Calibration of an automated high resolution gamma-ray spectroscopy system for in-situ real-time monitoring of a processed waste effluent stream

A pilot plant is being designed at the U. S. Department of Energy's Savannah River Site (SRS) to demonstrate the removal of ⁹⁰Sr, ¹³⁷Cs, and transuranics from a high-level liquid waste stream prior to encapsulation in a Saltstone Facility. In-line monitors are required to determine the concentration of all radionuclides on this processed waste stream. Calibration standards containing ⁶⁰Co, ¹³⁷Cs, and ⁹⁰Sr were prepared and counted. Efficiency curves were generated. Strontium-90 is readily observable above the system background in the calibration standard count, and is observable at less than 3 nCi/ml in a mixed solution having the maximum allowable concentration of all other activities present in the proposed SRS effluent stream.     

Diverse active well neutron coincidence counter utility at the Savannah River National Laboratory

In this paper we describe use of the Aquila active well neutron coincidence counter for nuclear material assays of ²³⁵U in multiple analytical techniques at Savannah River Site (SRS), at the Savannah River National Laboratory (SRNL), and at Argonne West National Laboratory (AWNL). The uses include as a portable passive neutron counter for field measurements searching for evidence of ²⁵²Cf deposits and storage; as a portable active neutron counter using an external activation source for field measurements searching for trace ²³⁵U deposits and holdup; for verification measurements of U-Al reactor fuel elements; for verification measurements of uranium metal; and for verification measurements of process waste of impure uranium in a challenging cement matrix. The wide variety of uses described demonstrate utility of the technique for neutron coincidence verification measurements over the dynamic ranges of 100–5000 g for U metal, 200–1300 g for U-Al, and 8–35 g for process waste. In addition to demonstrating use of the instrument in both the passive and active modes, we also demonstrate its use in both the fast and thermal neutron modes.     

Measurement of low concentrations of hydrogen in small samples by intermediate neutron moderation

An improved method based on the moderation of intermediate neutrons for the measurement of hydrogen in small samples is described. With the aid of boron and cadmium filters, a space shielded from slow neutrons is set up close to an isotopic neutron source shrouded by water moderator. A BF₃ proportional counter enclosed with a sample cell is placed in this space. The neutron count rate of the counter increases when a hydrogen-containing material is introduced into the cell, due to the moderation of intermediate neutrons passing through the filters. With a 1.3 μg²⁵²Cf neutron source, the lower limit of hydrogen detection for 200 ml samples in 10 min count time is 0.01 wt.%. This method is suitable for measuring the H₂O content of heavy water.     

Investigation of the effects of beam divergence on the response of neutron voluminous detectors

In the present study, the Monte Carlo calculations were carried out to evaluate the effect of beam divergence on the response of a spherical thermal-neutron counter at the center of a spherical moderating assembly. A model of ³He detector surrounded with 10 in. diameter polyethylene sphere was utilized to calculate the point isotropic and plane-parallel beams responses of twelve different energies, and obtain the geometry factor and its parameters (a ₄ and a ₅) based on general formulations. Furthermore, for 2, 3, 5, 8, 10 and 15 in. spheres exposed to three different radionuclide neutron sources with various energy spectra, the parameters of geometry factor were quantified, as a function of moderator radius and neutron energy. Obtained results based on this method were compared with the experimental data for one specific source. Finally, by these parameters the obtained values of the geometry factor based on two formulations were compared to each other.     

Measurements of the neutron spectra from the 7Li(p,n) accelerator based neutron source: Position and angular dependences

Thick target ⁷Li(p,n) neutron spectra were measured using a ³He ion chamber in the proton energy range of 1.95 to 2.30 MeV. The fast neutron spectra were collected for various distances from the lithium target as well as for various neutron emission angles. By unfolding the ³He raw data with the iterative van Cittert algorithm, the neutron fluence spectra were obtained. The ³He measured neutron spectra were compared with both analytically computed and Monte Carlo simulated spectra to account for neutron scatterings in the lithium target assembly and in the experimental area. To verify the accuracy of the neutron dose computation, the fast neutron kerma was obtained for each neutron spectrum using the fluence to kerma conversion coefficients and was compared with the measured neutron dose using tissue-equivalent proportional counters. In the position dependence investigation at the 0° emission angle, the analytically computed neutron kerma overestimates the experimental kerma by a factor of two mainly due to neutron moderation. The corresponding neutron kerma from the ³He measured spectra were in agreement with the neutron doses measured using tissue-equivalent proportional counters within 20% for lower proton energies, but the discrepancy increased to ～50% for higher proton energies. In the angular distribution investigation, a notable discrepancy between measured and computed neutron spectra were observed due to the neutron scattering effects in the target assembly and experimental room.     

Further study on NAA in characterizations of reference materials

A program was initiated at Chalk River Laboratories (CRL) to determine the physical, chemical and radiological properties of wastes intended for disposal in IRUS (Intrusion Resistant Underground Structure), a below ground vault to be constructed at CRL. One of the most restrictive radionuclides for IRUS is¹²⁹I, which has been assigned a maximum activity concentration in waste of 10⁶ Bq/m³. The limit of detection for radionuclides in waste has been set at 1% of the approximate maximum activity concentration, or 10⁴ Bq/m³ for¹²⁹I. A radiochemical instrumental neutron activation analysis method has been developed to determine¹²⁹I in two waste streams, incinerator ash and liquid feed to a bituminizer. Solid samples are spiked with¹²⁵I tracer, fused at 960°C with Li₂B₄O₇ in a platinum boat in a flowing oxygen stream inside a three zone tube furnace, and the volatilized I₂ is trapped on in-line charcoal filters. The charcoal filters are irradiated together with a filter containing a spiked¹²⁵I/¹²⁹I standard, in the NRU reactor, and then subjected to post-irradiation chemistry to remove⁸²Br interference. The¹²⁹I concentration in the sample is determined by comparing the activity of the activated¹³⁰I in the sample with that of the standard, and the chemical recovery for¹²⁹I is determined from the activity of¹²⁵I tracer. Limits of detection for¹²⁹I in solids are typically 0.005 Bq/g, based on a 4 hour counting period on a 10% efficient HPGe gamma-spectrometer at a source to detector distance of approximately 12 cm. This paper presents a summary of the method and the results from analysis of two waste streams.     

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.     

A borehole sonde using neutron activation technique

A simulated borehole sonde has been assembled, with an aluminium casing of 70 cm in length and 12 cm in inner diameter. It contains a 5 Ci Pu–Be source with a neutron yield of about 5.45·10⁶+10% n·cm^–2·s^–1, a shadow shield, and a Hp Ge detector of 15% efficiency and 2 keV FWHM for the 1.33 MeV line of⁶⁰Co. Evaluations of the assembly through the dependence of thermal neutron flux on water content, matrix composition and borehole configuration have been performed. Accordingly, thermal neutron flux distributions have been measured around the sonde and inside the ore in both the simulated dry and water filled borehole. From these measurements one could estimate that the effective moderating thickness of water is about 4 cm, while the volume matrix of the ore to be investigated by this assembly is a slab of about 8 cm width and a height of about 15 cm. It also follows that the uranium-thorium ore analysis method described in this work may become important as a field neutron activation technique.This work was financially supported by the IAEA under research project No. 3534/R.B.     

Distribution behavior of U(VI), Am(III) and Eu(III) on diglycolamide based extraction chromatographic resin in perchloric acid medium

An attempt has been made in the present work to investigate the role of anion for the uptake of Am(III)/Eu(III)/U(VI) by extraction chromatography (EXC) resin incorporating tetra-n-octyl-3-oxapentanediamide, commonly referred to as tetra-octyl diglycolamide (TODGA). In contrast to the nitric acid, perchloric acid medium favors extraction of trivalent metal ions even at low acidity (pH 2) and is almost insensitive to the acidity up to 5 M. Exceptionally large distribution coefficients (10⁵–10⁶) in the wide range of perchlorate concentration (10^?2–5 M) is quite unusual and is by far the largest reported in the literature for Am(III)/Eu(III). Thermodynamic data suggests the possibility of inner sphere/cation exchange mechanism involving TODGA aggregates at higher acidity but outer sphere/cation exchange mechanism at low acidity for Eu(III). There is a possibility of employing TODGA based EXC resin for the remediation of liquid waste (contaminated with long lived transuranics like ^241/243Am and ²⁴⁵Cm) in the wide range of acidity.     

Determination of water in zeolites by a neutron reflection method

Water content in zeolites has been determined by an improved neutron reflection method using a Pu-Be neutron source of 10⁶ n·s^–1 intensity and a BF₃ counter. It was found that the water content of different types of zeolites collected in Hungary varies between 9 and 12 wt.%. The standard deviation of the determination does not exceed 0.5 wt.%. The matrix effect on the sensitivity and accuracy was also studied. An approximate relation is given between the count rate and the neutron physical parameters of the samples.     

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.     

Measurement of moisture and hydrogen contents by intermediate neutron moderation

An intermediate neutron moderation method for measurement of moisture and/or hydrogen contents of small samples is presented. The sample is placed on the top face of a neutron howitzer, with a cadmium sheet between. Thermal neutrons resulting from intermediate neutron moderation in the sample are detected with a³He proportional counter placed on the sample, by a cadmium difference method. With a 500 mCi Am-Be neutron source, the limit of moisture detection for a 10×20×1.8 cm³ asbestos plate in 1 min count time is 0.5 wt.%. The precision of measuring the hydrogen contents of 250 ml hydrocarbons containing 112 mg H/ml is 0.9% under the same conditions.     

Neutron induced reactions at the Athens Tandem Accelerator NCSR “Demokritos”

The neutron facility at the 5.5 MV tandem T11/25 Accelerator of NCSR “Demokritos” can deliver monoenergetic neutron beams in the energy range from thermal to 450 keV, 4–11.5 MeV and 16–20.5 MeV via the ⁷Li(p,n), ²H(d,n) and ³H(d,n) reactions, respectively. The flux variation of the neutron beam is monitored by using a BF₃ counter and a liquid scintillator BC501A detector. The ²³²Th(n,2n)²³¹Th and ²⁴¹Am(n,2n)²⁴⁰Am as well as (n,2n), (n,p) and (n,α) reactions on natural Ge and Hf isotopes, have been investigated from threshold up to 11.5 MeV, by using the activation method. The cross section values have been determined relative to the ¹⁹⁷Au(n,2n)¹⁹⁶Au, ²⁷Al(n,α)²⁴Na and ⁹³Nb(n,2n) reference reaction cross sections.     

Measurement of the radionuclide 24Na produced by neutron exposure to the body

Three workers were exposed to neutrons at a criticality accident at a uranium conversion test plant. The (n,g) reaction with body sodium (²³Na) gave rise to ²⁴Na, which emits 1,369 keV and 2,754 keV gamma-rays. One of the workers was measured with a whole-body counter to estimate the specific activity of ²⁴Na [i.e., ²⁴Na (Bq)/²³Na (g)] in the body. The estimated specific activity (9,510 Bq^.g^-1) was converted to neutron dose, using a dose coefficient (0.066 mGy per specific activity). The method was useful for the retrospective estimation of neutron doses. This revised version was published online in July 2006 with corrections to the Cover Date.     

Analysis of 99Tc in the radioactive liquid waste after extraction into suitable solvent

⁹⁹Tc is one of the long lived fission product with high fission yield. From radioactive waste management point of view it is very much essential to evaluate the concentration of technetium in the radioactive liquid waste in order to finalise the treatment process to extract/isolate it from the stream which is discharged to the environment. For the estimation of ⁹⁹Tc in the radioactive liquid waste stream, extraction of the stable complex of technetium-tetraphenyl arsonium chloride (TPAC) into chloroform followed by beta counting was studied. Various parameters like pH, time of equilibration, concentration of TPAC in chloroform, use of other solvent for extraction as well as interference of various other radionuclides present in the waste were also studied. The radioactive liquid waste being handled in plant contains high concentrations of salts in the form of sodium nitrate. Hence effect of salt concentration on the percentage extraction was also evaluated. The extraction behavior does not dependent on change in the pH of the solution. Almost 99.5% extraction was observed in the pH range of 1?C13.0. High concentration of salt is affecting the extraction. However, this can be taken care by diluting the radioactive waste. It takes almost 90?min time for maximum extraction. Presence of radionuclides like ¹³⁷Cs, ⁹⁰Sr are not interfering the extraction of ⁹⁹Tc. However, ¹⁰⁶Ru is getting slightly extracted along with ⁹⁹Tc. The error due to ¹⁰⁶Ru can be eliminated by taking gamma spectrum and deducting the activity from the total beta activity to get ⁹⁹Tc activity. Nitrobenzene can be used for extraction of Tc?CTPAC complex in place of chloroform.     

Radioactivity measurement with a plastic scintillation vial

Liquid scintillation counting is widely used to measure radioactivity, but it generates radioactive organic liquid waste. Not to generate the liquid waste using a liquid scintillation counter, novel counting method with a plastic scintillation vial was designed. The counting efficiency for ³²P was 10–40% and that for ¹²⁵I was 4–8%. The efficiency depended on the sample volume. The color quenching effect was negligible. No radioactive liquid waste was generated by this method. In addition, you can reuse the measured sample.     

The application of a252Cf neutron multiplier for routine instrumental neutron activation analysis

The successful application of instrumental neutron activation analysis for routine determinations depends on the ability to produce accurate and precise analytical results in a relatively short time. An important factor in obtaining the desired speed has been the availability of a low-cost, moderate-flux neutron source for on-site use. The²⁵²Cf neutron multiplier (CFX), designed and constructed by Intelcom Rad Tech Corporation of San Diego, California, is a subcritical assembly capable of continuous, stable operation and has provided us with the ability to determine more than 35 elements as major and minor components. The CFX produces a thermal neutron flux of ∼2×10³ n/cm²-sec by a 100-fold multiplication of the neutrons emitted from a 1 mg²⁵²Cf source. Of particular importance in its application at Kodak has been the determination of the halogens Cl, Br, and I, both singly and simultaneously, in various matrices including photographic materials.     

Neutron activation analysis of low level lithium in water samples

For determining low level lithium concentrations in water, a neutron activation method based on the measurement of tritium radioactivity produced by⁶Li(n,)³H reaction has been developed. This method is specific and free from interference by other chemical elements. Using a low background liquid scintillation counter for tritium measurement, the detection limit is approximately 0.3 ppm during irradiation at a thermal neutron flux density of 1.1·10⁷n·cm^–2·s^–1 for 6 hours by a small nuclear reactor and liquid scintillation counting for 2000 minutes     

Determination of <Superscript>129</Superscript>I using volatilization method and liquid scintillation spectrometry

A simple and rapid separation method for ¹²⁹I determination in radioactive waste samples was developed. Suitable conditions for iodine volatilization were tested. Iodine was trapped in 1.5 mol L^?1 NaOH and precipitated as PdI₂·H₂O by addition of PdCl₂ with recoveries higher than 80%. The method was applied for analysis of contaminated soil, radioactive sludge, evaporator concentrate and heterogeneous waste samples from nuclear power plants in Slovak Republic. ¹²⁹I was measured on liquid scintillation counter TRI CARB 2900 TR using Ultima Gold AB scintillation cocktail.     

Determination of manganese in ores by activation analysis using a californium-252 neutron source

A²⁵²Cf neutron source has been used to analyse manganese in ores such as pyrolusite, rodonite (manganese silicate) and blends used in dry-batteries. Samples with about 150 mg and standards of manganese dioxide were irradiated for about 20 min and counted using a well-type NaI(Tl) scintillation counter and scaler, with or without pulse-height discriminator between the detector and the scaler. The interferences of nuclear reactions⁵⁶Fe(n,p)⁵⁶Mn and⁵⁹Co(n,α)⁵⁶Mn were studied, as well as problems in connection with neutron shadowing during irradiation, gamma-rays attenuation during counting and influence of granulometry of samples. Some of the samples were also analysed by wet-chemical method (sodium bismuthate) in order to compare results.