Measurements of fast neutron capture cross sections on <Superscript>63</Superscript>Cu and <Superscript>186</Superscript>W

Neutron capture cross sections on ⁶³Cu and ¹⁸⁶W were measured by fast neutron activation method at neutron energies from 1 to 2 MeV. Monoenergetic fast neutrons were produced by ³H(p,n)³He reaction. Neutron energy spread by target thickness, which was assumed to be the main factor of neutron energy spread, was estimated to be 1.5% at neutron energy of 2.077 MeV. Neutron capture cross sections on ⁶³Cu and ¹⁸⁶W were calculated by reference comparison method on those of ¹⁹⁷Au(n,γ). Not only statistical errors of gamma-counts from samples but also systematic errors in the counting efficiency for HP Ge detector and the uncertainty of areal density of samples were considered in calculating neutron capture cross section. Estimated neutron capture cross sections on ⁶³Cu and ¹⁸⁶W were also compared with ENDF-6 data.     

Accurate determination of half-life and radionuclidic purity of reactor produced <Superscript>177g</Superscript>Lu (<Superscript>177m</Superscript>Lu) for metabolic radiotherapy

Thanks to its favorable decay characteristics, ^177gLu is finding several applications in nuclear medicine, especially for palliative metabolic radiotherapy of cancer and radioimunotherapy. ^177gLu is produced in thermal nuclear reactor either by direct neutron capture ¹⁷⁶Lu(n,γ)^177(m+g)Lu on either natural or enriched ¹⁷⁶Lu target, or by reaction on enriched ¹⁷⁶Yb target followed by negatron decay. The latter method does produce a high radionuclidic purity and high specific activity radionuclide in no-carrier-added form, since ¹⁷⁷Yb decays solely to the ground state ^177gLu. Conversely, the first method does produce a low specific activity ^177gLu in carrier-added form,¹ contaminated by the long-lived radioisotopic impurity ^177mLu. The accurate determination of radionuclidic purity and half-life of ^177gLu carried out by HPGe and LSCS is presented in some details.     

Recent developments in the manufacture of 173lutetium targets

To better model nuclear processes there is an interest in measuring neutron capture cross sections of lanthanide radioisotopes. A natural hafnium target was irradiated with 100 meV protons at the Los Alamos Isotope Production Facility at the Los Alamos Neutron Science Center (LANSCE) to produce neutron poor lutetium radioisotopes. After irradiation, the target was allowed to cool to allow shorter lived lutetium isotopes to decay. This left predominately ¹⁷³Lu with small amounts of ¹⁷⁴Lu. The hafnium target was then chemically processed to isolate the lanthanide fraction through ion exchange techniques. Recent efforts have focused on the separation of lanthanide species to produce an elementally pure lutetium product, the manufacture of small high density lutetium targets, and neutron capture cross section measurements.     

Preparation of a one-curie <Superscript>171</Superscript>Tm target for the detector for advanced neutron capture experiments (DANCE)

Approximately one curie of ¹⁷¹Tm (T _1/2 = 1.92a) has been produced and purified for the purpose of making a nuclear target for the first measurements of its neutron capture cross section. Target preparation consisted of three key steps: (1) material production; (2) separation and purification; and (3) electrodeposition onto a suitable backing material. Approximately 1.5 mg of the target material (at the time of separation) was produced by irradiating ca. 250 mg of its stable enriched ¹⁷⁰Er lanthanide neighbour with neutrons at the ILL reactor in France. This production method resulted in a “difficult-to-separate” 1:167 mixture of near-neighboring lanthanides, Tm and Er. Separation and purification was accomplished using high-performance liquid chromatography (HPLC), with a proprietary cation-exchange column (Dionex, CS-3) and alpha-hydroxyisobutyric acid (α-HIB) eluent. This technique yielded a final product of ∼95% purity with respect to Tm. A portion (20 μg) of the Tm was electrodeposited onto thin Be foil and delivered to the Los Alamos Neutron Science Center (LANSCE) for preliminary analysis of its neutron capture cross section using the Detector for Advanced Neutron Capture Experiments (DANCE). This paper discusses the major hurdles associated with the separation and purification step, including scale-up issues related to the use of HPLC for material separation and purification of the target material from α-HIB and 4-(2-pyridylazo)resorcinol (PAR) colorant.     

Production of monoenergetic MeV-range neutrons by <Superscript>3</Superscript>H(p,n)<Superscript>3</Superscript>He reaction

Monochromatic MeV-energy neutron source for secondary reaction was developed utilizing tritium embedded titanium (Ti-³H) thin film via ³H(p,n)³He reaction. We have measured the neutron energies and the energy spread by resonance reactions of ¹²C(n,tot) and ²⁸Si(n,tot). The available energy was within the range from 0.6 to 2.6 MeV. Energy spread was 1.6% at energy of 2.077 MeV. The flux in the beam direction was determined to be 3.76·10⁷ n/s/sr by irradiating ¹⁹⁷Au by about 2 MeV neutrons. This source was shown to be useful for measurements of nuclear data by measuring the total cross sections of neutrons on Fe and comparing these data to the data of ENDF-6.     

Measurement of <Superscript>36</Superscript>Cl in nuclear wastes and effluents: Validation of a radiochemical protocol with an in-house reference sample

Summary ³⁶Cl is a beta-emitter with a very low specific activity. It is produced during the irradiation of nuclear fuel, in the reactor core of power plants, from neutron capture by stable ³⁵Cl that may be present at trace level in any part of the irradiated material. Due to its long half-life (T_1/2 = 3.01 ^. 10⁵ y), ³⁶Cl may be significant in impact assessment studies of disposal sites of nuclear wastes. Considering these different elements, the National Radioactive Waste Management Agency (Andra-France) requests information on the ³⁶Cl content of the waste packages destined to be stored at Andra sites. As for other halogens, the measurement of ³⁶Cl is a difficult analytical task in view of its potential losses during the different chemical steps and also because of the lack of international certified reference material needed to validate the chemical and measurement procedures. This paper describes the methodology processed to constitute an in-house solid reference sample with a known content of stable and radioactive chlorine. The original radiochemistry developed to extract ³⁶Cl from solid samples and purify it before a liquid scintillation counting is explained. The comparison of the results given by this radiochemical protocol and other methods allow its validation. The replication of the measurements on the constituted reference materials gives a repeatability around 8% at a confidence level of 95% that is very close to the calculated combined uncertainty value.     

Study of the <Superscript>241</Superscript>Am(n,2n)<Superscript>240</Superscript>Am reaction cross section in the energy range of <Emphasis Type="Italic">E</Emphasis><Subscript><Emphasis Type="Italic">n</Emphasis></Subscript> = 8.8−11.1 MeV

The measurement of the cross section of the reaction ²⁴¹Am(n,2n)²⁴⁰Am has been performed at neutron energies from 8.8 to 11.1 MeV, implementing the activation technique. The neutron beam was produced at the TANDEM accelerator of NCSR “Demokritos” by the ²H(d,n)³He reaction, using a deuterium gas target. During the 5-day long irradiation, the neutron beam fluctuations were monitored in 100 seconds intervals by a BF₃ counter connected with a multiscaling unit. The radioactive target consisted of a 37 GBq ²⁴¹Am source enclosed in a Pb container. A natural Au foil, a ²⁷Al foil and a ⁹³Nb foil were used as reference materials for the neutron flux determination. After the end of the irradiation the activity induced at the target and the reference foils, was measured off-line by a 56% HPGe detector.     

PC-controlled radiochemistry system for preparation of no-carrier-added <Superscript>64</Superscript>Cu

A homemade modular PC-controlled radiochemistry system, used to separate the no-carrier-added ⁶⁴Cu from irradiated electroplated solid cyclotron target ⁶⁴Ni layers was developed. Upon irradiation the target layer is dissolved in diluted nitric acid, followed by selective liquid–liquid extraction of the ⁶⁴Cu. The final purification step is achieved making use of an anion exchange column. All the separation procedures can be carried out remote-controlled with a minimized risk of operator errors within less than 2 h, with excellent processing yield (>95%).     

Fission studies in the reaction of <Superscript>19</Superscript>F with <Superscript>209</Superscript>Bi at 99.2 MeV

Fission product yield studies in the reaction of 99.2 MeV ¹⁹F with ²⁰⁹Bi have been carried out for the first time using gamma-ray spectrometry. The cross sections for the production of fission products have been determined. The yield distribution of fission products was found to be symmetric and broad with FWHM around 22 mass units and peak near mass 111. The average number of neutrons emitted per fission has been found to be around 6.7. The comparison of the fission products yield distribution of ²⁰⁹Bi using projectiles like ⁴He, ¹²C, ¹⁶O, and ¹⁹F have shown that the mass of symmetric peak increases as the mass of the compound nucleus increases. The high fission yield around mass 112 has been attributed to the presence of deformed neutron shells. The total fission cross section and width of the mass distribution have been found to be low in case of ¹⁶O induced fission as compared to the ⁴He, ¹²C, and ¹⁹F induced fission of ²⁰⁹Bi.     

Study of the cyclotron production of <Superscript>172</Superscript>Lu: an excellent radiotracer

¹⁷²Lu due to its suitable (T _1/2 = 6.7 days) and high detection sensitivity, is used as a radiotracer in different fields. ¹⁷²Lu appears to be suitable as a long-lived rare-earth tracer for compound labelling and biodistribution studies. In the present study, excitation functions via ¹⁷²Yb(p,n)¹⁷²Lu, ^natYb(p,xn)¹⁷²Lu, ¹⁷²Yb(d,2n)¹⁷²Lu and ^natYb(d,xn)¹⁷²Lu reactions were calculated by ALICE/91, ALICE/ASH and TALYS-1.2 codes. Deposition of ^natYb₂O₃ on Cu substrate was carried out via sedimentation method for the production of ¹⁷²Lu. Cementation separation process and liquid–liquid extraction (LLX) of no-carrier-added (nca) radiolutetium from irradiated ytterbium(III)oxide target hydrochloric solution was described using Na(Hg) amalgam, α-hydroxyisobutyric acid (α-HIB) and di-(2-ethylhexyl)phosphoric acid (HDEHP).     

Nuclear excitation in <Superscript>176</Superscript>Lu by positron annihilation on K-shell electrons

In the process of nuclear excitation in positron-electron annihilation (NEPEA) experimental results have been found to be larger than theoretical predictions. In our previous works we have analyzed the NEPEA experimental data for 1078 keV in ¹¹⁵In and for 1330 keV in ¹¹¹Cd, respectively, by applying our model of indistinguishable quantum oscillators. In the present work we extend this model to estimate the cross section of nuclear excitation in positron-electron annihilation for ¹⁷⁶Lu from the experimentally measured effective cross section of σ ^eff = (2.6±0.9)·10⁻²⁹ cm². The estimated NEPEA cross section of σ=2.7·10⁻²⁶ cm² is in very good agreement with that of the theoretical prediction σ ^th=2.2·10⁻²⁶ cm².     

Radionuclide purity assessment and calibration of <Superscript>90</Superscript>Y(<Superscript>177</Superscript>Lu) glass particles using gamma-ray spectrometry

Intra-hepatic administration of radioactive glass microspheres is a treatment for patients with primary liver cancer and hepatic metastases. The purpose of this study was radionuclide purity assessment of new glass particles containing two radionuclide, ⁹⁰Y as a therapeutic source and also ¹⁷⁷Lu as a source of diagnostic gamma. For the mixed source, activity measurement using a dose calibrator cannot be used and we need new calibration methods. YAS (Yb) and YAS compositions were sol–gel derived glass particles and production of ⁹⁰Y (¹⁷⁷Lu) and ⁹⁰Y particles was performed using the Tehran Research Reactor. The radionuclide purity was carried out using γ-spectrometry with HPGe detector. A non-destructive spectroscopic assay was employed due to a newly updated low uncertainty positron branching ratio of ⁹⁰Y that emit 511 keV annihilation radiations. In another method, a new calibration of ⁹⁰Y using a non-destructive spectroscopic assay of ⁸⁸Y were investigated. Potential radionuclide impurity include: ⁸⁸Y, ¹⁵²Eu, ⁶⁰Co with activity 100, 50 and 5 Bq per 1 mg of that are not harmful for patients due to delivering radioactive particles about 20–50 mg in ⁹⁰Y(¹⁷⁷Lu) glass microspheres. Among of radionuclide impurity, ¹⁵²Er with a half life of 13.54 years and ⁸⁸Y with a half life of 106.65 days was important in the residual delivery device. For calibration of ⁹⁰Y with monitoring of 511 keV, errors were12.2–21%. In calibration of ⁹⁰Y using gamma spectroscopic assay of ⁸⁸Y, there was an error less than 14%. Spectroscopic assay of ⁸⁸Y can be performed easily and has more repeat for our purpose.     

Development of <Superscript>177</Superscript>Lu-DOTA-anti-CD20 for radioimmunotherapy

Rituximab was successively labeled with ¹⁷⁷Lu-lutetium chloride. ¹⁷⁷Lu chloride was obtained by thermal neutron flux (4 × 10¹³ n cm⁻² s⁻¹) of natural Lu₂O₃ sample with a specific activity of 2.6–3 GBq/mg. The macrocyclic bifunctional chelating agent, N-succinimidyl-1,4,7,10-tetraazacyclododecane-1,4,7,10-tetraacetic acid (DOTA-NHS) was prepared at 25 °C using DOTA, N-hydroxy succinimide (NHS) in CH₂Cl₂. DOTA-rituximab was obtained by the addition of 1 mL of a rituximab pharmaceutical solution (5 mg/mL, in phosphate buffer, pH 7.8) to a glass tube pre-coated with DOTA-NHS (0.01–0.1 mg) at 25 °C with continuous mild stirring for 15 h. Radiolabeling was performed at 37 °C in 24 h. Radio-thin layer chromatography showed an overall radiochemical purity of >98% at optimized conditions (specific activity = 444 MBq/mg, labeling efficacy; 82%). The final isotonic ¹⁷⁷Lu-DOTA-rituximab complex was checked by gel electrophoresis for structure integrity control. Radio-TLC was performed to ensure that only one species was present after filtration through a 0.22 μm filter. Preliminary biodistribution studies in normal rats were carried out to determine complex distribution of the radioimmunoconjugate up to 168 h. The biodistribution data were in accordance with other antiCD20 radioimmunoconjugates already reported.     

Measurement of <Superscript>99</Superscript>Mo production cross-section from the <Superscript>100</Superscript>Mo(n,2n) reaction with quasi monoenergetic neutron based on the <Superscript>9</Superscript>Be(p,n) reaction

The production cross-section of the medical isotope, ⁹⁹Mo from the enriched ¹⁰⁰Mo(n,2n) reaction with the average neutron energies of 21.9 and 26.5 MeV has been determined for the first time by using an off-line γ-ray spectrometric technique. The average neutron energies were generated by using the ⁹Be(p,n) reaction with the proton energies of 35 and 45 MeV from the MC50 cyclotron of the Korea Institute of Radiological and Medical Sciences (KIRAMS) at Seoul, South Korea. The ¹⁰⁰Mo(n,2n) reaction cross-section as a function of neutron energy was also calculated theoretically by using the computer code TALYS-1.8 and EMPIRE-3.2 Malta. The experimental results are in close agreement with the theoretical values from TALYS-1.8. However, the present data at the neutron energy of 21.9 MeV is slightly lower and at 26.5 MeV is higher than the values from EMPIRE-3.2 Malta.     

A combined method of neutron activation analysis and radiometric measurements for <Superscript>234</Superscript>U and <Superscript>238</Superscript>U determination in soil samples of low uranium concentration

A method that combines the use of non-destructive neutron activation analysis and high-resolution α spectrometry has been developed for determination of the activities of ²³⁴U and ²³⁸U in geological samples of low uranium content. The ²³⁸U content is determined by k₀-based neutron activation analysis, whereas the ²³⁴U/²³⁸U relationship is measured by α spectrometry after isolation and electrodeposition of the uranium extracted from a lixiviation with 6 M HCl. The main advantage of the method is the simplicity of the chemical operations, including the fact that the steps destined to assure similar chemical state for the tracer and the uranium species present in the sample are not necessary. The method was applied to soil samples from sites of the North Peru Coast. Uranium concentration range 3–40 mg/kg and the isotopic composition correspond to natural uranium, with about 10% uncertainty.     

<Superscript>234</Superscript>U/<Superscript>238</Superscript>U and <Superscript>235</Superscript>U/<Superscript>238</Superscript>U ratios in domestic water from the environs of Obuasi mine in Ghana

Activity concentrations of ²³⁸U, ²³⁵U and ²³⁴U were determined in different sources of drinking water at the Obuasi gold mines and its surrounding areas in Ghana. Water samples collected from the mines and its surrounding areas were analyzed using direct gamma-ray spectrometry and neutron activation analysis. The ²³⁴U/²³⁸U and ²³⁵U/²³⁸U ratios were calculated and the mean values range from 1.27 to 1.38 and from 0.044 to 0.045 respectively. The average ²³⁴U/²³⁸U ratio was from 1.27 for groundwater to 1.38 for treated water, demonstrating the lack of equilibrium. The average ²³⁵U/²³⁸U activity ratio is 0.045, indicating that only natural uranium was detected in the samples investigated.     

Single nucleon transfer reactions near Coulomb barrier on <Superscript>197</Superscript>Au

Summary Single neutron transfer reactions on ¹⁹⁷Au induced by ¹²C and ¹⁶O at near Coulomb barrier energies have been studied using radiochemical and off-line gamma-ray spectrometric techniques. High spin yield fraction (HSF = σ_m/(σ_m+σ_g)) for ^196m,gAu and formation cross section ratio (σ₁₉₈/σ₁₉₆) of ¹⁹⁸Au and ¹⁹⁶Au have been determined for both ¹⁹⁷Au+¹²C and ¹⁹⁷Au+¹⁶O reactions at different excitation energies near the Coulomb barrier. The HSF of ^196m,gAu and σ₁₉₈/σ₁₉₆ values in both reactions are seen to sharply increase with increasing excitation energy near the Coulomb barrier and then slowly increase reaching a constant value at high projectile energy. These observations have been analyzed in terms of nuclear structural and kinematic parameters. A comparison the effects of Coulomb excitation and nucleon transfer process during heavy-ion interactions indicates a predominant effect of the latter process in the background of the Coulomb excitation. Observed reaction systematics seem to follow the quasi-elastic transfer process.     

Performance comparison of 2.8 MeV and <Superscript>241</Superscript>Am-Be neutrons based moisture measurement setups

Performance of a ²⁴¹Am-Be neutron source-based and 2.8 MeV neutrons-based moisture measurement setups have been compared using Monte Carlo simulation. In the setup fast neutrons transmitted through the sample were detected by a fast neutron detector, which was placed behind a massive long double truncated collimator. The setup geometry was optimized to detect maximum effect of 1–7 wt.% moisture on the neutron intensity transmitted through the sample. The yield of neutrons transmitted through concrete, coal, wood and soil samples containing 1–7 wt.% moisture was calculated for 2.8 MeV neutrons and neutrons from an ²⁴¹Am-Be source. The slopes of the fast neutron intensities transmitted through the samples vs. their moisture contents are very sensitive to the neutron energy and the sample composition. Higher slopes have been observed for the samples with larger bulk density. The slopes of fast neutron yield show dependence on the incident neutron energy. Larger slopes have been observed for neutrons with samller energy. Due to the overall large slopes of the transmitted intensity data of the samples for 2.8 MeV neutrons, it is expected to achieve better sensitivity in moisture measurements for a 2.8 MeV neutrons based moisture setup.     

INAA with anticoincidence counting significantly reduces interferences from the 554.3-keV photopeak of <Superscript>82</Superscript>Br to allow reliable measurements of nanogram levels of arsenic in solid biological materials via the 559.1-keV photopeak of <Superscript>76</Superscript>As

It is rather difficult to measure low levels of arsenic by instrumental neutron activation analysis (INAA) in biological materials containing high levels of bromine and antimony in particular, as well as sodium and potassium. The 559.1-keV photopeak of ⁷⁶As is interfered with by the 554.3- and 564.1-keV photopeaks of ⁸²Br and ¹²²Sb, respectively. The use of INAA in conjunction with anticoincidence spectrometry (INAA-AC) was found to reduce the background under the 559.1-keV photopeak by factors of 4–16 for the biological reference materials analyzed and to decrease the detection limit to 0.35 µg kg^?1 making the measurement of nanogram amounts arsenic in them possible.     

<Superscript>99</Superscript>Mo/<Superscript>99m</Superscript>Tc generators based on aluminum molybdate gel matrix prepared by nano method

A procedure for preparation of ⁹⁹Mo/^99mTc radioisotope generator based on low specific activity neutron activated ⁹⁹Mo was developed. Aluminum molybdate(VI)-⁹⁹Mo of high Mo(VI) content (~?364 mg/g Al⁹⁹Mo) was prepared by mixing low specific activity molybdate(VI)-⁹⁹Mo and aluminum mixture solution with isoamyl alcohol. Al⁹⁹Mo gel matrix was precipitated when the pH of the mixture solution was raised to ~?5 by addition of NaOH to the mixture. Radiometric measurements indicate the strong fixation of Molybdate(VI)-⁹⁹Mo species in the form of the sparingly insoluble Al⁹⁹Mo gel matrix. The prepared AlMo gel matrix was physiochemically characterized. Al⁹⁹Mo gel matrix was used as a base material for preparation of ⁹⁹Mo/^99mTc generator. The ^99mTc eluted from ⁹⁹Mo/^99mTc radioisotope generator was found to have relatively high elution yield (84?±?2.3%), radionuclidic (≥?99.99%), radiochemical (98.1?±?0.9%) and chemical purity.