margins. In addition, FeCrAl is a promising cladding material. However, because of its greater thermal neutron absorption cross section, FeCrAl has a neutronic penalty when compared to Zircaloy leading to an annual reload requirement with high fuel enrichment. Owing to their low power density and excellent design, small modular reactors (SMRs) have exceptional inherent safety. In this work, we determined the optimum design of an accident tolerant fuel (ATF), with respect to neutronics, in SMART reactor using dense UC FCM (in contrast to the conventional UO2 fuel) and a combination of zircalloy and FeCrAl (FCM -Zircaloy, FCM -FeCrAl, FCM -combining of zircaloy and FeCrAl) as cladding to achieve better thermal conductivity as well as safer performance, while considering the required fuel enrichment for the proposed designs. The effective multiplication factor and burnup were calculated for neutronic assessment of the SMART core, the outcomes are showcased alongside a comprehensive assessment of the characteristics and prospective. The neutronic parameters were estimated using the stochastic MCNPX 2.6 code. We found that Fully Ceramic Microencapsulated -Uranium Carbide (FCM-UC) fuel type was more attractive than conventional UO2. The optimum cladding design as compared to the reference case and the other cases of (FCM-UC), was determined. When 50% FeCrAl cladding was employed, case 15 in our study emerged as the superior design, demonstrating maximum production of 239Pu due to higher initial concentrations of 235U and quantities of 238U. This work may offer recommendations for future optimized designs of the SMART reactor parameters.

Monte Carlo simulation and optimization of neutron ray shielding performance of related materials

Polymers have become widely used substrate materials for shielding neutron rays because of their high hydrogen content and easy processing procedures. Rare-earth materials are also being gradually adopted as neutron absorbers because of their considerable thermal neutron absorption cross-sections. This paper utilizes the FLUKA Monte Carlo simulation program to compare the shielding effects of various polymers and rare-earth oxides on neutron rays across different energy ranges. The study investigates the superior shielding materials for neutron radiation in each energy range. Subsequently, a series of materials are simulated by combining the preferred shielding materials for neutron rays in each energy range, exploring the influence of material composition and composite structure on the effectiveness of neutron ray shielding. It is revealed that the preferred material for shielding neutron rays changes for different energy ranges. For low-energy neutron rays, rare-earth oxides such as Sm2O3 and Gd2O3 demonstrate the most effective shielding, whereas for high-energy neutron rays, polyethylene (PE) provides the best shielding performance. Materials with different compositions show varying preferred structures when dealing with a 252Cf neutron source. However, in mitigating the secondary gamma rays generated during the neutron shielding process, stacked-type materials exhibit the most effective shielding performance.

Neutron shielding analysis for a gadolinium doped nickel alloy

Rare-earth-nickel (RE-Ni) alloy is a kind of appealing neutron shielding material for advanced reactors and spent fuel storage containers. The incorporation of neutron-absorbing additive RE elements results in a RE-Ni alloy with integrated structural and functional features, which can drastically reduce shielding material weight. Gadolinium (Gd), in comparison to other rare-earth elements, has a much larger thermal neutron absorption cross section, necessitating only a low fraction to achieve the desired shielding capability. Hence, gadolinium is doped into a nickel-based alloy to form a gadolinium-nickel (Gd-Ni) alloy. Due to its extremely low solubility in the nickel phase, gadolinium is commonly manifested as heterogeneously spatially scattered Gd-rich phases in existing preparation procedures of alloys rather than homogeneously distributed gadolinium elements. This would result in a strong neutron tunneling effect, limiting the neutron shielding capability of Gd-Ni alloys. In this work, a superfine layered Gd-Ni alloy doped with Gd-rich phases and arranged in dislocations is proposed to provide an adequate dislocation density of Gd-rich phases to suppress the neutron tunneling effect. With the same Gd content, the thermal neutron shielding ability of a layered Gd-Ni alloy is comparable to the case that Gd is assumed to be an ideally homogeneous distribution in the material. The smaller the particle size of the Gd-rich phase, the weaker the neutron tunneling effect, and the lower the thermal neutron transmission ratio. For a 0.5 cm-thick Gd-Ni alloy, the particle size of Gd-rich phase should be kept within 20μm to achieve a thermal neutron transmission ratio of less than 10−5 for natural gadolinium content of 0.6 at.%. This work provides an important guidance for the design and fabrication of the next-generation neutron shielding materials.

Numerical analysis of multi-field coupling behaviors of delayed hydride cracking for thin plate-type Zr-2.5Nb specimens

Zirconium alloys are extensively utilized as structural materials in nuclear applications due to their exceptional mechanical strength, corrosion resistance, and low thermal neutron absorption cross-section. The stress concentration provoked by the defects in zirconium alloy components will initiate delayed hydride cracking (DHC), significantly jeopardizing nuclear safety. In the present investigation, an enhanced cohesive model that incorporates the effects of stress state has been proposed to realize the three-dimensional simulation of multi-field coupled DHC behaviors in thin plate-type zirconium alloy specimens. The simulation results indicate that: (1) the consistence of the predicted DHC velocities with the corresponding experimental data validates the newly developed cohesive model; (2) the strip characteristics presented in the cracked surface are captured, demonstrating that the varying fracture patterns across the thickness are well simulated; (3) the stress-state induced hydrostatic-stress gradients and solid-soluted hydrogen concentration gradients make against the formation of a hydride strip to induce hydride embrittlement near the plate surfaces, resulting in a slower cracking of the surface layers than that of the internal region; for the thinner specimens the asynchronous cracking phenomena through the thickness will greatly affect the multi-field coupling behaviors around the mid-plane; the non-cracked surface layers will apply the dragging effects to the new cracking of the internal region, leading to longer waiting time and thicker hydrided zone before fracture.

Assessment of minimum allowable thickness of advanced steel (FeCrAl) cladding for accident tolerant fuel

The ferritic iron-chromium-aluminum (FeCrAl) alloy cladding is considered to be the most promising for near-term application in the ATF framework to replace existing zirconium alloy cladding. Although FeCrAl cladding presents several advantages, it is well known that there are at least two main drawbacks, one is the increased thermal neutron absorption cross-section compared to the current Zr-based cladding resulting in a neutronic penalty and another is tritium higher permeation. In the present study, the minimum allowable thickness of cladding is addressed considering neutronic penalty reduction and the mechanical-structural behavior under the LOCA accident condition. The neutronic penalty assessment was performed using the Monte Carlo code and mechanical-structural performance of the FeCrAl cladding using the TRANSURANUS fuel code, which was modified to consider properly the FeCrAl cladding.

Parametric neutronic analysis of different cladding options for ThO2 pellet of advanced dual-cooled annular PWR assembly

This study presents a parametric neutronic evaluation of different candidate claddings for a purposed candidate fuel (Th-233U-235U)O2 in an advanced dual-cooled annular PWR assembly. The investigated claddings include Zircaloy-2/4, zirconium carbide (ZrC), silicon carbide (SiC), ferritic Fe-20Cr-5Al (FeCrAl) and APMT alloys, and austenitic 310 (SS-310), compared to the by design cladding for the AP1000 assembly, Zirlo™. Neutronic calculations were performed using the deterministic code DRAGON, and the parameters included infinite multiplication factor, criticality periods, neutron flux spectra, radial power distributions, rim effects, and reactivity feedback coefficients.The results show that the candidate fuel with Zirlo™ achieves a higher criticality period than UO2 fuel in both solid and annular configurations. The candidate fuel with SS-310 cladding exhibits a higher criticality period penalty of about 12.60% compared to Zirlo™, APMT, and FeCrAl claddings show a criticality period penalty of about 10.94% and 8.79%, respectively. The candidate fuel with ceramic claddings, SiC and ZrC, can achieve slightly higher criticality periods compared to Zirlo™ of about 1.11% and 0.24%, respectively, due to their low thermal neutron absorption cross-sections. Regarding the power profile, the annular (Th-233U-235U)O2 fuel exhibits a flatter behavior compared to UO2 fuel. However, the claddings with lower capture cross-sections exhibit slightly increased fission power through the fuel pellet. The fuel temperature coefficient for candidate fuel exhibits more negative values as the capture cross-section of the cladding increases, while the molybdenum and zirconium-containing claddings exhibit more favorable moderator temperature coefficients.

Characterization of local deformation around hydrides in Zircaloy-4 using conventional and high angular resolution electron backscatter diffraction

Zircaloy-4 is used as a fuel cladding material for water reactors, as it has good mechanical properties, corrosion resistance, and a low thermal neutron absorption cross section. However, the mechanical performance of Zircaloy-4 can be reduced during service due to hydrogen uptake and hydride formation. These hydrides are brittle, and often reduce the strength and toughness of materials as well as increase susceptibility to delayed hydride cracking (DHC). In this work, large grain Zircaloy-4 with hydrides was prepared and then cross sectioned using cryo-ion beam polishing, using plasma focused ion beam (pFIB) and broad ion beam (BIB) approaches to enable the preparation of a very high quality flat surface with no preferential etching of either the hydride or zirconium metal (typically metallographic polishing preferentially removes hydrides). Conventional and high angular resolution electron backscatter diffraction (EBSD) analysis were then used to explore morphology, deformation fields, and orientation relationships between the zirconium matrix and hydrides. Four maps were collected for analysis which included hydrides near grain boundaries: (a) where the hydride smoothly decorates across two of the connecting boundaries near a triple junction; (b) where the hydride smoothly decorates the boundary; (c) a mixture of smooth decoration of the interface and protrusion into the grains; (d) fine scale hydride that protrudes into one grain. This work highlights that incompatibility of the hydride within the zirconium matrix is strongly linked to the orientation relationship of the hydride and matrix, and the grain boundary character. These results may enable enhanced understanding of the role of hydrides in fracture as well as stress-induced hydride reorientation and DHC susceptibility.

Characterization and applications of highly optical transparency tellurite glasses doped with Er3+, Tm3+, and Nd3+

Highly optical transparent tellurite glasses are attractive and viable as alternative materials for absorbing gamma-ray photons in the design of radiation protection structures. This report examines the potential of new tellurite glass systems (TeO2-ZnO-WO3-Bi2O3) doped with Er3+ (TZWB-Er), Tm3+ (TZWB-Tm), and Nd3+ (TZWB-Nd) as shielding materials against charged and uncharged radiation. The mass attenuation coefficient, μ/ρ of the glasses was computed by XCOM software and data from simulations of photon transmission in the PHIT code. The total stopping powers and range of beta particles β, H+, He2+, and C6+ in the glasses were estimated by ESTAR, PSTAR, ASTAR, and SRIM software at particle kinetic energies within 0.1–10 MeV. The fast neutron (FN) removal and thermal neutron absorption cross-sections in the glasses were also computed. The stopping powers and ranges of the considered charged particles showed variations that depended on the kinetic energy of the particles and chemical composition of the glasses. The TZWB-M glasses are better alternatives as gamma-radiation shields compared to traditional shielding substances such as concrete, some commercial glass shields, and recently recommended glass shields. The glasses are thus recommended for the design of radiation safety structures.

Microstructural and mechanical property evolution of a nuclear zirconium-4 alloy fabricated via laser powder bed fusion and annealing heat treatment

ABSTRACT Zirconium (Zr) alloys are widely used in nuclear energy because of their excellent mechanical properties and low thermal neutron absorption cross-section. This work investigated the printability, microstructure, and mechanical properties of Zr-4 alloy additively manufactured by laser powder bed fusion (LPBF) for the first time. The effect of annealing temperature on the microstructural and the mechanical property evolution of the printed Zr-4 alloy was studied. The results exhibited that the Zr-4 alloy with a high relative density of 99.77% was obtained using optimised printing parameters. With an increase in the annealing temperature, the formed α phase of the Zr-4 alloy changed from an acicular shape to a coarse-twisted shape, and finally to an equiaxed shape. Such microstructure change endowed the alloy with a high compressive strength of 2130 MPa and compressive strain of 36%. When the annealing temperature exceeded 700°C, Zr x (Fe2Cr) compounds were precipitated, strengthening the alloy by pinning effect. These findings provide valuable guidance for the manufacture of geometrically complex Zr alloy parts for nuclear power applications.

A synergistic effect of heavy metal oxides to enhance the physical, optical, and radiation-absorption properties of TeO2-Li2O-BaO glasses

A series of 75TeO2-(20-x)Li2O-xBaO glasses was successfully synthesized using the melt-quenching method. The glass samples were labeled as TLB05, TLB10, TLB15, TLB20 and TLB25 for x = 5, 10, 15, 20 and 25 mol% respectively. The prepared glasses were studied for their structural, optical, and radiation shielding properties. The optical band gap of the glasses decreased linearly from 3.45 to 3.28 eV with BaO in the range 5−25 mol%. The analysis of the thermal properties shows an increase in Tg values with the replacement of Li2O by BaO. The gamma ray shielding coefficients for the glasses estimated by FLUKA simulation and validated by XCOM varied from 0.0329 to 1.4383 cm2g−1, 0.0332 – 1.4879 cm2g−1, 0.0334 – 1.5327 cm2g−1, 0.0335 – 1.5671 cm2g−1, and 0.0336 – 1.5964 cm2g−1 for BaO concentration of 5%, 10%, 15%, 20%, and 25% respectively. Generally, the introduction of BaO improved the photon and charged particles (electron, proton, alpha particle, carbon ions) absorbing parameters of the glasses. The fast neutron removal cross section of the TLB glasses grew from 0.0995 to 0.1014 cm−1 as the molar concentration of BaO declines from 25 to 5 mol%. The increase in the BaO content also quenched the thermal neutron absorption cross section. For TLB5 – TLB25, the coherent, incoherent, absorption, and total thermal neutron cross sections varied between 1.0582 and 1.2059 cm−1, 0.1251 – 0.0024 cm−1, 9.5411 – 0.0876 cm−1, and 10.7244 – 1.2958 cm−1 accordingly. In general, TLB25 displayed the most effective photon shielding ability but the weakest neutron absorber. The TLB glasses in addition showed superior radiation shielding ability compared to some conventional shields.

Effects of irradiation on the multi-field coupling delayed hydride cracking behavior of zirconium alloys

With high mechanical strength, good corrosion resistance and small thermal neutron absorption cross-section, zirconium based alloys are widely used as the nuclear structural materials. Stress concentration usually appears in zirconium-based-alloy components during their reactor service phase or the storage stage of spent fuels, which could induce the delayed hydride cracking (DHC) behavior to seriously threaten the nuclear safety. In this study, an innovative multi-field coupling theoretical frame is developed, with the effects of texture factor, with the irradiation hardening and embrittlement effects as well as the hydride orientation contributions involved in the cohesive model, with the hydrogen diffusion affected by the irradiation-induced variations of Nb concentration. The multi-field coupling simulation results indicate that (1) the predictions of the DHC velocities for different irradiation doses are in good agreement with the experimental data, and the subcritical cracking traits can be captured, which validates the effectiveness of the newly developed models; (2) the irradiation-enhanced hydrogen diffusion coefficient plays a dominant role in the increase of DHC velocity with the radiation dose; (3) the irradiation hardening leads to the increases of hydrostatic stress gradient and the concentration gradient of solid-soluted hydrogen atoms, which accelerates the diffusion flux of hydrogen atoms towards the crack tip region; (4) essentially, it is the irradiation hardening that shortens the required length of hydrided zone to initiate new fracture, and quicken the multi-coupling cracking process. This work lays a foundation for the further study of delayed hydride cracking behavior.

Evaluation of thermophysical properties of UO2-10 Vol% Mo nuclear fuel pellets

Uranium dioxide (UO 2 )−molybdenum (Mo) composites have been proposed as accident tolerant fuel candidates due to their high thermal conductivity , high melting point , high boiling point, and ability to have low thermal neutron absorption cross-section. It is therefore essential to evaluate the method of fabrication and thermophysical properties of this system. In this work, UO 2 -Mo samples with a Mo fraction of 10 Vol% and different densities were fabricated using spark plasma sintering (SPS). This method allows sintering at relatively low temperatures and within a short time. Experimental analysis was performed on UO 2 -Mo specimens to characterize thermal diffusivity and thermal conductivity as a function of temperature up to 900 °C. Thermophysical properties were analyzed for various densities, and grain sizes as a function of temperature. The grain size and microstructural differences related to porosity were also analyzed using the scanning electron microscopic (SEM) and electron backscatter diffraction (EBSD).

Preliminary neutronic analysis of alternative cladding materials for APR-1400 fuel assembly

This paper presents a preliminary neutronics analysis for Accident Tolerant Fuel (ATF) cladding materials for standard APR-1400 reactor A0 fuel assembly. AMPT, FeCrAl, 304 and 310 stainless steels, and SiC were considered and compared with the original Zircaloy-4 cladding material. A parametric evaluation was done for fuel and cladding materials to confirm the geometry requirements to achieve the end-of-cycle fuel reactivity, and the results were compared with the standard APR-1400 reference fuel-cladding system. Serpent Monte Carlo reactor physics code version 2.31 was utilized to evaluate the associated neutronics penalty when different cladding materials were used such as, fuel reactivity, thermal neutron spectrum, plutonium and isotopic inventory evolution, and fuel assembly linear pin power distribution. In the examined cases, fuel enrichments were fixed, and cladding thickness and pellet diameter were changed for zircaloy, cases 1 – 4. As a result, thermal neutron absorption cross section varies between materials due to higher absorbing cross section leads to hardening of the neutron spectrum in the fuel-cladding system. That leads to a slight increase in the production of actinides in the inventory such as plutonium.

Large-scale production by centrifugation of isotopically modified molybdenum for nuclear reactors and its cost evaluation

Molybdenum is considered as one of the substitutes for zirconium in making fuel cladding for safe reactor designs. The isotopically modified molybdenum (IMM) should be used instead of natural molybdenum to achieve higher energy efficiency, reduce uranium consumption, and mitigate the impact on the nuclear fuel cycle. This task can be fulfilled by using the same technology of gas centrifugation for uranium enrichment to remove the isotopes in natural molybdenum with larger thermal neutron absorption cross sections. Optimized double-cascade separation schemes consisting of the Q-cascade model are employed for studying massive production of IMM and evaluating the economic feasibility of production. A simple formula is derived to be able to assess the cost per kilogram of IMM, taking into account various cost-affecting factors. The results demonstrate that it is feasible to produce massively the IMM with a total thermal neutron absorption cross section equivalent to that of zirconium.

Explore the possible advantages of using thorium-based fuel in a pressurized water reactor (PWR) Part 1: Neutronic analysis

This study discusses the effect of using 232Th instead of 238U on the neutronic characteristics and the main operating parameters of the pressurized water reactor (PWR). MCNPX version 2.7 was used to compare the neutronic characteristics of UO2 with (Th, 235U)O2 and (Th, 233U) O2. Firstly, the infinity multiplication factor (Kinf), thermal neutron flux, and power distribution have been studied for the investigated fuel types. Secondly, the effect of Gd2O3 and Er2O3 on the Kinf and on the radial thermal neutron flux and thermal power has been investigated to distinguish which of them is more suitable than the other in reactivity management. Thirdly, to illustrate the effectiveness of 232Th in decreasing the inventory of both the actinides and non-actinides, the concentration of plutonium (Pu) isotopes and minor actinides (MAs) has been simulated with the fuel burnup. Besides, due to their large thermal neutron absorption cross-section, the concentrations of 135Xe, 149Sm, and 151Sm with the fuel burnup have been investigated. Finally, the main safety parameters such as the reactivity worth of the control rods (ρCR), the effective delayed neutron fraction βeff, and the Doppler reactivity coefficient (DRC) were calculated to determine to which extent these fuel types achieve the acceptable limits.

Determination of boron in in-house graphite reference material by instrumental charged particle activation analysis

Due to high thermal neutron absorption cross-section, boron (B) content is a binding specification for nuclear grade graphite. Reliable analytical methods are essential to determine boron at trace to ultra-trace levels. Analytical Chemistry Division, Bhabha Atomic Research Centre, India has prepared an in-house graphite reference material (GRM). A Charged particle activation analysis methodology has been developed for quantification of B in graphite matrix. Detection limit for boron using this methodology is 1.4 mg kg−1. This in-house GRM can be used for quantification of B in unknown graphite samples, validating existing analytical methods and for the development of new analytical methods.

Three-dimensional (X-Y-Z) Core Design of Long-Life Pressurized Water Reactor Using (Th-U)O2 Fuels with The Addition of Gd2O3 and Pa-231 as Burnable Poisons

Pressurized water reactors (PWRs) are one of the most dominant types of nuclear power plants that have been operated commercially to produce electricity in the world. The purpose of this study was to perceive a three-dimensional (X-Y-Z) core design of long-life PWR using Thorium-Uranium dioxide ((Th-U)O2) fuels with the addition of Gadolinium (Gd2O3) and Protactinium-231 (Pa-231) as the burnable poisons. A combination of Thorium and enriched Uranium fuels have a higher conversion ratio than other fuels, therefore can guarantee the reactor to operate longer. The burnable poison isotopes could be used to reduce excess reactivity due to the very high thermal neutron absorption cross-section. For core geometry analysis, a three-dimensional (X-Y-Z) geometry and a fuel volume fraction of 40% were applied. The computer code of SRAC 2006 from the Japan Atomic Energy Agency (JAEA) and the JENDL 4.0 as a nuclear data library were used for calculation. In this study, different fractions of Uranium dioxide, Uranium-235, Gadolinium, and Protactinium-231 in fuel were carried out. The result of this study was a three-dimensional core design of 800 MWt PWR using 60% Uranium dioxide fuel with enriched Uranium-235 of 12%-11% and the addition of 0,025% Gd2O3 and 1,0% Pa-231 which could operate for ten years without refueling. This research is expected to be a reference for long-life PWR design using the Thorium and Uranium fuel cycles.

TeO2–B2O3–ZnO–La2O3 glasses: γ-ray and neutron attenuation characteristics analysis by WinXCOM program, MCNP5, Geant4, and Penelope simulation codes

γ-ray, fast and thermal neutrons shielding features for five Pb-free {[(TeO2)0.70(B2O3)0.30]0.7 (ZnO)0.30}(1-x)(La2O3)x (x = 0.01, 0.02, 0.03, 0.04, and 0.05 mol%) glasses have been inspected in this work. For all samples, within the photon energy range of 0.015–15 MeV, linear attenuation coefficients (μ) and mass attenuation coefficients (μ/ρ) were estimated using WinXCOM program (theoretical approach), and obtained μ/ρ values are in fine agreement with computed μ/ρ results of respective MCNP5, Geant4, and Penelope codes. By utilizing μ/ρ values of WinXCOM, γ-ray attenuation parameters like effective atomic number (Zeff), effective electron density (Neff), half-value layer (HVL), and mean free path (MFP) were assessed within 0.015–15 MeV photon energy range, for all samples. Additionally, exposure buildup factors (EBFs) and energy absorption buildup factors (EABFs) were calculated for all samples by applying a five parameter geometric progression (G‒P) fitting method as a function of various penetration depths (1, 5, 10, 15, 20, 25, 30, 35, and 40 mfp) within 0.015–15 MeV photon energy range. The derived radiation protection efficiency (RPE) results indicate all samples' effective attenuation competence for lower energy γ-rays. Among all glasses, 46.55TeO2-19.95B2O3-28.5ZnO–5La2O3 (mol%) sample, by having relatively bigger μ/ρ and Zeff, minimal HVL and MFP, and lowest EBF and EABF, endorse its' excellent γ-ray shielding effectiveness. Further, neutron shielding parameters such as macroscopic removal cross-section for fast neutrons (ΣR), coherent and incoherent scattering cross-sections (σcs & σics), absorption cross-section (σA), and total cross-section (σT) for thermal neutrons were explored. By using the Geant4 code, ‘σT’ values have been deduced within 10−4‒10−8 MeV neutron energy range. The 46.55TeO2-19.95B2O3-28.5ZnO–5La2O3 (mol%) glass owns relatively larger ΣR (0.1474 cm−1) for fast neutrons attenuation and ‘σT’ (10.1444 cm−1 at 0.0253 eV neutron energy, 15.9824 cm−1 → 0.551 cm−1 from 1 × 10−8 MeV →1 × 10−4 MeV neutron energy) for thermal neutrons absorption, respectively, revealing its' superior neutrons shielding ability.