High Neutron Absorption Cross Section Research Articles

Effectiveness of green-synthesized nickel-doped calcium ferrite nanoparticles in the X-ray/gamma radiation shielding applications

Nanoparticles comprising calcium ferrite doped with Nickel within a concentration range of 10 to 50 mol% were successfully produced using a solution combustion technique, employing neem leaf extract as a reducing agent. The subsequent calcination process at 500 °C validated the formation of orthorhombic calcium ferrite, featuring distinct Bragg reflections indicative of Nickel doping. An analysis via scanning electron microscopy (SEM) affirmed the presence of irregularly shaped nanoparticles with pores and voids within their structure. The optical energy band gap, determined through Wood and Tauc’s relation, ranged from 2.89 to 2.97 eV. Notably, an increase in Nickel substitution was found to correlate with an increase in the energy band gap and a reduction in crystallite size. The research extensively evaluated the nanoparticles’ efficacy in radiation protection, encompassing parameters such as mass attenuation coefficient, mean free path, half-value layer, and effective atomic number. The mass attenuation coefficient is found to be 0.05353 cm2g−1 for 40 mol% at 1332 KeV. EABF is found to be large for 40 mol% at 0.89 MeV energy. Likewise, 50 mol% has a high Neutron absorption cross section (σab) which is 1.5(barn). Also, the bremsstrahlung dose rate is found to be 1.35 (R/h)*103, at 1.511 MeV.Additionally, the study comprehensively explored variations in the energy-absorbing buildup factor across diverse energy ranges, taking into account the Nickel concentration within the calcium ferrite nanoparticles. This aspect holds significant importance in the field of radiation dosimetry, particularly in applications related to neutron shielding and bremsstrahlung radiation.

Study on damage of Gd2O3-CeO2 under electronic energy loss: comparison between bulk-like and nanostructure.

To understand the physical phenomena responsible for radiation damage of the materials used in nuclear reactors, and thus study their operation life and/or efficiency, it is required to simulate the conditions by exposing the materials to energetic ions. Ceria (CeO2) has been proposed as one of the inert matrices for the transmutation of minor actinides in the futuristic inert matrix fuel (IMF) concept. The inert matrix should also contain burnable poison to compensate for the initial reactivity of fuel. In this context, gadolinium (Gd) is an excellent burnable poison with a high neutron absorption cross-section. In view of this, Gd2O3-CeO2 nano-powders were synthesized and sintered at 800 °C and 1300 °C to obtain different grain sizes and morphologies. FESEM and TEM were carried out to study the grain size of pristine pellets. The sintered pellets were irradiated with 80-MeV Ag ions (electronic energy loss (Se) regime) at room temperature to emulate the effect of fission fragments. For analysis of the effect of grain size on the irradiation-induced structural degradation at different fluences, GIXRD and Raman spectroscopy were performed. Significantly large damage has been observed for the smaller grain-sized samples (sintered at 800 °C) as compared to the large grain-sized sample (sintered at 1300 °C). Neither of the samples amorphized under the present experimental conditions as indicated by the presence of the Raman-active T2g mode (centred at 462 cm-1) and all the XRD peaks of fluorite cubic structure up to the highest fluence employed (1 × 1014 ions cm-2). X-ray photoelectron spectroscopy results demonstrate that Ce4+ to Ce3+ and vacancy-related isolated clusters are the main defects produced in the systems. The radiation tolerance behaviour of the samples is understood with the help of thermal spike simulation, which indicates higher transient lattice temperatures with longer duration in the smaller grain-sized sample upon irradiation. Gd-doped ceria thus possesses good radiation stability in the Se regime, indicating its potential for application in IMFs.

Density functional study on the mechanics, thermodynamics, and H diffusion mechanism of LiH

LiH is valued by researchers and engineers for its high-temperature resistance, high strength, high hydrogen storage capacity of up to 12.6 wt%, and high neutron absorption cross section. However, sintering and application of this material are difficult. Lithium hydride is a highly effective neutron shielding material. The neutron shielding ability of lithium hydride is directly affected by the homogeneity of the total amount and distribution of hydrogen. The heat migration of hydrogen reduces the neutron shielding effectiveness of lithium hydride. It is critical to forecast H's diffusion behavior. The diffusion coefficient is a critical metric for determining the diffusion efficiency. As a result, the stability, electronic structure, and bonding properties of LiH, LiOH, and H entering the tetrahedral interstitial sites of LiH were calculated using first principles based on density functional theory (DFT). The elastic modulus of the system was also examined to find the lowest thermal conductivity, and some thermodynamic parameters were derived by calculating LiH phonons. Finally, static calculations were used to determine the diffusion coefficient and diffusion activation energy of H in the LiH solid solution. The results revealed that the addition of H altered the characteristics of the LiH semiconductor, giving it metallic features. The lowest thermal conductivity of the LiH system was 2.9. W·m−1 K−1. The diffusion coefficient was D=(2.638×10−2cm2/s)exp(−5.811×10−3cal/KBT). The activation energy for diffusion was 0.252 eV. The diffusion coefficients, elastic constants, and lattice parameters all corresponded well with previously published results. The density functional theory was employed to study the material's system energy and electrical structure, and the microscopic migratory behavior of hydrogen was predicted theoretically. The calculation of the lowest thermal conductivity provided a reference for the screening of ceramic materials.

Synthesis and Thermal Oxidation Resistance of Boron-Rich Boron-Carbide Material.

A boron-rich boron-carbide material (B4+δC) was synthesized by spark plasma sintering of a ball-milled mixture of high-purity boron powder and graphitic carbon at a pressure of 7 MPa and a temperature of 1930 °C. This high-pressure, high-temperature synthesized material was recovered and characterized by X-ray diffraction, X-ray photoelectron spectroscopy, Raman spectroscopy, Vickers hardness measurements, and thermal oxidation studies. The X-ray diffraction studies revealed a single-phase rhombohedral structure (space group R-3m) with lattice parameters in hexagonal representation as a = 5.609 ± 0.007 Å and c = 12.082 ± 0.02 Å. The experimental lattice parameters result in a value of δ = 0.55, or the composition of the synthesized compound as B4.55C. The high-resolution scans of boron binding energy reveal the existence of a B-C bond at 188.5 eV. Raman spectroscopy reveals the existence of a 386 cm-1 vibrational mode representative of C-B-B linear chain formation due to excess boron in the lattice. The measured Vickers microhardness at a load of 200 gf shows a high hardness value of 33.8 ± 2.3 GPa. Thermal gravimetric studies on B4.55C were conducted at a temperature of 1300 °C in a compressed dry air environment, and its behavior is compared to other high-temperature ceramic materials such as high-entropy transition metal boride. The high neutron absorption cross section, high melting point, high mechanical strength, and thermal oxidation resistance make this material ideal for applications in extreme environments.

Synthesis and characterization of Lead–Nickel–Boron nanocomposite for radiation shielding application

We have synthesized the Lead–Nickel–Boron (LNB) nanocomposite using Green synthesis approach for the first time. LNB is synthesized by solution combustion method using Alovera as a reducing agent. The released combustion products do not affect the environment to a greater extent resulting in less air pollution compared to that of chemical synthesis. The synthesized nanocomposite is characterized using various techniques such as PXRD, SEM, UV–Vis and FTIR spectroscopy. The non-spherical shape of the nanocomposite has been confirmed through SEM images. XRD & FTIR studies reveal the physical and chemical structure of the compound. Both experimental and theoretical approach were used to study the gamma shielding ability of the nanocomposite and its efficiency is comparable to that of pure lead. The value of mass attenuation coefficient of the LNB nanocomposite is found to be 0.44, 0.36, 0.13, 0.10, 0.06, 0.06 cm2/g for 0.276, 0.356, 0.511, 0.661, 1.173 and 1.332 MeV gamma rays. Highest attenuation coefficient and lowest HVL (half value layer-the thickness of any given material where 50% of the incident energy has been attenuated) values are observed at 0.276MeV gamma ray energy. Thus proving excellent efficiency in attenuating low energy gamma rays. It is evident from the results of high neutron absorption cross section and scattering length of value 8.5fm which is comparable to that of Lead. Thus, LNB nanocomposite has also proved to be efficient in absorbing neutrons as well as Bremsstrahlung radiation and neutron shielding efficiency is better than that of lead thus acting as a good neutron absorber. Hence, LNB nanocomposite is efficient and environmental friendly material which can absorb multiple radiations like X/Gamma rays, neutrons and Bremsstrahlung radiation.

Evolution and characteristics of defect clusters in H2+ irradiated Al-based Gd2O3–W composites used as spent fuel shielding materials

Al-based composites containing high neutron absorption cross-section elements are the most promising shielding materials for spent fuel storage, but their irradiation performance is the key to their long-term service stability. In the present work, an in-situ study on the evolution and characteristics of irradiation defect clusters including dislocation loops, gas bubbles and hydrogen platelets in Al-based Gd2O3–W composites during 30 keV H2+ irradiation is carried out using a transmission electron microscope and analyzed in detail. The delamination of bubbles and the formation of a tunnel structure in the Al matrix are found for the first time. Hydrogen platelets in Al5W particles are first observed, and their formation and evolution and corresponding mechanism are investigated in detail. Platelets with different orientations have different growth rates, and the relationship between growth rate and dose is constructed. In addition, the irradiation-induced increase in yield strength in Al the matrix is estimated. This study provides new insights for the design and development of new Al-based rare earth composites with high resistance to irradiation damage.

Towards a Large-Scale “Green Energy”: Multifunctional Protective Coatings for Nuclear Engineering Structural Materials

The main task of the research is to analyze the possibility of using tungsten protective coatings in fast nuclear power reactors cooled by a lead. The author uses computational and optimization research methods. A comprehensive analysis of the properties of tungsten, including the properties of thin tungsten coatings, has been carried out. The results obtained make it possible to consider tungsten coatings of structural materials as multifunctional protective coatings. They perform the functions of protecting against corrosion and erosion of structural materials in liquid lead. They increase the thermal conductivity of fuel claddings and contribute to a decrease in the maximum fuel temperature in the nominal and emergency operating modes of the reactor. The only disadvantage of tungsten when used in nuclear technology is associated with a high neutron absorption cross section. This disadvantage can be an advantage in minimizing the void reactivity effect. The necessary neutron balance is ensured by using a coolant based on lead thorium ores and innovative ceramic fuel with the addition of uranium metal nanopowder. The significance of research is determined by the possibility of the transition of nuclear energy to the status of “green” environmentally acceptable and safe energy in the near future.

Selective thermal neutron transmutation doping with Gd masks in GaN semiconductors

For the first time, selective neutron transmutation doping was successfully performed on GaN through Gd masks, showing the feasibility of patternable doping through neutron irradiations. In collaborating with Oak Ridge National Laboratory, GaN with Gd masks was irradiated by directional neutrons. By using the Gd properties with high neutron absorption cross-sections, it was confirmed that the area covered by Gd masks on GaN was completely shielded from neutrons, and the uncovered area was doped with the generated Ge and 14C. Property differences were clearly identified in various ways when comparing the area exposed by neutrons and the area unexposed by neutrons. The discoloration by the ejected protons and elastic scattering of neutrons appeared only in the exposed area while the unexposed area remained the same as before irradiation. The aspect ratio of the patterned profile of GaN was estimated by comparing patterns on the front and backside. From energy dispersive X-ray spectroscopy spectra and secondary ion mass spectrometry measurement, the concentration difference between Ge and 14C as well as the exact concentration of produced 14C was also checked. Schottky barrier diodes on the selectively doped GaN were fabricated and investigated to study the electrical properties.

Dual X-ray- and Neutron-Shielding Properties of Gd2O3/NR Composites with Autonomous Self-Healing Capabilities.

The neutron- and X-ray-shielding, morphological, physical, mechanical, and self-healing properties were investigated for natural rubber (NR) composites containing varying gadolinium oxide (Gd2O3) contents (0, 25, 50, 75, and 100 parts per hundred parts of rubber; phr) to investigate their potential uses as self-healing and flexible neutron- and X-ray-shielding materials. Gd2O3 was selected as a radiation protective filler in this work due to its preferable properties of having relatively high neutron absorption cross-section (σabs), atomic number (Z), and density (ρ) that could potentially enhance interaction probabilities with incident radiation. The results indicated that the overall neutron-shielding and X-ray-shielding properties of the NR composites were enhanced with the addition of Gd2O3, as evidenced by considerable reductions in the half-value layer (HVL) values of the samples containing 100 phr Gd2O3 to just 1.9 mm and 1.3 mm for thermal neutrons and 60 kV X-rays, respectively. Furthermore, the results revealed that, with the increase in Gd2O3 content, the mean values (± standard deviations) of the tensile strength and elongation at break of the NR composites decreased, whereas the hardness (Shore A) increased, for which extreme values were found in the sample with 100 phr Gd2O3 (3.34 ± 0.26 MPa, 411 ± 9%, and 50 ± 1, respectively). In order to determine the self-healing properties of the NR composites, the surfaces of the cut samples were gently pressed together, and they remained in contact for 60 min; then, the self-healing properties (the recoverable strength and the %Recovery) of the self-healed samples were measured, which were in the ranges of 0.30-0.40 MPa and 3.7-9.4%, respectively, for all the samples. These findings confirmed the ability to autonomously self-heal damaged surfaces through the generation of a reversible ionic supramolecular network. In summary, the outcomes from this work suggested that the developed Gd2O3/NR composites have great potential to be utilized as effective shielding materials, with additional dual shielding and self-healing capabilities that could prolong the lifetime of the materials, reduce the associated costs of repairing or replacing damaged equipment, and enhance the safety of all users and the public.

Progress in Hexagonal Boron Nitride (h-BN)-Based Solid-State Neutron Detector

This article will briefly review the progress of h-BN based solid-state metal semiconductor metal (MSM) neutron detectors. In the last decade, several groups have been working on hexagonal boron nitride (h-BN)-based solid-state neutron detectors. Recently, the detection efficiency of 59% has been reported. Efficient, low-cost neutron detectors made from readily available materials are essential for various applications. Neutron detectors are widely used to detect fissile materials and nuclear power plants for security applications. The most common and widely used neutron detectors are 3He based, which are sometimes bulky, difficult to transport, have high absorption length, need relatively high bias voltage (>1000 V), and have low Q-value (0.764 MeV). In addition, 3He is not a readily available material. Thus, there is a strong need to find an alternative detection material. The 10B isotope has a high neutron absorption cross-section, and it has been tested as a coating on the semiconducting materials. Due to the two-step process, neutron capture through 10B and then electron–hole pair generation in a typical semiconducting material, the efficiency of these devices is not up to the mark. The progress in h-BN based detectors requires a review to envision the further improvement in this technology.

Determination of Impurities in Graphite Using Proton Induced Gamma Ray Emission, Total Reflection X-ray Fluorescence and Instrumental Neutron Activation Analysis

Determination of impurities in graphite is very important for its quality control, as their presence even at trace level can affect the performance of graphite in various applications. Graphite with equivalent boron content (EBC) less than 5 mg kg-1 is considered as nuclear grade. Elements with high neutron absorption cross section (boron and rare earths) contribute significantly to EBC. Non-destructive method is preferred as there is no sample processing and probability of loss of volatile elements while digestion. Proton Induced Gamma Ray Emission (PIGE), Instrumental Neutron Activation Analysis (INAA) were utilized for the non- destructive determination of impurities in both nuclear and commercial grade graphite. Low Z elements like Li, B, F, Na, Al and Si were detected in graphite by PIGE whereas Na, K, Sc, Cr, Mn, Fe, Co, Zn, Rb, Zr, Sb, Cs, La, Ce, Nd, Sm, Eu, Tb, Yb, Hf, Ta, Th were determined using INAA. Few elements like Ca, Ti, V, Ni, Sr and Pb remained undetected by both the non-destructive techniques. These elements were determined by Total Reflection X-ray Fluorescence (TXRF) after digestion of the graphite samples by dry ashing. Combinations of these techniques were utilized to get maximum information regarding the impurities present in graphite.

Development of an innovative external (in air) Particle Induced Gamma-ray Emission method for rapid non-destructive determination of isotopic composition of boron in “As received” boron based ceramic neutron absorbers

Boron is an important element in nuclear reactor technology due to its high neutron absorption cross section of 10B isotope. Isotopic composition of B (IC, 10B/11B atom ratio) determination in finished neutron absorbers is a necessity under chemical quality control (CQC). We report an innovative greener method for rapid and non-destructive approach of isotopic composition determination of B in “as received” boron based ceramic neutron absorbers including boron carbides and hexa-borides by external (in air) Particle Induced Gamma-ray Emission (PIGE) using 3.5 MeV proton beam. It involves irradiation of “as received” powder samples wrapped in a thin Mylar film and measurement of prompt gamma rays at 429, 718 and 2125 keV from 10B(p,αγ)7Be, 10B(p,p’γ)10B and 11B(p,p’γ)11B, respectively, using a HPGe detector system. The method was standardized with natural and enriched B4C powders. For validation, the results of isotopic composition obtained from “as received” samples were compared with that obtained from pellet samples using both external and vacuum chamber PIGE methods. IC values obtained for natural to 10B enriched samples (19.8–67 atom % of 10B) are very encouraging with 1–2% and 0.3–0.7% uncertainties from single and replicate sample experiments. The method is truly non-destructive as the samples can be returned back as such after the experiment as they are not radioactive. Compared to existing PIGE method for isotopic composition of B, the developed method keeps promise for wide applications as it is simple, sensitive and rapid and it does not require vacuum, pellet preparation with a binder, exact mass of the sample and beam current measurement.

Neutron imaging based on transfer foil activation and COTS CMOS image sensors

In this paper we present a method for obtention of neutron images with Commercial-Off-The-Shelf (COTS) CMOS image sensors through the activation of indium foils. This detection method has been designed specifically for the acquisition of thermal and epitermal neutron images in mixed beams with a high gamma flux, and also for the study of high radioactive samples that are usually placed into research reactor pools. We also present a technique to obtain multi-spectral neutron images taking advantage of the high neutron absorption cross-section of this material in the thermal energy range, as well as around the 1.45 eV resonance. Measurements performed in a neutron beam of the RA6 nuclear research reactor located in Bariloche, Argentina, confirm the capability of the proposed technique.

The calculation of self-shielding correction factors for large samples in 241Am–Be isotopic neutron source

In the present study, it is aimed to calculate neutron self-shielding correction factors for cylinder sample geometry using different methods in the literature in the application of k0-NAA standardization method with 241Am–Be isotopic neutron source. These correction factors are important in calculating the accurate concentration values for large samples with a high neutron absorption cross-section in the application of the k0-NAA method. In order to see the validity of the applied theoretical and experimental methods, the obtained concentration values are given comparatively with respect to the known Mn concentrations. The mean concentration difference ranges from −3.4% to −9.7%. The thermal neutron self-shielding correction depends on all elements in the sample that absorb thermal neutrons. In k0-NAA applications with large samples, although time-consuming, it is recommended to use the experimental methods after calculating the necessary parameters with the help of samples with known composition.

Performance and economic assessment of enriched gadolinia burnable absorbers

As a burnable absorber, gadolinium oxide (Gd2O3) is widely used in light water reactors due to the high neutron absorption cross section of several gadolinium isotopes and its good solid solubility in UO2. However, some isotopes of natural Gd cause residual reactivity suppression, while some are not efficient neutron absorbers, reducing the efficiency of the burnable absorber when implemented. In this study, fuel assemblies utilizing gadolinium oxide enriched with 157Gd isotope were modelled using Monte Carlo particle transport methods and compared to fuel with a natural Gd based absorber. Reactivity gains were examined over the life of the assembly utilizing 157Gd-enriched absorber as compared to natural gadolinia. A preliminary economic evaluation is also made to assess the commercial benefits of using 157Gd-enriched burnable absorber. Use of enriched gadolinium oxide is shown to eliminate residual reactivity caused by natural gadolinium oxide, and similar reactivity properties (and therefore criticality margins) can be achieved with less burnable absorber in the fuel. The financial cost of incorporating enriched Gd isotopes into nuclear fuel has also been estimated.

Synthesis and Characterization of a Carborane-containing Precursor for B<sub>4</sub>C Ceramics

B₄C ceramic is one of widely used ceramics with properties of low density, high hardness, high thermal-stability and high neutron absorption cross-section. However, the high melting point and high hardness of B₄C make it difficult to be prepared as materials with complex shape. Precursor derived method is an effective approach to prepare special-shaped ceramics. In the early research of B₄C ceramic precursors, highly toxic and unstable decaborane was generally used as the boron source material, and precursors were also unstable and difficult to synthesize. In this work, a new B₄C precursor is designed and synthesized, and m-carborane, which has stable chemical properties and high thermal stability, is used as the boron source. Firstly, m-carborane is converted into lithium salt after removing active hydrogen by n-butyllithium. Then, catalyzed by Cu(I), the lithium salt reacts with m-chloroiodobenzene so that monomer, bis(3-chlorophenyl)-m-carborane, is prepared. The monomer is polymerized through Ullmann Coupling reaction at the present of Ni(0) catalyst and Zinc powder, and precursor poly[bis(3-chlorophenyl)-m-carborane] is finally synthesized. The precursor is stable at room temperature in air and soluble in Tetrahydrofuran, with Mw of 1013g/mol, and ceramic yield of 62.20% at 900°C in N₂. The precursor pyrolytic product is composed of B₄C and graphite after heat treated at 1600°C in inert atmosphere.

Computational and experimental comparison of boron carbide, gadolinium oxide, samarium oxide, and graphene platelets as additives for a neutron shield

Neutron shielding materials are used in aerospace, space, medical, and nuclear applications and there is a growing need of neutron shielding materials with increased temperature resistance. This is difficult to achieve with the current approach of using borated or lithium enriched polyethylene or polyamide composites because their hydrogen content limits temperature resistance to 200 °C. To address this challenge, an elastomer-based neutron shielding material has been developed with a degradation temperature greater than 300 °C and a macroscopic neutron cross section better than that of 5% borated polyethylene. By compensating for the low number of hydrogen atoms in this material, compared to polyethylene or polyamides, iron and additives with a high neutron absorption cross section such as gadolinium and samarium are incorporated to achieve the necessary attenuation properties. Graphene platelets were also added to evaluate graphene's ability to mitigate secondary gamma from the neutron absorption by gadolinium and samarium. A computational study was first completed in Monte Carlo N-Particle (MCNP) to benchmark existing materials and compare the loading of the potential additives. The maximum experimental fast neutron cross-sections for each absorber additive were 0.021 cm−1 in 9% B4C, 0.031 cm−1 with 27% Gd2O3, and 0.030 cm−1 with 3% Sm2O3 compared to only 0.023 cm−1for 5% borated polyethylene. The sample containing 10% Gd2O3 and 2% B4C attenuated neutrons better than the borated polyethylene with a fast neutron cross section of 0.026 cm−1 and exhibited the lowest amount of secondary gamma photons of all the material combinations.

Assembly-level analyses of accident-tolerant cladding concepts for a long-life civil marine SMR core using micro-heterogeneous duplex fuel

In this reactor physics study, we examine the neutronic performance of accident-tolerant fuel (ATF) claddings – austenitic type 310 stainless steel (310SS), ferritic Fe-20Cr-5Al (FeCrAl), advanced powder metallurgic ferritic (APMT), and silicon carbide (SiC)-based materials – as alternative cladding materials compared with Zircaloy-4 (Zr) cladding. The cores considered use 18% 235U enriched micro-heterogeneous ThO2-UO2 duplex fuel and, for purposes of comparison, 15% 235U enriched homogeneously mixed all-UO2 fuel, loaded into 13×13 pin arrays. A constant cladding coating thickness of 655 μm is assumed. We use the WIMS reactor physics code to analyse the associated reactivity, achievable discharge burnup, spectral variations, rim effect and reactivity feedback parameters for the candidate cladding materials at the assembly level.The results show that candidate fuels with 310SS cladding exhibit a ∼13% discharge burnup penalty compared to Zr due to the presence of a very high nickel (Ni) concentration. The high neutron absorption cross-sections of iron (Fe) in the FeCrAl and APMT claddings also lead to a ∼10% discharge burnup penalty. The fuels with SiC cladding can achieve a ∼1% higher discharge burnup compared to Zr due to the low thermal neutron absorption cross-sections of its constituents and the softer neutron spectrum. The claddings with lower capture cross-sections (SiC and Zr) exhibit higher relative fission power at the pellet periphery. For both candidate fuels, the end-of-life 239Pu (for UO2 fuel) and 233U (for duplex fuel) inventories are higher for the claddings (Fe-based: FeCrAl, APMT and steel-based: 310SS) with higher thermal capture cross-sections, unlike for SiC and Zr, where SiC provides higher end-of-life 239Pu and 233U inventories despite having lower capture cross-section than that of the Zr. Reactivity feedback parameter values (moderator and fuel temperature coefficients) are more negative for the duplex fuel than the UO2 fuel for all the candidate claddings, with claddings with harder spectra exhibiting more negative values. The duplex fuel yields a softer spectrum than the UO2 fuel with the candidate claddings, which improves neutron economy and thus discharge burnup.

Results from the dynamic albedo of neutrons (DAN) passive mode experiment: Yellowknife Bay to Amargosa Valley (Sols 201–753)

The Mars Science Laboratory (Curiosity rover) Dynamic Albedo of Neutrons (DAN) experiment detects neutrons for the purpose of searching for hydrogen in the shallow subsurface of Mars. DAN has two modes of operation, active and passive. In passive mode, the instrument detects neutrons produced by Galactic Cosmic Ray interactions in the atmosphere and regolith and by the rover's Multi-Mission Radioisotope Thermoelectric Generator. DAN passive data from Yellowknife Bay to Amargosa Valley (sols 201 through 753) are presented and analyzed here. Water equivalent hydrogen (WEH) estimates from this portion of Curiosity's traverse range from 0.0 wt. % up to 15.3 wt. %. Typical uncertainties on these WEH estimates are ∼0.5 wt. % but in some cases can be as high as ∼4.0 wt. % depending on the specific circumstances of a given measurement. Here we also present a new way of reporting results from the passive mode of the experiment, the DAN passive geochemical index (DPGI). This index is sensitive to some key geochemical variations, but it does not require assumptions about the abundances of high thermal neutron absorption cross section elements, which are needed to estimate WEH. DPGI variations in this section of the traverse indicate that the shallow regolith composition is changing on both the local (∼meters) and regional (∼100 s of meters) scales. This variability is thought to be representative of the diverse composition of source regions for sediments within the crater floor. Kolmogorov-Smirnov Tests on the populations of WEH estimates and DPGI values demonstrate there are statistically significant differences between nearly all of the geologic units investigated along the rover's traverse. We also present updated previous DAN passive results from Bradbury Landing to John Klein that make use of revised DAN active mode results for calibration, however, no qualitative changes in the interpretations made in Tate et al. (2015b) are incurred.

Removal of gadolinium, a neutron poison from the moderator system of nuclear reactors

Gadolinium as gadolinium nitrate is used as neutron poison in the moderator system for regulating and controlling the power generation of Pressurized Heavy Water Reactors (PHWR) and proposed to be used in Advanced Heavy Water Reactors (AHWR) owing to its high neutron absorption cross section. Removal of the added gadolinium nitrate (Gd3+ and NO3−) from the system after its intended use is done using ion exchange resins. In the present investigation, attempts have been made to optimize the ion exchange process for generation of low radioactive waste and maximize utilization of the ion exchange resins by employing different types of resins and different modes of operation. The investigations revealed that use of mixed bed (MB) resin column consisting of Strong Acid Cation (SAC) resin and Strong Base Anion (SBA) resin followed by SAC resin column is efficient in removing the Gd3+ and NO3− from the system besides maintaining the pH of the moderator system in the desirable regime, where gadolinium does not get precipitated as its hydroxide.