Radiotracer technique in adsorption study: Part XVII. Removal Behaviour of Alkali Metal (K- and Li-) Titanates for Cd(II)

Alkali metal (potassium and lithium) titanates were synthesized and employed for the efficient removal of Cd(II) ions from aqueous solutions using the radiotracer technique. The possible mechanism involved at the solid/solution interface was deduced with the help of various physico–chemical data, i.e. effect of adsorptive concentration temperature and pH. The effect of added cations and H⁺ (HCl/H₂SO₄) in the uptake process was also seen. The radiation stability of these materials in the adsorption process was assessed by employing a 11.1 GBq (Ra–Be) neutron source having an integral neutron flux of 3.85×10⁶ n·cm⁻² s⁻¹ and associated with a nominal γ-dose of ca. 1.72 Gy/h.     

Towards a laser neutron driver

During the last few years, important experimental investigations have been made concerning the possibility of induced nuclear fission of high-Z elements by electromagnetic interaction (photofission, electron fission, neutron fission). Fast ions, neutrons and fission fragments from such interactions can be used to pump a laser medium, to produce energy from the ²³²Th–²³³U nuclear fission cycle. The main aim of the present work is to study a three-step process, in a relatively new experimental scheme, in order to improve the number of both neutrons and fast ions. In the proposed scheme, high-energy particles and photons are produced by high-intensity laser beam interaction with a solid or gas target, which are utilized later on to trigger the nuclear reactions for the production of (photo) neutrons. These neutrons can give rise to fission of ²³²Th that leads through a cascade of decays to ²³³U —a highly fissionable material. Such a process will enhance, by an important factor, the final neutron flux and the energetic fission fragments. The use of a high intensity pulsed laser beam will control the turn-on and turn-off of the nuclear reactions and allow one to ensure the security of the whole operation. Finally, the produced neutrons are used to accomplish a major population inversion in an appropriate gas medium for the last stage of amplification of a high-contrast ultra-short laser seed pulse.     

Measurement of Pu γ-ray intensities between 148 and 161 keV

Utilizing Pu isotopic standards characterized by the Institute for Nuclear Materials and Measurement (IRMM), New Brunswick Laboratory (NBL) measured the γ-ray branching ratios of the 148.567-keV ²⁴¹Pu, the 152.72-keV ²³⁸Pu, the 159.995-keV ²⁴¹Pu and the 160.28-keV ²⁴⁰Pu γ-rays. The study reported here includes the measurement of four IRMM standards, utilizing three different detectors for a total of 16 spectra, finding branching ratios of (1.863 ± 0.008) × 10⁻⁶, (9.230 ± 0.068) × 10⁻⁶, (6.321 ± 0.040) × 10⁻⁸ and (4.065 ± 0.017) × 10⁻⁶, γ/disintegration, respectively.     

Ultra-low-level determination of U in IAEA marine reference materials by ICPMS and AMS

The development of analytical procedures for the measurement of ultra-low levels of ²³⁶U in marine samples using high-resolution inductively coupled plasma mass spectrometry (HR-ICPMS) and accelerator mass spectrometry (AMS) techniques are discussed and results are presented for IAEA reference materials—marine sediments (IAEA-135, 306, 384 and 385), marine biota (IAEA-134 and 414) and seawater (IAEA-381), collected in areas affected by nuclear reprocessing plants and nuclear weapons tests. The obtained minimum detection limit of the ²³⁶U/²³⁸U atom ratio was 1×10⁻⁸ for AMS and 1×10⁻⁶ for HR-ICPMS.     

Stability of biodegradable radioactive rhenium (Re-186 and Re-188) microspheres after neutron-activation

Our objective was to determine if microspheres made from the biodegradable polymer poly(lactic acid) that contained rhenium could withstand the conditions of direct neutron activation necessary to produce therapeutic amounts of radioactive rhenium. The radiation damage of the polymer produced by γ-doses of up to 1.05 MGy from Re-186 and Re-188 was examined by scanning electron microscopy and size exclusion chromatography. At a thermal neutron flux of 1.5×10¹³ n/cm²/s the microspheres melted after 3 h in the nuclear reactor, but suffered little damage after 1 h of radiation and released less than 5% of the radioactivity during incubation in buffer at 37°C. The radioactive microspheres produced in this manner have a specific activity too low for radioembolization for treatment of liver tumors, but could be injected directly into tumors or applied topically to the wound bed of partially resected tumors.     

肿瘤硼中子俘获治疗的理论基础与近期研究进展

硼中子俘获治疗(BNCT)是一种新型肿瘤精准治疗方法 ,通过肿瘤细胞内的10B俘获热中子发生核裂变反应产生a粒子和反冲7Li核选择性地杀死肿瘤细胞.将足量的10B选择性递送到肿瘤细胞内部是BNCT成功的关键.本文简要介绍了BNCT治疗肿瘤的理论基础,综述了BNCT所用的中子源和硼递送剂的近期研究进展,简述了BNCT临床...     

Determination of 4.438 MeV γ-ray to neutron emission ratio from a Am–Be neutron source

Gamma-ray (γ-ray) spectra of a 1.49×10¹¹ Bq ²⁴¹Am–⁹Be source and background were measured using a 2 in×2 in, NaI(Tl) detector. Backgrounds due to the neutron interactions and energy deposition were calculated with MCNP4C. By subtracting the backgrounds from the experimental spectra, the S_γ is obtained and the R=S_γ/S_n was estimated. The final result of R=S_γ/S_n=0.596 is in agreement with result reported in literature.     

Effect of Ar ion irradiation on electrochemical behaviors of zircaloy-4

In order to investigate the effect of the damage on the electrochemical behavior, zircaloy-4 specimens were irradiated by Ar ions using an accelerator at an energy of 50 keV, with doses from 3×10¹⁴ to 3×10¹⁶ ions/cm² at liquid nitrogen temperature. Potentiodynamic polarization measurements were employed to evaluate the corrosion resistance of zircaloy-4 in a 0.5 M H₂SO₄ water solution at room temperature. A damage efficiency function on the passive current density of the polarization curve was defined and deduced theoretically.     

The feasibility of accelerator-based in vivo neutron activation analysis of nitrogen

The feasibility of accelerator-based in vivo neutron activation analysis of nitrogen has been investigated. It was found that a moderated neutron flux from 10 μA of 2.5 MeV protons on a ⁹Be target performed as well as, and possibly slightly better than the existing isotope-based approach in terms of net counts per unit subject dose. Such a system may be an attractive alternative to the widespread use of ^238,239Pu/Be or ²⁵²Cf neutron sources, since there is more flexibility in the energy spectrum generated by accelerator-based neutron sources. From a radiation safety standpoint, accelerators have the advantage in that they only produce radiation when in operation. Furthermore, an accelerator beam can be pulsed, to reduce background detected in the prompt-γ measurement, and such a device has a wide range of additional biological and medical applications.     

Characterization of trace elements in Casearia medicinal plant by neutron activation analysis.

Leaves of Casearia sylvestris, Casearia decandra and Casearia obliqua plant species, collected at the Atlantic Forest in Brazil, were analyzed by using instrumental neutron activation analysis (INAA). Short and long irradiations using thermal neutron flux of the IEA-R1 nuclear reactor were carried out for these analyses. Concentrations of Ca, K and Mg were found in these samples at the percentage levels, Br, Cl, Fe, Mn, Na, Rb and Zn at the μg g⁻¹ levels and Co, Cr, Cs, La, and Sc at the μg kg⁻¹ levels. Comparisons were made among the element concentrations obtained in these three Casearia species and significant differences were found for the elements Cl, Co, Cs, Cr, La, Mn, Na and Sc. The precision and the accuracy of the results were evaluated by analyzing the certified reference materials NIST-1515 Apple Leaves and NIST-1573a Tomato Leaves.     

Investigation of the NMR relaxation rate dose-response of a ceric sulphate dosimeter

The relationship between the radiation absorbed dose and the NMR longitudinal and transversal relaxation rates, R₁ and R₂, respectively, of a ceric sulphate dosimeter was examined. By adding copper sulphate, the R₁ and R₂ dose-responses were found to be linear up to 60 kGy with dose sensitivities of 13×10⁻⁶ and 15×10⁻⁶ s⁻¹ Gy⁻¹, respectively. There is thus the potential for a three-dimensional ceric dosimeter for high dose applications, provided a suitable gelling substance is used.     

Neutron flow measurements in the d(14) + Be neutron radiation field from the cyclotron in Essen

It exists the possibility in the neutron therapy of specific tumours to increase the local energy dose in the tumour by treatment with fast neutrons under use of B-10. Thermal neutrons cause the neutron capture reaction B-10(n, alpha)Li-7 in tissue. A sufficiently high number of thermic neutrons is necessary combined with a possibly selective concentration of B-10 in the tumour as target volume for the clinical application. Gold foils were activated in a water phantom at several depths for the quantitative determination of the thermal neutron fluence. A result of this measurements is a maximal thermal neutron fluence of 1.44 X 10(10) cm-2 at depth of 5 cm with applied total energy dose of 0.8 Gy.     

Radiochemical purification of no-carrier-added scandium-47 for radioimmunotherapy

A procedure, adaptable to large-scale remote operation, was developed to purify no-carrier-added (NCA) ⁴⁷Sc from irradiated Ti targets. Methods based on extraction chromatography and cation-exchange chromatography were compared. Results of this comparison led to the development of an optimized procedure based on cation-exchange with Dowex AG 50W-X4 and ⁴⁷Sc elution with HCl/HF. This method gave 90–97% overall ⁴⁷Sc recovery, with a Ti separation factor greater than 2.4×10⁻⁵, and specific activities ≥0.9 GBq μg⁻¹. Use of the ⁴⁷Sc product, for labeling monoclonal antibodies, resulted in consistent labeling yields of ≥90%.     

A feasibility study of the Tehran research reactor as a neutron source for BNCT

Investigation on the use of the Tehran Research Reactor (TRR) as a neutron source for Boron Neutron Capture Therapy (BNCT) has been performed by calculating and measuring energy spectrum and the spatial distribution of neutrons in all external irradiation facilities, including six beam tubes, thermal column, and the medical room. Activation methods with multiple foils and a copper wire have been used for the mentioned measurements. The results show that (1) the small diameter and long length beam tubes cannot provide sufficient neutron flux for BNCT; (2) in order to use the medical room, the TRR core should be placed in the open pool position, in this situation the distance between the core and patient position is about 400 cm, so neutron flux cannot be sufficient for BNCT; and (3) the best facility which can be adapted for BNCT application is the thermal column, if all graphite blocks can be removed. The epithermal and fast neutron flux at the beginning of this empty column are 4.12×10⁹ and 1.21×10⁹ n/cm²/s, respectively, which can provide an appropriate neutron beam for BNCT by designing and constructing a proper Beam Shaping Assembly (BSA) structure.     

Dependency of the [18F]fluoromisonidazole uptake on oxygen delivery and tissue oxygenation in the porcine liver

We have previously shown that the accumulation of fluorine-18-labeled fluoromisonidazole ([¹⁸F]FMISO) is inversely correlated to tissue oxygenation, allowing the quantification of porcine liver tissue hypoxia in vivo. We determined the activity from administered [¹⁸F]FMISO in relation to the hepatic oxygen availability and the partial pressure of oxygen in tissue (tPO₂) to define a critical oxygen delivery on a regional basis. [¹⁸F]FMISO was injected 2 h after onset of regional liver hypoxia due to arterial occlusion of branches of the hepatic artery in 10 domestic pigs. During the experimental procedure the fractional concentration of inspired oxygen (FiO₂) was set to 0.67 in group A ( N=5) and to 0.21 in group B ( N=5) animals. Immediately before sacrifice, the tPO₂ was determined in normal flow and flow-impaired liver segments. The standardized uptake values (SUV) for [¹⁸F]FMISO was calculated from 659 single tissue samples obtained 3 h after injection of approximately 10 MBq/kg body weight [¹⁸F]FMISO and was compared with the regional total hepatic oxygen delivery (DO₂) calculated from the regional arterial and portal venous flow (based on ¹⁴¹Ce- and ^99mTc-microspheres measurements) and the oxygen content of the arterial and portal venous blood. In 121 tPO₂-measured liver tissue samples, the mean DO₂ was significantly decreased in occluded liver tissue samples [group A: 0.063 (0.044–0.089); group B: 0.046 (0.032–0.066)] compared to normal flow segments [group A: 0.177 (0.124–0.252); group B: 0.179 (0.128–0.25) mL·min⁻¹·g⁻¹; geometric mean (95% confidence limits); p < 0.01 in group A and p < 0.001 in group B]. The tPO₂ of occluded segments [group A: 5.1 (3.2–8.1); group B: 3.9 (2.4–6.2) mm Hg] was significantly decreased compared to normal flow segments [group A: 20.2 (12.6–32.5); group B: 22.4 (14.3–35.2) mm Hg; p < 0.01 in group A and p < 0.001 in group B]. Three hours after [¹⁸F]FMISO administration, the mean [¹⁸F]FMISO SUV determined in tPO₂-measured occluded segments was significantly higher [group A: 4.08 (3.12–5.34), group B: 5.43 (4.14–7.13)] compared to normal liver tissue [group A: 1.57 (1.2–2.06), group B: 1.5 (1.16–1.93); p < 0.001 for both groups]. The [¹⁸F]FMISO SUV allowed prediction of the tPO₂ with satisfying accuracy in hypoxic regions using the exponential regression curve { [¹⁸F]FMISO=1.05+6.7^{(−0.117 tPO₂}); r²=0.75;p < 0.001}. In addition, regardless of ventilation conditions, a significant exponential relationship between the DO₂ and the [¹⁸F]FMISO SUV was found ( r²=0.39,p < 0.001). Our results suggest that the reduction of the oxygen delivery below the critical range of 0.1–0.11 mL·min⁻¹·g⁻¹ regularly causes liver tissue hypoxia. The severity of hypoxia is reflected by the [¹⁸F]FMISO accumulation and allows the in vivo estimation of the tPO₂ in hypoxic regions.     

Light weight concrete: Ra, Th, K contents and dose reduction assessment

Four types of light weight concrete (LWC) commonly used in Hong Kong, namely, autoclave aerated concrete (plus lime), autoclave aerated concrete (plus Pulverized Fuel Ash or PFA), concrete with synthetic aggregate ‘Leca’ and concrete with polystyrene bean as aggregate were measured for their ²²⁶Ra, ²³²Th and ⁴⁰K contents using high-resolution gamma spectrometry. All the radionuclide contents except those for the PFA autoclave aerated concrete were below the world averages of building materials. The Ra-equivalents for these four LWC were 49.6, 249, 122 and 44.2 Bq kg⁻¹, respectively, and were much smaller than a recommended limit of 370 Bq kg⁻¹ for construction materials for dwellings. The gamma-dose rate for an indoor environment with partition walls built with LWC was estimated to be about 20 × 10⁻⁸ Gy h⁻¹, which corresponded to a reduction in the effective dose of about 0.25 mSv y⁻¹ when compared to that obtained for an indoor environment built with normal concrete (NC) only. The Rn exhalation rates from the three lowest Ra-equivalent LWC were calculated as 0.509, 1.28, and 0.335 mBq m⁻² s⁻¹, respectively, which corresponded to a reduction in the indoor Rn concentration of around 14 Bq m⁻³ and reduction in the tracheobronchial dose reaching 1 mSv y⁻¹ by using the James lung-dosimetry model, when compared to the case of NC.     

硼中子俘获治疗的加速器中子源7Li(p,n)7Be的中子特性研究

目的 通过研究质子加速器 ⁷Li(p,n)⁷Be 反应的中子特性,为研究和制作适用于硼中子俘获治疗(BNCT)的加速器中子源提供基础数据。方法 加速质子使其轰击Li靶后产生中子;通过金属箔活化法,测量中子与In箔发生阈值反应后放出的γ射线;然后计算出In箔的放射性活度、加速器反应后放出中子的注量和反应的微分截面。结果 质子加速轰击Li靶后,在不同方向产生不同能量和注量的中子。加速器电压分别为3.0、2.8和2.6 MV,出射中子与入射质子束的方向一致时, ⁷Li(p,n)⁷Be 反应的微分截面约为50 mb/mr;夹角为60°时,反应的微分截面减小到30 mb/mr左右。由于部分中子与其他金属原子等发生弹性散射而射向后方,提高了这一范围内In箔的比放射性活度,影响了其微分截面的准确性。结论 用金属箔活化法测定中子简便易行,可同时测得多个方向的中子分布,但需对中子与其他金属弹性散射产生的影响进行进一步的研究; ⁷Li(p,n)⁷Be 反应后发射出的中子经慢化后,能得到适于BNCT治疗的热中子和超热中子;若作为BNCT的中子源,加速器的质子束流需达到10 mA。     

Feasibility study of explosive detection for airport security using a neutron source

Application of on-line prompt gamma activation analysis technique to the detection of explosive-like materials in airports is presented. The explosive detecting assembly consists of five bismuth germanate scintillation detectors, five timing single-channel analyzers and one 20 μg ²⁵²Cf isotopic neutron source. Low-level discriminator and energy window for each timing single channel analyzer were adjusted to the proper energy range to detect the 10.8 MeV high-energy prompt photon emitted from the ¹⁴N(n, γ) reaction, which the nitrogen in chemical explosives has in high proportions. Urea is used in laboratory tests to simulate an explosive-like material. The space-flux distribution of thermal neutrons near inspected luggage and the detection limit of explosives with various scanning periods were determined. The features and possible modifications of this preliminary setup for airport security inspection are discussed in detail.     

Standardization of I by the coincidence method and practical applications

A new coincidence system, based on two thin NaI(Tl) crystals, 50×2 mm², with beryllium windows, 1.5 mg cm⁻², positioned approximately symmetrically from the source holder (at various distances) was constructed. The electronic chains of a classical 4πPC–γ system were utilized. A multichannel analyzer was employed for energy calibration, and an oscilloscope for visualization of the pulses. After calibration, the system was used for the standardization of an ¹²⁵I solution. An energy window of 12–95 keV containing the entire spectrum (single and sum peak) was employed, and a simple demonstration of Taylor's equation [1967. X-ray–X-ray coincidence counting methods for the standardization of ¹²⁵I and ¹⁹⁷Hg. In: Standardization of Radionuclides: Proceedings of a International Atomic Energy Agency Symposium, Vienna, 1966, pp. 341–353, SM-79/65] for whole spectrum counting, was developed. The method and system were tested during the calibration of a LKB 2104 multi-gamma counter, by comparing the detection efficiency values obtained for this prepared standard solution and that utilized in the CIPM-CCRI(II) key comparison. The efficiency of the equipment was calculated as the mean of the two individual values. The result validates the method and demonstrates the accomplishment of the national traceability assurance in nuclear medicine for ¹²⁵I.     

Pulsed thermal neutron source at the fast neutron generator

A small pulsed thermal neutron source has been designed based on results of the MCNP simulations of the thermalization of 14 MeV neutrons in a cluster-moderator which consists of small moderating cells decoupled by an absorber. Optimum dimensions of the single cell and of the whole cluster have been selected, considering the thermal neutron intensity and the short decay time of the thermal neutron flux. The source has been built and the test experiments have been performed. To ensure the response is not due to the choice of target for the experiments, calculations have been done to demonstrate the response is valid regardless of the thermalization properties of the target.