Neutron Capture Research Articles

Neutron capture cross section measurement of $$^{129}$$I and $$^{127}$$I using the NaI(Tl) spectrometer of the ANNRI beamline at J-PARC

Neutron capture cross section measurement of $$^{129}$$I and $$^{127}$$I using the NaI(Tl) spectrometer of the ANNRI beamline at J-PARC

Wettability, imbibition and interdiffusion of lithium-based water-in-salt electrolytes in nanoporous carbon in relation to energy storage: A neutron radiography study

Superconcentrated lithium salt electrolytes, or lithium-based Water-in-Salt (WiS) electrolytes emerge as alternative electrolytes in supercapacitors and advanced batteries, such as the Li–O2 ones, especially when combined with a mesoporous carbon with very high surface area. In this work, we analyze the wettability, imbibition, and diffusivity of very concentrated lithium salts in a bimodal mesoporous carbon with pores of 5 nm and 25 nm in diameter. For the first time a neutron radiography technique is used to determine the imbibition and interdiffusion of lithium salts in mesoporous carbon. We found that a 20 m LiTFSI aqueous solution wets the surface of graphite and mesoporous carbon better than water. As a consequence, this WiS electrolyte is embedded rather fast in the mesoporous carbon by the action of capillary forces, exhibiting an intrinsic permeability K = (1.16 ± 0.03).10−18 m2. Finally, by resorting to the different cross-section of the 6Li and 7Li isotopes for the neutron capture, we could analyze the diffusion of LiCl in D2O confined in the mesoporous carbon. The interdiffusion coefficient of the confined LiCl is 5.0 10−7 cm2 s−1, corresponding to a tortuosity coefficient τ = 1.32. This moderate confinement effect results from the strong ion association of the LiCl that precludes the electrostatic interactions of Li+ ions with the negative charge on the pore walls. In summary, the combined properties of the lithium-based WiS electrolytes and the mesoporous carbon allow us to glimpse their use in supercapacitors and Li–O2 batteries.

Implementation of thorium-based fuel for Indonesia Micro Reactor (IMR)

Indonesia has committed to reducing greenhouse gas emissions in the framework of net zero emissions, and nuclear energy will be a part of the energy mix. Indonesia has been developing a 10 MWt micro reactor of high temperature gas-cooled reactor (HTGR) with pebble fuel containing 17 % enriched UO2 kernel with OTTO cycle scheme. For the government’s interest to use thorium, this paper assesses the possibility of replacing uranium with thorium. The assessment is performed by calculating the effective neutron multiplication factor (keff). Neutrons captured by 232Th lead to the formation of two intermediate isotopes, namely 233Th and 233Pa, the latter being a significant neutron absorption cross-section. Neutronics analysis is performed using parameters of thorium mass fraction, core temperature, and active core height. The thorium mass fractions varies from 0 % to 90 % Th. The core temperatures are in the range of the normal operating temperature at 900 K, assumed superheat temperature at 1200 K, and shutdown temperature at 300 K. The active core height varies at 197, 180, 150, 130, and 125 cm. This paper aims to determine the keff of the IMR at maximum thorium/uranium mass fraction composition, maximum active core height, core temperature, and fuel burn-up. The MCNP code is used for analysis. The results show that keff decreases with increasing thorium mass fraction and core temperature, and decreasing active core height. Calculations with thorium mass fraction of 15 %, core active height of 197 cm in 20 steps show that the reactor may be continuously operated for more than 920 days without refueling. The analysis results show that IMR design has negative temperature reactivity, inherent safety characteristics, and the ability to use thorium-based fuels.

Evaluating optimal quality assurance and quality control conditions of activation measurements at the accelerator-based boron neutron capture therapy system employing a lithium target

Evaluating neutron output is important to ensure proper dose delivery for patients in boron neutron capture therapy (BNCT). It requires efficient quality assurance (QA) and quality control (QC) while maintaining measurement accuracy. This study investigated the optimal measurement conditions for QA/QC of activation measurements using a high-purity germanium (HP-Ge) detector in an accelerator-based boron neutron capture therapy (AB-BNCT) system employing a lithium target. The QA/QC uncertainty of the activation measurement was evaluated based on counts, reproducibility, and standard radiation source uncertainties. Measurements in a polymethyl methacrylate (PMMA) cylindrical phantom using aluminum-manganese (Al–Mn) foils and aluminum-gold (Al–Au) foils and measurements in a water phantom using gold wire with and without cadmium cover were performed to determine the optimal measurement conditions. The QA/QC uncertainties of the activation measurements were 4.5% for Au and 4.6% for Mn. The optimum irradiation proton charge and measurement time were determined to be 36 C and 900 s for measurements in a PMMA cylindrical phantom, 7.0 C and 900 s for gold wire measurements in a water phantom, and 54 C and 900 s at 0–2.2 cm depth and 3,600 s at deeper depths for gold wire measurements with cadmium cover. Our results serve as a reference for determining measurement conditions when performing QA/QC of activation measurements using HP-Ge detectors at an AB-BNCT employing a lithium target.

Effect of oxidation on radiation shielding capacity of ZrNbTaMoW Refractory High Entropy Alloys (RHEA) for nuclear reactor applications: Experimental and theoretical assessment

This study presents the production and comprehensive assessment of a Refractory High Entropy Alloy (RHEA) with a ZrNbTaMoW composition for possible usage nuclear reactor applications. The alloy was synthesized using the mechanical alloying (MA) method, and its radiation properties were examined. Elemental powders of Zr, Nb, Ta, Mo, and W were utilized in equimolar quantities during the 120-hour MA process, conducted under a high-purity Ar atmosphere to prevent oxidation. The obtained HEA was then subjected to a mechano-thermal process to transform it into Refractory High-Entropy Oxide (RHEO) for comparing their radiation-shielding performance under oxide conditions. The characterization of the produced alloys through X-ray diffraction (XRD), scanning electron microscopy (SEM), and electron dispersive spectroscopy (EDS). The photon shielding parameters of the alloys were experimentally obtained for various energies emitted from a 133Ba radioactive source, employing a Canberra UltraLEGe semiconductor detector. Furthermore, theoretical calculations were performed to validate the experimental data and to evaluate the alloys' photon shielding capabilities comprehensively. For this purpose, Phy-X/PSD software was employed to determine mass attenuation coefficients, half-value layer, effective atomic number, and effective electron density. Fast neutron absorption capabilities were determined through the calculation of average absorption cross-sections in the 2–12 MeV energy range. The results showed that the oxidation process significantly diminished the photon absorption capacity of RHEA.

DNA-damage RBE assessment for combined boron and gadolinium neutron capture therapy.

Neutron capture therapy (NCT) by 10B and 157Gd agents is a unique irradiation-based method which can be used to treat brain tumors. Current study aims to quantitatively evaluate the relative biological effectiveness (RBE) and dose distributions during the combined BNCT and GdNCT modalities through a hybrid Monte Carlo (MC) simulation approach. Snyder head phantom as well as a cubic hypothetical tumor was at first modeled by Geant4 MC Code. Then, the energy spectra and dose distribution relevant to the released secondary particles during the combined Gd/BNCT were scored for different concentrations of 157Gd and 10B inside tumor volume. Finally, the scored energy spectra were imported to the MCDS code to estimate both RBESSB and RBEDSB values for different 157Gd concentrations. The results showed that combined Gd/BNCT increases the fluence-averaged RBESSB values by about 1.7 times when 157Gd concentration increments from 0 to 2000µg/g for both considered cell oxygen levels (pO2=10% and 100%). Besides, a reduction of about 26% was found for fluence-averaged RBEDSB values with an increment of 157Gd concentration in tumor volume. From the results, it can be concluded that combined Gd/BNCT technique can improve tumor coverage with higher dose levels but in the expense of RBEDSB reduction which can affect the clinical efficacy of the NCT technique.

Study of activation cross sections of double neutron capture reaction on 193Ir for the reactor production route of radiotherapeutic 195mPt

The radiotherapeutic 195mPt is among the most effective Auger electron emitters of the currently studied radionuclides that have a potential theranostic application in nuclear medicine. Production of 195mPt through double neuron capture of enriched 193Ir followed by β−-decay to the radioisotope of interest carried out at the research reactor IBR-2 is described. Because of the high radiation background, radiochemical purification procedure of 195mPt from bulk of iridium was needed to be developed and is detailed here as well. For the first time, cross section and resonance integral for the reaction 194Ir(n,γ)195mIr were determined. Resonance neutrons contribution was established to exceed that of thermal neutrons, and resonance integral for the reaction 194Ir(n,γ)195mIr is calculated to be 2900 b. Specific activity of 195mPt was estimated to reach a value of 38.7 GBq/(g Pt) at IBR-2 by the end of bombardment (EOB).

Combustion Synthesis of Functionalized Carbonated Boron Nitride Nanoparticles and Their Potential Application in Boron Neutron Capture Therapy

In this research, we developed boron-rich nanoparticles that can be used for boron neutron capture therapy as potential carriers for boron delivery to cancerous tissues. Functionalized carbonated boron nitride nanostructures (CBNs) were successfully synthesized in self-propagating combustion waves in mixtures of high-nitrogen explosives and boron compounds. The products’ composition, morphology, and structural features were investigated using Fourier transform infrared spectroscopy, powder X-ray diffraction, low-temperature nitrogen sorption analysis, thermogravimetric analysis, high-resolution scanning electron microscopy, and high-resolution transmission electron microscopy. The extreme conditions prevailing in combustion waves favor the formation of nanosized CBN hollow grains with highly disordered structures that are properly functionalized on the surface and inside the particles. Therefore, they are characterized by high porosity and good dispersibility in water, which are necessary for medical applications. During biological tests, a concentration-dependent effect of the obtained boron nitride preparations on the viability of normal and neoplastic cells was demonstrated. Moreover, the assessment of the degree of binding of fluorescently labeled nanoparticles to selected cells confirmed the relationships between the cell types and the concentration of the preparation at different incubation time points.

An Effective Method for the Evaluation of the Enantiomeric Purity of 1,2-Diacyl-sn-glycero-3-phosphocholine-Based Lipids by NMR Analysis

In this article, we report a very efficient method for the determination of the enantiopurity of 1,2-diacyl-sn-glycero-3-phosphocholine by 1H NMR analysis using a readily available chiral derivatizing boronic acid (CDA), (R)-(2-(((1-phenylethyl)amino)methyl)phenyl)boronic acid. After the removal of the acyl groups of 1,2-diacyl-sn-glycero-3-phosphocholine via methanolysis and washing fatty acid byproducts with CHCl3, the obtained sn-glycero-3-phosphocholine (GPC) with the free diol moiety is derivatized by the chiral boronic acid and analyzed by 1H NMR analysis. The choline methyl resonance of each diastereomer is observed at distinctive chemical shifts in the 1H NMR spectrum. Integration of the respective resonances allows direct determination of the enantiomeric purity. The procedure was tested successfully using 1,2-dipalmitoyl-sn-glycero-3-phosphocholine (DPPC) with different enantiomeric purities and with commercially available 1,2-dimyristoyl-sn-glycero-3-phosphocholine (DMPC) and 1,2-distearoyl-sn-glycero-3-phosphocholine (DSPC).

Cold-Sprayed W/B4C/Al Coatings for Neutron and γ-Rays Absorption

Cold-Sprayed W/B4C/Al Coatings for Neutron and γ-Rays Absorption

A method for delivering the required neutron fluence in an accelerator-based boron neutron capture therapy system employing a lithium target

Accelerator-based boron neutron capture therapy (BNCT) systems employing a solid-state lithium target indicated the reduction of neutron flux over the lifetime of a target, and its reduction could represent the neutron flux model. This study proposes a novel compensatory approach for delivering the required neutron fluence and validates its clinical applicability. The proposed approach relies on the neutron flux model and the cumulative sum of real-time measurements of proton charges. The accuracy of delivering the required neutron fluence for BNCT using the proposed approach was examined in five Li targets. With the proposed approach, the required neutron fluence could be delivered within 3.0%, and within 1.0% in most cases. However, those without using the proposed approach exceeded 3.0% in some cases. The proposed approach can consider the neutron flux reduction adequately and decrease the effect of uncertainty in neutron measurements. Therefore, the proposed approach can improve the accuracy of delivering the required fluence for BNCT even if a neutron flux reduction is expected during treatment and over the lifetime of the Li target. Additionally, by adequately revising the approach, it may apply to other type of BNCT systems employing a Li target, furthering research in this direction.

Actinide isotope analysis and trace element composition of bones of prehistoric animals and humans by inductively coupled plasma mass spectrometry (ICP-MS)

Inductively coupled plasma mass spectrometry (ICP-MS) has been used for analysis of the bones of archanthropus, ancient bear, southern mammoths, dinosaurs, as well as soils collected in various parts of Uzbekistan. The concentrations of 64 elements have been detected including thorium and uranium for which isotope compositions were also analysed. The comparison of the ICP-MS data with the results we obtained earlier for some bones using instrumental neutron activation analysis (INAA) is presented. The concentrations of 234, 235, 238U are up to two orders of magnitude elevated compared to the soils (e.g.238U in south mammoth bone—130.1 mg/kg, and that of soil is only 1.6 mg/kg). The levels of 236U, and 239, 240, 242, 244Pu isotopes (possible neutron capture products of 235U and 238U) correspond to a count rate of blank samples. In addition, the isotope analysis confirms the uranium in the bones and surrounding soils is natural (the average determined for all samples uranium ratios 235U/238U = 0.0071 ± 0.0003 (2-sigma errors). The concentration of 230Th (which is the part of 234U and 238U decay chain) is also elevated in the bones compared to the surrounding soils, however, the concentrations of stable 232Th largely correspond to those of the soil. The excess uranium detected in bones is most probably due to the preferential accumulation from soil, and not to the paleo diet.

Effects of neutron sources and moderator materials on therapeutic dose delivery to deep-seated lesions in proton accelerator-based boron neutron capture therapy

This study aimed to identify the optimal conditions for delivering sufficient doses to deep-seated lesions within short irradiation times for two boron carriers of different T/N ratios. The therapeutic depth and irradiation time of a neutron beam for beam shaping assemblies (BSAs) with a Li or Be target and a MgF2 or CaF2 moderator were examined with the fast-neutron dose per epithermal neutron (FNR) as a parameter. When T/N = 3.61, the therapeutic depth was almost saturated at an FNR of about 10 × 10−13 Gy cm2; when the FNR value was about 10 × 10−13 Gy cm2, the therapeutic depth of the neutron beam for the BSA with a Be target and a MgF2 moderator was almost identical to that for the neutron beam for the BSA with a Be target and a CaF2 moderator, and slightly greater than those for the neutron beams for the BSAs with a Li target and a MgF2 or CaF2 moderator; moreover, the irradiation time of the neutron beam for the BSA with a Be target and a MgF2 moderator was shorter than that for the neutron beam for the BSA with a Be target and a CaF2 moderator. When T/N = 100, the therapeutic depths of the neutron beams for the BSAs varied greatly depending on the FNR, and were greater than the corresponding values for T/N = 3.61. We therefore concluded that the BSA with a Be target and a MgF2 moderator that produced a neutron beam with an FNR of about 10 × 10−13 Gy cm2 is optimal for delivering sufficient doses to deep-seated lesions in short irradiation times when T/N = 3.61, and stricter control over FNR is required when T/N = 100.

Improving instrumental neutron activation analysis detection limit using Monte Carlo simulation for lanthanides determination in Algerian phosphate

ABSTRACT Instrumental Neutron Activation Analysis (INAA) is one of the best methods to analyse Rare Earth Elements (REE) with a low detection limit in phosphate samples. Since most lanthanides emit gamma rays at the energy below 1000 keV, this energy range is affected by the Bremsstrahlung noise of beta emission produced by the 31P(n,γ)32P reaction, when the irradiation of the P2O5 composing the phosphate matrix (approximately 30%). This beta (β−) spectrum with a maximum energy of 1.71 MeV (average energy of 695 keV) affects the gamma spectrum and will reduce the detection limit significantly, other factors decrease the detection limit of lanthanides like neutron self-shielding and interference. In the present work, an analysis methodology based on optimising the measurement parameters of the INAA technique was implemented to improve the detection limits for the precise evaluation of lanthanide concentrations in Algerian phosphate deposits. The complete procedure was developed using a combination of a Monte Carlo simulation approach and experimental measurements. Two models were developed based on Monte Carlo simulation for the calibration of HPGe detector efficiency and the optimisation of type and filter thickness to reduce the beta spectrum (Bremsstrahlung) emitted by the 32P produced by neutron capture reaction 31P(n, γ) 32P. The developed model allowed a considerable improvement in the detection limit of the INAA technique and reduced the uncertainty in the lanthanide concentration. Finally, the methodology was applied to determine lanthanide elements in phosphate samples with good precision.

Efficient optimization of an accelerator neutron source for neutron capture therapy using genetic algorithms.

In recent years, genetic algorithms have been applied in the field of nuclear technology design, producing superior optimization results compared to traditional methods. They can be employed in the design and optimization of beam shaping assemblies (BSA) BSA to obtain the desired neutron beams. But it should be noted that the direct combination of Monte Carlo methods with genetic algorithms requires a significant amount of computational resources and time. Design and optimize BSA more efficiently to achieve neutron beams that meet specified recommendations. We propose an approach of NSGA II with crucial variables which are identified by multivariate statistical techniques. This approach significantly reduces the problem sizes, thus reducing the time required for optimization. We illustrate this methodology using the example of BSA design for AB-BNCT. The computational efficiency has tripled with crucial variables. By using NSGA II, we obtained optimized models conforming to both the new and old version IAEA BNCT guidelines through a single optimization process and subjected them to phantom analysis. The results demonstrate that models obtained through this method can meet the IAEA recommendations with deep advantage depth (AD) and high absorbed ratio (AR). The genetic algorithm with crucial variables displays tremendous potential in addressing BSA optimization challenges.

Matrix effect correction in active neutron interrogation for residual fissile mass assessment in radioactive waste drums

Metallic waste in the form of shells and nozzles remains following the reprocessing of spent fuel at La Hague plant (France). Before the final disposal, drums containing the waste are transported to the compaction facility where its volume is reduced by a factor of 5. With the objective of controlling the criticality/safety levels, an active neutron interrogation system at the entrance of the facility is used to assess the residual fissile mass remaining in the waste, which relies on prompt fission neutron detection (known as the differential die-away technique, DDA). The measured signal is proportionally linked to the fissile mass by a calibration coefficient. However, two effects can produce an inaccurate prediction. The number of induced fissions for the same fissile mass varies according to the waste matrix composition, particularly the neutron slow-down and absorption ratio. Secondly, the presence of fissile clusters can impact the neutron interrogation due to an increase in self-shielding effect. This work presents a matrix effect correction method based on the use of internal flux monitors (sensitive to the neutron absorption ratio) and the transmission signal (sensitive to the slow-down ratio). It relies on a numerical model of the whole measurement cell developed with the Monte-Carlo N-Particle transport (MCNP) code. The calibration coefficient of 72 different matrices, representative of the waste produced at La Hague, were simulated and a statistical model using a multilinear regression was then established. This simulated-based approach provides a more robust and comprehensive estimation of the residual fissile mass without impacting accuracy thanks to the flux monitors and the transmission signals.

Monitoring for Actinide Neutron Emissions From Spent Nuclear Fuel Under Extreme Gamma Fields Using Centrifugally Tensioned Metastable Fluid Detector Sensor Technology

Abstract This paper presents outcomes from research studies conducted independently at Purdue University as part of a collaborative project with the staff of Pacific Northwest National Laboratory (PNNL). Research work focused on assessments for meeting the challenge of monitoring actinide content in spent nuclear fuel (SNF) via characteristic neutron emissions [from spontaneous fission and (α,n) reactions] using the centrifugally tensioned metastable fluid detector (CTMFD) sensor technology using an inert organic decafluoropentane (DFP)(C5H2F10) sensing fluid. Traditional detectors readily saturate and/or cannot monitor neutron emissions under the expected 1012:1 gamma to neutron radiation environment. A challenge problem was posed to examine if a CTMFD could operate reliably over 1 h for conducting neutron spectroscopy at a 1 m standoff from a 30-y cooled SNF, in a ∼1012:1 gamma:neutron intensity environment resulting in a 150 Gy (15 kRad) accumulated dose for the CTMFD. The impacts on reliable operability were studied separately under expected gamma radiation energy and intensity for possible effects of: (i) radiolysis in the CTMFD sensing fluid from absorbed (<1.5 MeV) gamma radiation; (ii) photoneutron contamination signals from < 3 MeV high energy gamma photons interacting with the sensing fluid; and, (iii) malfunction of CTMFD component electronics from the absorbed gamma radiation. A Co-60 gamma irradiator was used for dose accumulation in the CTMFD electronic components and sensing fluids. A 14 MeV DT accelerator was used with a NaCl target to produce 3–4 MeV photons from activated 37S (via. neutron absorption in 37Cl) at SNF-commensurate intensities from SNF at 1-m standoff. Our examinations revealed the absence of any significant impact on CTMFD performance for meeting and exceeding the challenge problem metric. That is, we validated for no discernible impact of: 3–4 MeV gamma-produced photoneutrons when combined with a fission neutron source and radiolysis in the DFP sensing fluid through a 150 Gy absorbed dose. Past research results at Purdue University have validated for survivability of the key electronic components for absorbed gamma doses above the targeted 150 Gy level. This paper also provides extended evidence for survivability (from radiolysis) at higher gamma doses through 750 Gy with a borated DFP-sensing fluid formulation-based CTMFD.

BNCT pancreatic cancer treatment strategy with glucose-conjugated boron drug

Multidisciplinary therapy centered on radical surgery for resectable pancreatic cancer is expected to prolong prognosis, but relies on CA19-9 biomarker levels to determine treatment strategy. Boron neutron capture therapy (BNCT) is a chemoradiotherapy using tumor hyperaccumulator boron drugs and neutron irradiation. The purpose of this study is to investigate novel boron drug agents for BNCT for pancreatic cancer. Bioinformatics was used to evaluate the uptake of current boron amino acid (BPA) drugs for BNCT into pancreatic cancer. The expression of the amino acid transporter LAT1, a BPA uptake transporter, was low in pancreatic cancer and even lower in high CA19-9 pancreatic cancer. In contrast, the glucose transporter was high in high CA19-9 pancreatic cancers and inversely correlated with LAT1 expression. Considering the low EPR effect in pancreatic cancer, we synthesized a small molecule Glucose-BSH, which is boron BSH bound to glucose, and confirmed its specific uptake in pancreatic cancer. uptake of Glucose-BSH was confirmed in an environment compatible with the tumor microenvironment. The therapeutic efficacy and safety of Glucose-BSH by therapeutic neutron irradiation were confirmed with BNCT. We report Glucose-BSH boron drug discovery study of a Precision Medicine BNCT with application to high CA19-9 pancreatic cancer.

Apatite–Wollastonite (AW) glass ceramic doped with B2O3: Synthesis, structure, SEM, hardness, XRD, and neutron/charged particle attenuation properties

In the study, the influence of doping apatite-wollastonite (AW) glass-ceramics (GCs) with B2O3 on the structural, physical, microstructural, and light and heavy particle interaction properties is presented. Using the solid-state reaction and the cold isostatic press method, pristine AW, 10 wt% (AW-B10), and 20 wt% (AW-B20) B2O3-doped AW GCs were prepared. The prepared GCs were subject to structural, chemical compositional, and surface morphological investigation through standard experimental processes. Furthermore, the neutron (fast and thermal) and charged radiation (electron, proton, alpha, and carbon ion) interactional parameters of the GCs were obtained through standard theoretical models and software. The density of AW (2.91 gcm−3) increases to about 2.957 and 2.986 gcm−3 as B2O3 increases to 10 and 20 wt%, respectively. The presence of boron oxide supported the wollastonite phase to remain glass and suppressed crystallisation in the AW GCs. Doping AW with B2O3 improved the interaction probabilities of the GCs with fast and thermal neutrons, electrons, protons, alpha particles, and carbon ions. B-rich AW is thus potentially useful as a target material in human tissues for boron neutron capture therapy for the management of cancer and tumours, isolating or shielding specific tissues in ion beam therapy processes, and other techniques that require the exposure of human tissues to irradiation.

Early Stage In Vitro Bioprofiling of Potential Low-Molecular-Weight Organoboron Compounds for Boron Neutron Capture Therapy (BNCT)-Proposal for a Guide.

Given the renewed interest in boron neutron capture therapy (BNCT) and the intensified search for improved boron carriers, as well as the difficulties of coherently comparing the carriers described so far, it seems necessary to define a basic set of assays and standardized methods to be used in the early stages of boron carrier development in vitro. The selection of assays and corresponding methods is based on the practical experience of the authors and is certainly not exhaustive, but open to discussion. The proposed tests/characteristics: Solubility, lipophilicity, stability, cytotoxicity, and cellular uptake apply to both low molecular weight (up to 500 Da) and high molecular weight (5000 Da and more) boron carriers. However, the specific methods have been selected primarily for low molecular weight boron carriers; in the case of high molecular weight compounds, some of the methods may need to be adapted.