Evaluation of the radioactivity in concrete from accelerator facilities

For evaluation of radioactivity induced in the concrete samples from accelerator facilities, the residual radioactivity in concrete sample, collected from seven accelerator facilities, was determined by γ-ray spectrometry. The tritium was extracted by the heating method using an IR furnace, and measured with a liquid scintillation counter. It was found that the major radioisotopes activated mainly by neutrons in the concrete samples were ¹⁵²Eu, ⁶⁰Co, ¹³⁴Cs and ³H. The concentrations of radioactivities induced by thermal neutron capture are the highest at a depth of 10 cm in the concrete wall. The correlation between tritium, ⁶⁰Co and ¹⁵²Eu activity was investigated by measuring many concrete samples for seven accelerator facilities. The results indicate that their activities are strongly correlated with each other. So it would also be concluded that the total activity in shielding concrete could be estimated on the basis of the activities of ⁶⁰Co and ¹⁵²Eu.     

Neutron activation analysis for determination of induced radioactivity in concrete of nuclear reactor shielding

The content of binding agent components and aggregates of shielding concrete (portland cement, gabbro, silicate, granite, hematite, magnetite, limestone) has been studied to identify nuclides contributing to the long-lived radioactivity of concretes used in nuclear reactor shielding. Elemental content was determined by neutron activation analysis at the IBR-2 reactor of the Laboratory of Neutron Physics (JINR). It is shown that the nuclides mainly responsible for the long-lived activity are¹⁵²Eu,¹⁵⁴Eu,⁶⁰Co and¹³⁴Cs. Based upon the experimental data obtained, calculations of the specific activity of concrete ingredients being irradiated for 30 years by a unit neutron flux density have been performed, and diagrams of the dependence of the specific activity versus time are presented. It is recommended that the application of such media as magnetite, serpentinite, chromite to shielding structures should be abandoned. Estimations of the dose rate from concrete structures containing various aggregates for the limiting case of semi-infinite space are given. The results obtained can form the basis for the optimal choice of the materials when constructing the concrete shielding of neutron sources.     

Determining trace amounts of nickel in plant samples by neutron activation analysis

Neutron activation analysis of plant samples for trace amounts of nickel has applications in various fields. Landsberger and Robinson (Trans Am Nucl Soc 102:187–188, 2010) found their measurement of the concentrations of nickel for different NIST reference materials to be significantly greater than the certified values when measuring nickel from the 810.8 keV gamma ray that comes from ⁵⁸Co from the ⁵⁸Ni(n,p) reaction. They determined that this overestimation was due to a significant interference by the presence of ¹⁵²Eu at 810.5 keV, and presented a method for correcting this interference. Their method involved a long thermal irradiation and correction of gamma ray counts based on the 1,408 keV belonging to ¹⁵²Eu. This paper presents an alternative approach, which involves irradiating the samples with epithermal neutrons, which suppressed the ¹⁵²Eu to the point of being negligible for samples with low levels of europium. Both methods were determined to work well for the identification of nickel concentrations by neutron activation analysis.     

Measurement of k0-Factors in Prompt Gamma-Ray Neutron Activation Analysis

The guided thermal neutron beam at 100 MW Dhruva research reactor facility of Bhabha Atomic Research Centre (BARC) was used to carry out prompt gamma-ray neutron activation analysis (PGNAA). The prompt k ₀-factors have been determined for the isotopes of the elements H, B, K, Co, Cu, Ca, Ti, Cr, Cd, Ba, Hg and Gd with respect to 1951 keV gamma-line of ³⁶Cl. The prompt k ₀-factors for H, Cl and Cu were also measured with respect to the 1381 keV gamma-line of ⁴⁹Ti. Different samples like NH₄Cl, Ti metal, cobalt chloride and other stoichiometric compounds and pure metals were used for this purpose. Prompt gamma-rays were accumulated using a 22% HPGe detector connected to a PC based 8k MCA in single mode counting. The energy calibration in the range of 100–8500 keV was carried out using gamma-rays from ¹⁵²Eu and ⁶⁰Co, and the prompt gamma-rays from ³⁶Cl whereas the absolute detection efficiency for this energy range was determined using ¹⁵²Eu and prompt gamma-rays from ³⁶Cl and ⁴⁹Ti.     

Tritium activity induced in the accelerator building and its correlation to radioactivity of gamma-nuclides

An infrared furnace (ULVAC RHL-410P) was newly applied to the extraction of tritium from concrete samples. After studying the tritium recovery yield regarding temperature and time, the best extraction conditions were set to 800 °C (setting temperature) for 30 minutes under Ar-gas flow of 200 ml/min. Tritium was collected in two cold traps and transferred to a vial for liquid scintillation counting. It took about one hour for the extraction of tritium. Reproducibility and recovery yield of tritium were about 100% compared to the values obtained by the ordinary heating method using an electric furnace. Gamma-ray emitters and tritium of concrete samples collected from several accelerator facilities have been determined. The specific activity of tritium strongly correlated with that of ¹⁵²Eu and ⁶⁰Co, so it was found that tritium was produced by thermal neutron reaction by the ⁶Li(n,)³H reaction. The results indicate that the tritium specific activity in concrete can be estimated from the ⁶⁰Co specific activity obtained easily by -ray measurement.On leave from IHEP, Beijing.This work was supported by Japan Society for the Promotion of Science. The authors also are grateful to Dr. S. Watanabe, Center for Nuclear Study, University of Tokyo, Dr. T. Ohtsuki, Laboratory of Nuclear Science, Tohoku University, and Dr. T. Saito, Research Center for Nuclear Physics, Osaka University, and their colleagues for concrete sampling.     

Precise assay of weight per cent sodium in high-level processed nuclear tank waste using a <Superscript>252</Superscript>Cf neutron source

An infrared furnace (ULVAC RHL-410P) was newly applied to the extraction of tritium from concrete samples. After studying the tritium recovery yield regarding temperature and time, the best extraction conditions were set to 800 °C (setting temperature) for 30 minutes under Ar-gas flow of 200 ml/min. Tritium was collected in two cold traps and transferred to a vial for liquid scintillation counting. It took about one hour for the extraction of tritium. Reproducibility and recovery yield of tritium were about 100% compared to the values obtained by the ordinary heating method using an electric furnace. Gamma-ray emitters and tritium of concrete samples collected from several accelerator facilities have been determined. The specific activity of tritium strongly correlated with that of ¹⁵²Eu and ⁶⁰Co, so it was found that tritium was produced by thermal neutron reaction by the ⁶Li(n,)³H reaction. The results indicate that the tritium specific activity in concrete can be estimated from the ⁶⁰Co specific activity obtained easily by -ray measurement.On leave from IHEP, Beijing.This work was supported by Japan Society for the Promotion of Science. The authors also are grateful to Dr. S. Watanabe, Center for Nuclear Study, University of Tokyo, Dr. T. Ohtsuki, Laboratory of Nuclear Science, Tohoku University, and Dr. T. Saito, Research Center for Nuclear Physics, Osaka University, and their colleagues for concrete sampling.     

Decontamination of radionuclides on construction materials

A wide variety of materials can become contaminated by radionuclides, either from a terrorist attack or an industrial or nuclear accident. The final disposition of these materials depends, in large part, on the effectiveness of decontamination measures. This study reports on investigations into the decontamination of a selection of building materials. The aim has been to find an effective, easy-to-use and inexpensive decontamination system for radionuclides of cesium and cobalt, considering both the chemical and physical nature of these potential contaminants. The basic method investigated was surface washing, due to its ease and simplicity. In the present study, a basic decontamination formulation was modified by adding isotope-specific sequestering agents, to enhance the removal of cesium(I) and cobalt(II) from such construction materials as concrete, marble, aluminum and painted steel. Spiking solutions contained ¹³⁴Cs or ⁶⁰Co, which were prepared by neutron activation in the SLOWPOKE-2 nuclear reactor facility at the Royal Military College of Canada. Gamma spectroscopy was used to determine the decontamination efficiency. The results showed that the addition of sequestering agents generally improved the radiological decontamination. Although the washing of both cesium and cobalt from non-porous materials, such as aluminum and painted steel, achieved a 90–95 % removal, the decontamination of concrete and marble was more challenging, due to the porous nature of the materials. Nevertheless, the removal efficiency from 6-year-old concrete increased from 10 % to approximately 50 % for cobalt(II), and from 18 to 55 % for cesium(I), with the use of isotope binding agents, as opposed to a simple water wash.     

Environmental radiation and radioactivity in the vicinity of the Semipalatinsk Test Site

The International Atomic Energy Agency's Mission to Kazakhstan studied the present environmental contamination and the exposure risk from nuclear tests performed at the Semipalatinsk Test Site. On this mission, EML scientists collected in-situ gamma-ray spectra, measured external gamma dose rates, and collected soil samples. With the exception of plowshare areas and near ground zero, all the areas visited had external dose rates within typical environmental levels. Measurements within a 15 km radius of ground zero showed elevated levels of¹³⁷Cs,¹⁵²Eu, and⁶⁰Co. The dose rate within ≈1 km of ground zero ranged from 500 to 30000 nGy·h⁻¹.     

Effect of humic acid on absorption-release processes in the bottom sediments-Yenisei river water system as studied by dual-column ion chromatography and γ-ray spectrometry

The effect of humic acid on absorption-release processes in the bottom sediments-Yenisei river water system was studied by dual-column ion chromatography and γ-ray spectrometry. With the use of ion chromatography, it was found that processes related to the absorption of SO ₄ ^2? and Cl^? anions by a solid phase with the release of NO ₃ ^? , PO ₄ ^3? , and F^? to a liquid phase competed in the test systems as the concentration of water-soluble organic carbon (WSOC) was increased. Only the test anions were released in the systems without the introduction of an additional amount of WSOC as humic acid. With the use of γ-ray spectrometry, it was found that the release of ⁶⁰Co, ¹⁵²Eu, and ²⁴¹Am radionuclides to the liquid phase in the systems with added humic acid began much earlier than in the system without the addition of humic acid. In this case, the amount of released radionuclides was greater than the amount of radioisotopes released in the system without the addition of humic acid: ～25% ²⁴¹Am, ～3% ¹⁵²Eu, and ～0.8% ⁶⁰Co in the system with added humic acid or 0.8% ¹⁵²Eu and <0.1% ⁶⁰Co in the system without the addition of humic acid. The ²⁴¹Am radionuclide was not determined in the system without the addition of humic acid. An increase in the concentration of WSOC in the experimental system of bottom sediments-Yenisei river water initiated the release of ⁶⁰Co, ¹⁵²Eu, and ²⁴¹Am anthropogenic radionuclides from bottom sediments because of the formation of soluble complexes capable of migration. An increase in the concentration of WSOC had almost no effect on the release of ⁴⁰K and ¹³⁷Cs radionuclides.     

Radiochemical study of Cd,Co and Eu coprecipitation with iron hydroxide in seawater

The co-precipitation of Cd, Co and Eu from seawater onto ferric hydroxide was studied, using the radioactive tracers¹⁰⁹Cd,⁶⁰Co and^152–154Eu, as a function of the pH, the concentration of the collector iron, the concentration of the metal spike and the aging of the internally formed precipitate. For preconcentration purpose, a pH 9 and a 35 ppm iron concentration appeared to be suitable. At pH 8, the natural pH of seawater, the distribution coefficients for Cd, Co and Eu are 10^4.5, 10⁵ and 10⁷, respectively.On leave from: Department of Chemistry, Rammohan College, Calcutta, India.     

Utilization of low-cost sorbent for removal and separation of 134Cs, 60Co and 152+154Eu radionuclides from aqueous solution

Eggshell material was used as low-cost and eco-friendly biosorbent for removal of ¹³⁴Cs, ⁶⁰Co and ^152+154Eu radionuclides from aqueous solution. The eggshell material was calcined at 500 and 800 °C, and then characterized. Comparative studies on the natural and calcined eggshell for sorption of the three radionuclides were carried out. It was found that, the uptake is in the order: Eu(III) > Co(II) > Cs(I). Further, column chromatography was used in separation of ¹³⁴Cs, ⁶⁰Co and ^152+154Eu using 0.15, 0.2 and 0.5 mol/l nitric acid, respectively. Eggshell material can be considered as a promising material for separation of radionuclides from radioactive waste solution.     

Radiological dose assessment of naturally occurring radioactive materials in concrete building materials

Previous studies have shown that the natural radioactivity contained in building materials have significantly influenced the dose rates in dwelling. Exposure to natural radiation in building has been of concerned since almost 80 % of our daily live are spend indoor. Thus, the aim of the study is to assess the radiological risk associated by natural radioactivity in soil based building materials to dwellers. A total of 13 Portland cement, 46 sand and 43 gravel samples obtained from manufacturers or bought directly from local hardware stores in Peninsular of Malaysia were analysed for their radioactivity concentrations. The activity concentrations of ²²⁶Ra, ²³²Th and ⁴⁰K in the studied building materials samples were found to be in the range of 3.7–359.3, 2.0–370.8 and 10.3–1,949.5 Bq kg^?1 respectively. The annual radiation dose rates (μSv year^?1) received by dwellers were evaluated for 1 to 50 years of exposure using Resrad-Build Computer Code based on the activity concentration of ²²⁶Ra, ²³²Th and ⁴⁰K found in the studied building material samples. The rooms modelling were based on the changing parameters of concrete wall thickness and the room dimensions. The annual radiation dose rates to dwellers were found to increase annually over a period of 50 years. The concrete thicknesses were found to have significantly influenced the dose rates in building. The self-absorption occurred when the concrete thickness was thicker than 0.4 m. Results of this study shows that the dose rates received by the dwellers of the building are proportional to the size of the room. In general the study concludes that concrete building materials; Portland cements, sands, and gravels in Peninsular of Malaysia does not pose radiological hazard to the building dwellers.     

Determination of trace amounts of gold in the presence of rare earth elements in rock samples from Makhtesh Ramon (Southern Israel), by instrumental epithermal neutron activation analysis

Epithermal neutron activation analysis (ENAA), followed by high resolution gamma-ray spectrometry, was applied to determine trace amounts of Au in the presence of rare earth elements (REE) from vein samples in the basaltic rocks of Makhtesh Ramon, located in southern Israel. The contribution of¹⁵²Eu (411.1 KeV) to the 411.8 keV peak of¹⁹⁸Au was determined using multiple gamma-peak, ratios derived from Eu standards and mixtures of Au and Eu. The concentration of Au was found to be in the range of 10–80 ppb. A group of rare earth elements: La, Eu, Ce, Tb, Sm, Lu, Yb was identified; the concentration of Eu was found to be 0.5 ppm.     

Determination of radionuclides in human and environmental samples from the criticality accident in Tokai-mura, Japan

For three heavily exposed workers in the Tokai-mura criticality accident, the levels of ²⁴Na, ⁴²K and ⁸²Br were measured in blood, urine, vomit and hair. ²⁴Na showed the highest activity and was found to have a specific activity of 8.24·10⁴, 4.34·10⁴ and 1.23·10⁴ Bq ²⁴Na per g Na in blood of the three exposed workers. Based on these data the individual neutron fluences were calculated. Radionuclides were also measured in environmental samples from the JCO campus. The fission products ¹³¹I, ¹³³I, ¹⁴⁰Ba-¹⁴⁰La and ¹³⁷Cs were determined in plant samples and the activation products ⁵⁹Fe, ⁴⁶Sc, ⁶⁰Co and ¹⁵²Eu were identified in soil samples. ²³⁵U/²³⁸U ratios in soil and plant samples were measured by ICP-MS. This revised version was published online in July 2006 with corrections to the Cover Date.     

Radiochemical separation and determination of europium by Ge(Li) detector in biological tissues

A simple neutron activation method has been developed for the determination of europium in biological tissues and applied in the analysis of marine organism samples with ±9% precision at the nanogram level. The method is based on the separation, by ion-exchange, of the rare earth group from dry or ashed irradiated tissues and subsequent determination of^152mEu, by γ-spectrometry using a lithium drifted germanium detector.^152mEu, separated almost completely from other than rare earth elements, with better than 98% chemical yield, is counted on the 121.8 keV photopeak which then is practically free from any other γ-ray energy interfering in this counting.     

Monitoring of environmental radiation on the Spratly Islets in the South China Sea

The concentration of radionuclides in environmental samples and the environmental radiation level (including cosmic rays) was monitored on Spratly Islets in South China Sea prior to the commercial operation of nearby Chinese nuclear power plants. Samples of coral sand, sea food, vegetation, water, and accumulated fallout on the islet were obtained and measured for radioactivity. Except for some exposure originating from minute amounts of⁶⁰Co and¹³⁷Cs, the radiation background is mainly emitted from naturally occurring radionuclides. Radiation levels including beta, gamma, and neutron on the islet have also monitored. Gamma dose rates for outdoor exposure averaged approximately 0.09 Sv/h; the neutron dose rate was about 0.003 Sv/h. Automatic radiation surveillance and a routine sampling program on the Spratly Islets are viewed as essential to provide an early warning in the Far East in case of a nuclear emergency.     

Prompt gamma neutron activation analysis of boron with a 241Am-Be neutron source

A prompt gamma neutron activation analysis (PGAA) setup installed at ANRTC has been used to analyze boron. It consists of a 22.6% REGe detector and a 740 GBq ²⁴¹Am-Be neutron source moderated with water and paraffin. At the sample irradiation position, the thermal neutron fluence rate measured was 2.36·10⁴ n·m^–2· s^–1 and the corresponding Cd-ratio was 22 for gold monitor. The absolute detection efficiency in the range of 120–1500 keV was determined using ¹⁵²Eu standard solution. The sensitivity and detection limit for standard boric acid samples has been determined. The boron content in boric acid prepared from Turkish borate ores is measured to be 15.91±0.46% wt.     

Prompt gamma neutron activation analysis of boron with a 241Am-Be neutron source

A prompt gamma neutron activation analysis (PGAA) setup installed at ANRTC has been used to analyze boron. It consists of a 22.6% REGe detector and a 740 GBq ²⁴¹Am-Be neutron source moderated with water and paraffin. At the sample irradiation position, the thermal neutron fluence rate measured was 2.36·10⁴ n·m^–2· s^–1 and the corresponding Cd-ratio was 22 for gold monitor. The absolute detection efficiency in the range of 120–1500 keV was determined using ¹⁵²Eu standard solution. The sensitivity and detection limit for standard boric acid samples has been determined. The boron content in boric acid prepared from Turkish borate ores is measured to be 15.91±0.46% wt.     

Simultaneous determination of 152Eu and 241Am in liquid solution by liquid scintillation counting

¹⁵²Eu and ²⁴¹Am are the most frequently used radiotracers in the separation studies on trivalent minor actinides and lanthanides. In almost all those studies, the determination of ¹⁵²Eu and ²⁴¹Am has been based on measuring their γ radiation by using a NaI(Tl) scintillation detector and/or a germanium detector. In this study, based on measuring the β particles and mono-energy electrons from ¹⁵²Eu and the α particles from ²⁴¹Am, we provide a new option to simultaneously determine the radioactivities of ¹⁵²Eu and ²⁴¹Am by liquid scintillation counting (LSC) with the aid of α/β discrimination. If the count rate ratio of ²⁴¹Am to ¹⁵²Eu is within the range of 100:1–1:100, the radioactivities of ¹⁵²Eu and ²⁴¹Am in mixed samples can be simultaneously determined by LSC with the errors less than 5 %. In addition, the interferences of ²⁴¹Am on Eu are divided into two parts: inside and outside the ²⁴¹Am region of interest. Only if the count rate ratio of ²⁴¹Am to Eu is more than 10:1, should the latter interference be in consideration.     

Neutron Equivalent Dose Rate Measurements of Gypsum-Waste Tire Rubber Layered Structures

The neutron equivalent dose rates (µSv/h) of gypsum, steel-reinforced rubber waste tire, and gypsum-waste tire rubber sandwich composite samples were investigated. Prepared samples were irradiated with ²⁴¹Am-Be neutrons and transmission values were obtained using dose equivalent rates measured with a BF₃ neutron detector. Results were compared to those of concrete, and as a result of neutron shielding, the performance of gypsum, waste tire, and waste tire (steel-reinforced rubber) embedded gypsum samples was higher than that of concrete. This information may be useful for shielding design of nuclear application areas.