Thermal Neutron Irradiation Research Articles

Therapeutic Effect of Boron Neutron Capture Therapy on Boronophenylalanine Administration via Cerebrospinal Fluid Circulation in Glioma Rat Models.

In recent years, various drug delivery systems circumventing the blood-brain barrier have emerged for treating brain tumors. This study aimed to improve the efficacy of brain tumor treatment in boron neutron capture therapy (BNCT) using cerebrospinal fluid (CSF) circulation to deliver boronophenylalanine (BPA) to targeted tumors. Previous experiments have demonstrated that boron accumulation in the brain cells of normal rats remains comparable to that after intravenous (IV) administration, despite BPA being administered via CSF at significantly lower doses (approximately 1/90 of IV doses). Based on these findings, BNCT was conducted on glioma model rats at the Kyoto University Research Reactor Institute (KUR), with BPA administered via CSF. This method involved implanting C6 cells into the brains of 8-week-old Wistar rats, followed by administering BPA and neutron irradiation after a 10-day period. In this study, the rats were divided into four groups: one receiving CSF administration, another receiving IV administration, and two control groups without BPA administration, with one subjected to neutron irradiation and the other not. In the CSF administration group, BPA was infused from the cisterna magna at 8 mg/kg/h for 2 h, while in the IV administration group, BPA was intravenously administered at 350 mg/kg via the tail vein over 1.5 h. Thermal neutron irradiation (5 MW) for 20 min, with an average fluence of 3.8 × 1012/cm2, was conducted at KUR's heavy water neutron irradiation facility. Subsequently, all of the rats were monitored under identical conditions for 7 days, with pre- and post-irradiation tumor size assessed through MRI and pathological examination. The results indicate a remarkable therapeutic efficacy in both BPA-administered groups (CSF and IV). Notably, the rats treated with CSF administration exhibited diminished BPA accumulation in normal tissue compared to those treated with IV administration, alongside maintaining excellent overall health. Thus, CSF-based BPA administration holds promise as a novel drug delivery mechanism in BNCT.

Photon and thermal neutron shielding behaviors of aluminum calcium fluoroborate glass modified with barium oxide: FLUKA Monte Carlo, XCOM and experimental investigations

The borate glass system was prepared using the melt-quenching method. The XRD was used to confirm the amorphous structure of present samples, the FTIR was utilized to observe vibration modes, and the UV–Vis-NIR spectrophotometer was employed to measure transmittance spectra. The radiation shielding properties were examined by FLUKA Monte Carlo simulation, XCOM programs, and experimental methods. The density increased, while the molar volume decreased as the BaO content increased. For the X-ray and gamma-ray radiation, indicate that the addition of BaO instead of B2O3 improve the shielding properties, while the Σ is decreased for thermal neutron radiation. According to the results, adding BaO enhanced the X-ray and gamma-ray shielding properties. However, adding BaO led to the reduces its ability of shielding the thermal neutrons of the glass samples. These data are expected to be accommodate to considering for any applications which requires high efficiency of photon or thermal neutrons shielding.

Feasibility of producing 225Ac via thermal neutron irradiation of 226Ra: A systematic theoretical study

With a suitable half-life and abundant radiolabeling strategy, 225Ac has become one of the most promising radionuclides in the area of targeted alpha therapy. However, limited radionuclide supply is threatening the development of 225Ac related endoradiotherapy dramatically. As the parent nuclide of 225Ac, 229Th can be produced via 226Ra(3n, 2β)229Th reaction in a nuclear reactor. However, related practice has not been conducted in large scale, since the nuclear reaction pathway for producing 229Th is complicated. In this work, the feasibility of producing 225Ac/229Th in a reactor was confirmed by systematic theoretical calculations, and a procedure that combines irradiation with separation process was proposed. The results show that 176 MBq of 229Th can be produced by irradiating 1.0 g of 226Ra with a neutron flux density of 1 × 1015 n cm−2 s−1 for 90 days. This will generate 150 MBq of 225Ac monthly from a radionuclide generator, which is sufficient for the single treatment cycle of 200 patients each year considering the radioactivity loss in radiochemical separation, transfer and radiolabeling process. In addition, this irradiation process will also produce 37.8 GBq 227Ac for the preparation of 227Ac-227Th-223Ra generator. In general, the production of 225Ac by neutron irradiation of 226Ra in reactor is practicable and holds potential to alleviate the shortage of current supply of 225Ac.

Measurement of potassium in rats using an In vivo neutron activation analysis system

Abnormal levels of potassium are linked to several health conditions, including high blood pressure, cardiac dysfunction, kidney damage, and osteoporosis. Given the limited availability of in vivo measurement techniques, there is a need for novel methods to measure potassium to enhance the diagnosis and management of potassium metabolism related diseases. This study aimed to evaluate the feasibility of compact neutron generator based in vivo measurement system for quantification of potassium using rat carcasses. A cohort of thirty-nine rats (n = 20 males and 19 females, average weight 255 ± 15 and 163 ± 7 g) were sacrificed, and their carcasses were placed in polyethylene bottles. The rats were then positioned and irradiated in a carefully designed irradiation cave built alongside the neutron generator with an optimized thermal neutron flux and radiation dose ratio. The irradiation time was 10 min, followed by a 5-min decay and 2-h measurement using a high efficiency high purity germanium detector(HPGe). ResultsThe average potassium concentration in male and female rats was found to be comparable (male 2874 ± 161 and female 2866 ± 144 μg/g). A marginally positive correlation between potassium concentration and weight was found in female rats only (male(20) = 0.07, P = 0.76 and female r(19) = 0.34, P = 0.15). We assessed the influence of manganese toxicity on potassium levels and observed no significant impact. These results were consistent with our previous study in mice. ConclusionThis study suggests that in vivo neutron activation analysis could serve as a promising method to quantify potassium and to investigate the storage and metabolism of potassium in human and in animals.

Radiation shielding parameterizations of FeSO4, CuSO4, NiSO4 and ZnSO4 Compounds: using (XRF) technique & Monte Carlo FLUKA approach

The mass attenuation coefficient (MAC) for different sulphate compounds can be estimated by using the Energy Dispersive X-ray Fluorescence (EDXRF), also known as X-ray Fluorescence (XRF) technique. The X-ray photons emitted have different energies depending on incident photon energy, atomic weight, and molecular structure of tested material. The excitation energy of the gamma rays source with 59.53 keV was obtained by using 241Am (40 µci). The (MAC) for sulphate compounds of a different element (Fe, Ni, Cu, Zn) have been calculated by measuring the intensity difference for kα in pure elements and their compounds. The determined results showed that the maximum value for (µm) was in FeSO4. These results are consistent with the theoretical value obtained by the XCOM software in addition to investigating the wide energy response of photon interaction with the introduced compounds using the FLUKA Monte Carlo simulation software. The mass attenuation coefficient (MAC) of these compounds is numerically evaluated in the energy range 0.015-15 MeV using the FLAIR code. The computed (µm) is used to generate significant radiation protection factors such as the linear attenuation coefficient (LAC), half-layer value (HVL), effective (Zeff), and equivalent (Zeq) atomic number. For studying the shielding effectiveness and efficiency, for fast and thermal neutron radiation, the removal cross-section ƩR was given.

Albumin nanoparticle electrostatically loaded with highly anionic Gd-thiacalixarene complex for MRI-guided neutron capture therapy

As a neutron capture therapy (NCT) agent, 157Gd garners significant attention due to the largest neutron capture cross section, emission of γ rays and Auger electrons, and facilitation of magnetic resonance imaging-guided NCT. Owing to its high kinetic stability, 1H relaxivity, and facile conjugation to albumin via electrostatic interactions, the negatively charged Gd3TCAS2 complex {TCAS = thiacalix[4]arene-p-tetrasulfonate (TCAS)} was incorporated into albumin nanoparticles (ANPs) in various configurations, including shell, core, and core-shell types, contingent on the positioning of Gd3TCAS2 within the ANP. The characteristics of the ANPs optimized with respect to loading capacity (LC) were explored, all of which demonstrated particle sizes suitable for passive delivery. The LC of Gd3TCAS2 per ANP increased in the following order: shell (1.1 %) < core (1.4 %) < core-shell (8.1 %). The leakage of Gd3TCAS2 after 7 d of storage was approximately 2 % for all ANPs, indicating stable Gd3TCAS2 loading. The 1H relaxivity (e.g., core-shell: r1 = 10.7 mM–1s–1) surpassed that of the Gd3TCAS2 complex alone (3.15 mM–1s–1). The luminescence emanating from Tb in MCF-7 cells co-cultured with Tb3TCAS2-loaded ANPs corroborated this cellular uptake. Gd uptake into MCF-7 cells via Gd3TCAS2-loaded ANPs was most pronounced for core-shell ANP at 1.33 ± 0.06 nmol/106 cells. The cytotoxicity of Gd3TCAS2-loaded ANPs was mitigated compared to free Gd3TCAS2 (CC50 = 52 µM), with shell and core-shell ANPs demonstrating nearly 100 % cell survival even at 50 µM. Cell viability post thermal neutron irradiation revealed the highest NCT efficacy for free Gd3TCAS2, attributed to its elevated cellular uptake (15.71 ± 0.57 nmol/106 cells). Core-shell ANP also exhibited a notable reduction in cell viability relative to the control, validating their potential as a promising GdNCT agent.

Neutron detection properties of Ti-doped LiGaO2 single crystalline scintillator

Photoluminescence (PL) and scintillation properties of undoped and Ti-doped LiGaO2 single crystals were investigated. In both PL and scintillation, the Ti-doped LiGaO2 crystals showed luminescence due to charge transfer between Ti4+–O2- while the undoped LiGaO2 crystal showed luminescence due to oxygen vacancies. Under thermal neutron irradiation, all the LiGaO2 crystals exhibited a clear thermal neutron peak, and the highest light yield was observed from the 3% Ti-doped LiGaO2 crystal. The light yield of the 3% Ti-doped LiGaO2 crystal was ten times higher than that of the undoped LiGaO2 crystal and comparable with that of a commercial neutron scintillator GS20. In addition, the Ti-doped LiGaO2 crystals showed high α/γ ratios up to about 0.7.

90Y]Yttria Alumino Silicate Glass Microspheres: A Biosimilar Formulation to "TheraSphere" for Cost-Effective Treatment of Liver Cancer.

Background: Selective internal radiation therapy (SIRT) using a suitable β--emitting radionuclide is a promising treatment modality for unresectable liver carcinoma. Yttrium-90 (90Y) [T1/2 = 64.2 h, Eβ(max) = 2.28 MeV, no detectable γ-photon] is the most preferred radioisotope for SIRT owing to its favorable decay characteristics. Objective: The present study describes indigenous development and evaluation of intrinsically radiolabeled [90Y]yttria alumino silicate ([90Y]YAS) glass microsphere, a formulation biosimilar to "TheraSphere" (commercially available, U.S. FDA-approved formulation), for SIRT of unresectable liver carcinoma in human patients. Methods: YAS glass microspheres of composition 40Y2O3-20Al2O3-40SiO2 (w/w) and diameter ranging between 20 and 36 μm were synthesized with almost 100% conversion efficiency and >99% sphericity. Intrinsically labeled [90Y]YAS glass microspheres were produced by thermal neutron irradiation of cold YAS glass microspheres in a research reactor. Subsequent to in vitro evaluations and in vivo studies in healthy Wistar rats, customized doses of [90Y]YAS glass microspheres were administered in human patients. Results: [90Y]YAS glass microspheres were produced with 137.7 ± 8.6 MBq/mg YAS glass (∼6800 Bq per microsphere) specific activity and 99.94% ± 0.02% radionuclidic purity at the end of irradiation. The formulation exhibited excellent in vitro stability in human serum and showed >97% retention in the liver up to 7 d post-administration when biodistribution studies were carried out in healthy Wistar rats. Yttrium-90 positron emission tomography scans recorded at different time points post-administration of customized dose of [90Y]YAS glass microspheres in human patients showed near-quantitative retention of the formulation in the injected lobe. Conclusions: The study confirmed the suitability of indigenously prepared [90Y]YAS glass microspheres for clinical use in the treatment of unresectable hepatocellular carcinoma.

Scattering-free Ce:LYBO single crystals for thermal neutron detection

Ce:Li6Y(BO3)3 (LYBO) is a well-known candidate for thermal neutron detection with a very high Li concentration (3.06 × 1022/cm3). So far, as-grown crystals exhibit a milky appearance that compromises their performance as scintillators. Current work demonstrates, for the first time, the growth of scattering-free undoped and Ce-doped LYBO by a thermal quenching process. The origin and features of the scattering centers are investigated in detail. Furthermore, the annealing treatment for the scintillation activation is studied, finding that the reduction in oxygen vacancies is mandatory. Under thermal neutron irradiation, the annealed scattering-free Ce:LYBO single crystal achieves a record-high light yield of 6200 ph/n in a single decay with a lifetime as short as 24 ns.

Comparison of H*(10) estimations, due to neutrons and γ-rays, around FANT with two 241Am/9Be sources

FANT (Fuente Ampliada de Neutrones Térmicos; in Spanish) is a thermal neutron irradiation facility with an extended and very uniform irradiation area, which has been developed by the Neutron Measurement Laboratory of the Energy Engineering Department at Universidad Politécnica de Madrid (LMN-UPM). In FANT, an isotopic neutron source (241Am/9Be) produces the primary neutrons. The design and facility optimization were carried out by extensive Monte Carlo calculations. In addition, Monte Carlo methods were used to evaluate the facility's performance to produce a constant and uniform thermal neutron field; these results were validated through experimental methods. FANT is designed to have two neutron sources; the objective of this work is to estimate the ambient dose equivalent due to neutrons and gamma-rays by Monte Carlo methods, and to compare these values with measured experimental doses. Thus, the performance of FANT with the two 241Am/9Be sources of LMN-UPM, with regard to the ambient dose equivalent H*(10) produced by both neutrons and photons around the facility, is analyzed in this work. The results are compared with those previously obtained in the framework of the results obtained with the LB6411 device around FANT.

Dissolution behavior and aging of iron–uranium oxide

Understanding the dissolution behavior of the fuel debris is necessary for the safe decommissioning of Fukushima Daiichi Nuclear Power Plants. The dissolution behavior of FeUO4 compounds formed by a high-temperature reaction of UO2 with iron, a stainless-steel component of reactor structural materials, was investigated under atmospheric conditions. The compounds were prepared in an electric furnace using U3O8 and Fe3O4 as starting materials, and their solid states were analyzed using X-ray diffraction, scanning electron microscopy–energy-dispersive X-ray spectroscopy, and X-ray absorption fine-structure spectroscopy. The fission products were produced via thermal-neutron irradiation. The concentration of nuclides dissolved in water was examined by performing static leaching tests of FeUO4 compounds for up to three months. A redox reaction was proposed to occur between trivalent Fe and pentavalent U ions in the early stage of FeUO4 dissolution. It was thermodynamically deduced that the reduced divalent Fe ion was finally oxidized into a trivalent ion in the presence of dissolved oxygen, and iron hydroxide limited the solubility of Fe. Meanwhile, the concentration of hexavalent U (i.e., uranyl ion) was limited owing to the presence of secondary minerals such as metaschoepite and sodium uranate and subsequently decreased, possibly owing to sorption on Fe oxides, for example. The concentrations of multivalent ions of fission products, such as Ru and Ce, also decreased, likely for the reason above. By contrast, the concentration of soluble Cs ions did not decrease. The validity of this interpretation was supported by comparing the results with the dissolution behavior of a reference sample (Fe-free U3O8).

HER-2-Targeted Boron Neutron Capture Therapy with Carborane-integrated Immunoliposomes Prepared via an Exchanging Reaction.

Boron neutron capture therapy (BNCT) is a promising modality for cancer treatment because of its minimal invasiveness. To maximize the therapeutic benefits of BNCT, the development of efficient platforms for the delivery of boron agents is indispensable. Here, carborane-integrated immunoliposomes were prepared via an exchanging reaction to achieve HER-2-targeted BNCT. The conjugation of an anti-HER-2 antibody to carborane-integrated liposomes successfully endowed these liposomes with targeting properties toward HER-2-overexpressing human ovarian cancer cells (SK-OV3); the resulting BNCT activity toward SK-OV3 cells obtained using the current immunoliposomal system was 14-fold that of the l-BPA/fructose complex, which is a clinically available boron agent. Moreover, the growth of spheroids treated with this system followed by thermal neutron irradiation was significantly suppressed compared with treatment with the l-BPA/fructose complex.

Gamma, neutron, and charged particle shielding performance of ABKT glass system

In this study, the gamma, charged, thermal neutron, and fast neutron radiation shielding parameters of (5+x)K2O-(25-x)TeO2-(10-x)As2O3-(60-x)B2O3 (where x ranged from 0 (ABKT1) to 20 (ABKT5) mol% in 5 mol% steps) were calculated and simulated. In addition, important optical parameters such as the reflection loss (Rloss), molar refractivity (Rm), molar polarizability (αm), optical transmission (TOpt), metallization criterion (M(n)), static dielectric constant (Ɛsta), optical dielectric constant (Ɛopt) of the glasses, were computed and juxtaposed with the chemical composition of the glasses. The values of Rm and αm decreased from 17.213 cm3/mol to 15.247 cm3/mol and 6.830 x10−24 cm3 to 6.05 x10−24 cm3, respectively, as K2O replaced TeO2 in the ABTK glass system. Similarly, Rloss reduced from 0.173 to 0.164, while TOpt increased from 0.705 to 0.719. The values of M(n), εsta, and εOpt were in the range 0.381–0.397, 5.565–5.87, and 4.565–4.871, respectively. The mass attenuation coefficients of ABKT1–ABKT5 were within the range of 0.0276–30.8789, 0.0269–29.6485, 0.0262–28.3373, 0.0255–29.9370, and 0.0249–25.4384 cm2/g, respectively. ABKT1 is the best gamma photon absorber among the ABKT glasses for narrow and broad beam transmission modes. For electrons, protons, alpha particles, and heavy carbon ions, the mass stopping powers of the glasses follow the order ABKT1>ABKT2>ABKT3>ABKT4>ABKT5. ABKT1 had the best photon and ion absorbing capabilities, while ABKT5 has the highest thermal and fast neutron interaction cross-sections. The ABKT1 glass demonstrated the potential to perform better than some commercial glass shields, standard shielding glasses, light and heavy concretes, and a host of other materials as a transparent gamma radiation shielding material.

Development of a Gadolinium-Boron-Conjugated Albumin for MRI-Guided Neutron Capture Therapy.

Noninvasive monitoring of boron agent biodistribution is required in advance of neutron capture therapy. In this study, we developed a gadolinium-boron-conjugated albumin (Gd-MID-BSA) for MRI-guided neutron capture therapy. Gd-MID-BSA was prepared by labeling bovine serum albumin with a maleimide-functionalized gadolinium complex and a maleimide-functionalized closo-dodecaborate orthogonally. The accumulation of Gd-MID-BSA in tumors in CT26 tumor-bearing mice reached a maximum at 24 h after the injection, as confirmed by T1-based MRI and biodistribution analysis using inductively coupled plasma optical emission spectrometry. The concentrations of boron and gadolinium in the tumors exceeded the thresholds required for boron neutron capture therapy (BNCT) and gadolinium neutron capture therapy (GdNCT), respectively. The boron concentration ratios of tumor to blood and tumor to normal tissues satisfied the clinical criteria, indicating the reduction of undesired nuclear reactions of endogenous nuclei. The molar ratio of boron to gadolinium in the tumor was close to that of Gd-MID-BSA, demonstrating that the accumulation of Gd-MID-BSA in the tumor can be evaluated by MRI. Thermal neutron irradiation with Gd-MID-BSA resulted in significant suppression of tumor growth compared to the group injected with a boron-conjugated albumin without gadolinium (MID-BSA). The neutron irradiation with Gd-MID-BSA did not cause apparent side effects. These results demonstrate that the conjugation of gadolinium and boron within the albumin molecule offers a novel strategy for enhancing the therapeutic effect of BNCT and the potential of MRI-guided neutron capture therapy as a promising treatment for malignant tumors.

A KGd(WO4)2 powder/Er2Ce2O7–SiO2 aerogel with low thermal conductivity and excellent radiation shielding performance for thermal insulation of nuclear power plants

To meet the application requirements of thermal insulation materials for nuclear power plants, Rare earth composite aerogel (KGd(WO4)2 powder/Er2Ce2O7–SiO2) were prepared by compounding rare earth precursor sol and rare earth powder with silica aerogel based on directional freeze drying method. The effect of thermal diffusivity of rare earth salt crystals on the thermal conductivity of rare earth composite aerogel was found, and the mechanism of the effect of the type and morphology of rare earth powder on the γ-ray shielding performance was explored. The irradiation and temperature resistance measurement shows that the KGd(WO4)2 powder/Er2Ce2O7–SiO2 aerogel can withstand 1 × 107 Gy γ-ray and high temperature baking above 1100 °C. KGd(WO4)2 powder/Er2Ce2O7–SiO2 aerogel is an ideal new generation multi-functional shielding and thermal insulation material for nuclear power plants because of its low thermal conductivity, excellent γ-ray and neutron radiation shielding, high temperature resistance and radiation aging resistance.

Rational Design, Multistep Synthesis and in Vitro Evaluation of Poly(glycerol) Functionalized Nanodiamond Conjugated with Boron-10 Cluster and Active Targeting Moiety for Boron Neutron Capture Therapy.

Boron neutron capture therapy (BNCT), advanced cancer treatment utilizing nuclear fission of 10 B atom in cancer cells, is attracting increasing attention. As 10 B delivery agent, sodium borocaptate (10 BSH, 10 B12 H11 SH ⋅ 2Na), has been used in clinical studies along with L-boronophenylalanine. Recently, this boron cluster has been conjugated with lipids, polymers or nanoparticles to increase selectivity to and retentivity in tumor. In this work, anticancer nanoformulations for BNCT are designed, consisting of poly(glycerol) functionalized detonation nanodiamonds (DND-PG) as a hydrophilic nanocarrier, the boron cluster moiety (10 B12 H11 2- ) as a dense boron-10 source, and phenylboronic acid or RGD peptide as an active targeting moiety. Some hydroxy groups in PG were oxidized to carboxy groups (DND-PG-COOH) to conjugate the active targeting moiety. Some hydroxy groups in DND-PG-COOH were then transformed to azide to conjugate 10 B12 H11 2- through click chemistry. The nanodrugs were evaluated in vitro using B16 murine melanoma cells in terms of cell viability, BNCT efficacy and cellular uptake. As a result, the 10 B12 H11 2- moiety is found to facilitate cellular uptake probably due to its negative charge. Upon thermal neutron irradiation, the nanodrugs with 10 B12 H11 2- moiety exhibited good anticancer efficacies with slight differences with and without targeting moiety.

Efficient neutron capture therapy of glioblastoma with pteroyl-closo-dodecaborate-conjugated 4-(p-iodophenyl)butyric acid (PBC-IP).

Boron neutron capture therapy (BNCT) has been applied for clinical trials on glioblastoma patients since 1950s, however, the low survival rate under the treatments has hampered the widespread use of BNCT. In this study, we developed a novel boron agent, PBC-IP, which consists of three functional groups: FRα-targeting, 10B resource (twelve 10B atoms in the molecule), and albumin-binding moieties. PBC-IP was selectively taken up by glioma cell lines such as C6, F98, and U87MG cells and accumulated 10- to 20-fold higher than L-4‑boronophenylalanine (BPA). PBC-IP administrated intravenously to the human glioblastoma (U87MG) xenograft model showed higher boron accumulation in tumors (29.8 μg [10B]/g at 6 h) than BPA (9.6 μg [10B]/g at 3 h) at a 25 mg [10B]/kg dose, effectively suppressing tumor growth after thermal neutron irradiation. PBC-IP administrated via convection-enhanced delivery (CED) accumulated in the F98 glioma orthotopic rat model, achieving 26.5 μg [10B]/g in tumors with tumor/normal (T/N) brain and tumor/blood (T/B) boron ratios of 37.8 and 94.6, respectively, 3 h after CED. Survival at 180 days after BNCT was 50% in the PBC-IP group and 70% in the combined BPA and PBC-IP groups, with no residual brain tumors.

Radiation characterizations of two isotopic neutron sources merging in one irradiator for experimental applications in the laboratory

Neutron sources are utilized for different aims, such as studying the material’s internal structure, investigating the materials’ crystal structures, and neutron therapy. Using more than one isotopic neutron source in some aims could be beneficial. Herein, a neutron irradiator consisting of two 241Am-9Be neutron sources has been designed. Radiation characterizations, including the neutron fluence and dose rate, were investigated. The Monte Carlo simulation code Monte Carlo N-particle transport code version 5 (MCNP5) was used to conduct all the required investigations to achieve an optimal design. Two factors were considered; the first was the dose rate due to both neutron and gamma radiations to meet radiation protection requirements; MCNP5 was used to determine the adequate thickness of the shielding material. The second parameter is the neutron fluence. Two irradiation sites have been proposed, one for fast neutron irradiation and the other for thermal neutron irradiation. By utilizing MCNP5, we have determined the radiation characterization at the two sites to be used as the irradiation sites. After the optimal design had been produced, the thermal and the epithermal neutron fluences inside the two irradiation sites were determined experimentally. The experimental results showed a perfect agreement with the simulation ones. The neutron irradiator designed from merged two isotopic neutron sources could be a benefit in the neutron radiation applications in the labs.

Thermoluminescent properties of Tb3+-doped B2O3–Na2O–CaO–P2O5 glasses for neutron detection

We fabricated Tb3+-doped B2O3–Na2O–CaO–P2O5 glasses using 10B- or 11B-enriched raw materials to develop new thermoluminescent (TL) glass materials for neutron detection. We investigated TL properties of the glass materials after X-ray and thermal neutron irradiation. A TL glow peak was observed at 376 K after X-ray irradiation, and the integrated TL intensities of the 10B- and 11B-enriched glasses were similar. This result indicates that the 10B- and 11B-enriched glasses had similar sensitivities to X-rays. A broad peak was observed in the 350–650 K range for the 10B- and 11B-enriched glasses after neutron irradiation, and the 10B-enriched glass exhibited higher TL intensities than the 11B-enriched glass in the neutron fluence range of 1.0 × 105–1.0 × 1011 neutron cm−2. This result indicates that γ-rays and high-energy particles in the 10B(n, α)7Li reaction efficiently induced thermoluminescence in the 10B-enriched glass. The TL intensity derived from the 10B(n, α)7Li reaction linearly increased with the thermal neutron fluence in the neutron fluence range of 1.0 × 105–1.0 × 1011 neutron cm−2. The 10B- and 11B-enriched glasses exhibited a wider dynamic range and lower minimum detectable neutron fluence than quite a few TL materials for neutron detection. Therefore, TL glass materials for neutron detection were successfully developed.

Measuring the thermal neutron fluence of NTD-Ge using the self-monitoring method

Thermal neutron irradiation in reactors is widely used in the different frontier fields, including advanced materials and semiconductor devices. The total thermal neutron fluence is a key parameter in various applications. In this paper, we demonstrate a method to determine the total thermal neutron fluence during the manufacture of Neutron Transmutation Doping Germanium (NTD-Ge). The NTD-Ge samples are irradiated at the China Advanced Research Reactor, and their total neutron fluences are measured using the activity of radionuclide 71Ge through the reaction process 70Ge(n,γ)→71Ge ⟶EC71Ga. The activity of 71Ge in a NTD-Ge sample is obtained by measuring the event rates of the gallium KX-rays with energies of 9.2 and 10.3 keV using a Micro-Megas Detector (MMD) and a Silicon Drift Detector (SDD) individually, and incorporating with the correction of detection efficiencies estimated with the Monte Carlo simulation based on the GEANT4 toolkit. The measurements yield the total thermal neutron fluence with high precision, and the results are well consistent with each other between the MMD and SDD method.