Application of multiple γ-ray detection to long-lived radioactive nuclide determination in environmental samples

ABSTRACT

Gamma-gamma coincidence measurement utilized in γ-ray spectroscopy experiments is well known to be effective for the improvement of signal-to-noise ratio in a γ-ray spectrum. We study its applicability to the determination of long-lived radioactive nuclides in environmental samples. The γ-ray simulation code Geant 4.10.2 was used. A conventional and effective detector system comprising five Ge detectors was assumed. We took up 38 nuclides which need to be determined for the evaluation of fission product leakage at the nuclear accident in the Fukushima nuclear power plants in Japan. Among them 12 nuclides emit γ-rays and five nuclides of ⁶⁰Co, ⁹⁴Nb, ¹³⁴Cs, ¹⁵²Eu, and ¹⁵⁴Eu can be the objectives of the multiple γ-ray detection methods. The simulation results indicate that the signal-to-noise ratio can be improved by a factor between 9.84 and 283, and the detection limit by a factor between 2.71 and 8.53 relative to the singles measurement, implying that the method can be well applied to the determination of the long-lived radioactive nuclides and will provide a quick and non-destructive analysis method.     

Estimation of the activation of local reactor shielding concretes

Many elements in reactor shielding concrete become radioactive due to interaction with neutrons during the operation of a nuclear reactor. These radioactive elements will build up over the time due to (n,γ) reaction and contribute to the radioactive waste during decommissioning of the reactor, they will increase the dose from the inner part of the biological shielding (concrete) surrounding a nuclear reactor during maintenance works within reactor containment, and their effect should be taken in the calculations of the dose behind the concrete shield.Six mixtures of local ordinary and special shielding concretes have been investigated in this study. Their compositions were determined using X-ray diffraction and neutron activation analysis using the Syrian Miniature Neutron Source Reactor (MNSR) and consequently the dose rates.Based on the results of the analyses and comparing to the published limits in the references, it has been found that ⁶⁰Co, ¹⁵²Eu, ¹⁵⁴Eu and ¹³⁴Cs will be the most effective radioactive nuclides existing in the local concrete, but Dolomite aggregates contain the minimum limits of them and it has been found that Fe, Mg, and Ti the major elements existing in the local concrete.     

Long-lived activation products in TRIGA Mark II research reactor concrete shield: calculation and experiment

In this paper, a process of long-lived activity determination in research reactor concrete shielding is presented. The described process is a combination of experiment and calculations. Samples of original heavy reactor concrete containing mineral barite were irradiated inside the reactor shielding to measure its long-lived induced radioactivity. The most active long-lived (γ emitting) radioactive nuclides in the concrete were found to be ¹³³Ba, ⁶⁰Co and ¹⁵²Eu. Neutron flux, activation rates and concrete activity were calculated for actual shield geometry for different irradiation and cooling times using TORT and ORIGEN codes. Experimental results of flux and activity measurements showed good agreement with the results of calculations. Volume of activated concrete waste after reactor decommissioning was estimated for particular case of Jo?ef Stefan Institute TRIGA reactor. It was observed that the clearance levels of some important long-lived isotopes typical for barite concrete (e.g. ¹³³Ba, ⁴¹Ca) are not included in the IAEA and EU basic safety standards.     

Activation Analysis of Gadolinium Impurity in High Purity Europium Oxide

Abstract

An analytical method for determining Gd impurity in high purity Eu₂O₃ is proposed, which makes use of neutron activation and cation-exchange separation to examine its suitability as target material for the production of ^{152 m, 152, 154}Eu.

Long-term irradiation of an Eu sample resulted in ¹⁵³Gd activity amounting to 1.8 times that of the same nuclide produced from an equal quantity of Gd by (n, γ) reaction. This experimental value is quite consistent with that obtained by calculation under the assumption that the ¹⁶³Gd results from nuclear reaction on ¹⁵¹Eu induced by secondary neutrons.

For the accurate determination of Gd, Gd impurity was separated from the Eu sample prior to neutron irradiation in order to reduce the self-shielding effect in the Eu sample. Separation by cation-exchange with α-hydroxyisobutyrate (0.33 M, pH 3.77) made it possible to reduce the content of Eu in the Gd fraction below 7×10^?3%. This sufficed to assure that the ¹⁵⁹Gd content in ^152mEu was smaller than 1×10^?3%. For the determination of Gd content below 10⁴ ppm, however, should necessitate further purification of the irradiated Gd fraction.     

Towards more precise Be and Cl data from measurements at the 10 level: Influence of sample preparation

The spreading application of accelerator mass spectrometry (AMS) to the geosciences will require measurement of increasing numbers of samples with low ¹⁰Be/⁹Be and ³⁶Cl/Cl isotopic ratios. To distinguish radionuclide concentrations in samples from corresponding processing blanks, samples must be prepared using stable isotope carriers with low intrinsic radionuclide concentrations. Measurements at different AMS facilities have shown that commercially available ⁹Be carriers rarely meet these requirements and use of ⁹Be carriers prepared from Be-containing minerals such as Be₂SiO₄ is advisable. For precise determinations at the low 10⁻¹⁴ level, samples need to produce the highest ⁹Be currents possible. Measurements performed at the two AMS facilities in France show that mixing BeO with Nb powder generally yields higher ⁹Be currents than mixing with Ag powder. Contamination of BeO by other elements such as Ti or Al will reduce the current by a simple dilution effect, thus limiting the statistical precision with which ¹⁰Be can be determined. In the case of ³⁶Cl analyses, repeated water-leaching of calcite samples sufficiently removes all atmospheric ³⁶Cl contamination, allowing determination of terrestrial cosmogenic in situ produced ³⁶Cl concentrations for surface exposure dating.     

Sensitivity of charged particle activation analysis for long-lived radioactive nuclide determination

ABSTRACT

Charged Particle Activation Analysis (CPAA) utilizing an 8-MeV proton beam has been studied for determination of 35 long-lived radioactive nuclides. We accumulated the reaction cross section and nuclear decay data by referring to nuclear database supplied by National Nuclear Data Center in Brookhaven National Laboratory. We also calculated the reaction cross sections by using statistical model code ALICE. By using the nuclear data, we have derived determination sensitivity of the radioactive nuclides relative to unit weight and specific radioactivity. The result indicates that several hardly measurable nuclides with long half-lives such as ¹³⁵Cs, ²⁴⁴Pu, ¹²⁹I, ¹²⁶Sn, ⁹³Mo, ¹⁰⁷Pd, ²³⁶U, ²⁴⁸Cm, and ²³⁷Np have high sensitivity. It may be concluded that CPAA can be applied to determination of several long-lived nuclei and will provide a quick and non-destructive analysis method.     

Nuclear Data Sheets for A = 36

Nuclear spectroscopic information for experimentally investigated nuclides of mass 36 (Mg, Al, Si, P, S, Cl, Ar, K, Ca) has been evaluated. The principal sources of the ‘adopted levels’ presented for nuclides close to the stability line are Endt?s evaluations (1990En08, 1978En02). The data sets for reactions and decays, including all available gamma-ray data, are based mostly on the original literature. The ³⁶Na has been looked for but not yet experimentally detected. There are no data available for the excited states in ³⁶Al, and for ³⁶Mg and ³⁶Ca, only one excited state is known.     

Trace Elements with Large Activation Cross Section in Concrete Materials in Japan

Amounts of trace elements with large activation cross section in concrete materials were measured to offer the basic data for developing of low activation concrete. From the measurements, the quantities of the activated radioactivities in biological shielding concrete were measured and evaluated for the clearance level.

The average concentrations of ⁶⁰Co, ¹⁵²Eu and ¹³⁴Cs formed in concrete were 21.9, 1.08 and 3.21 ppm, respectively. The combination of the concrete materials for the most lowering concentrations of ⁶⁰Co, ¹⁵²Eu and ¹³⁴Cs was the limestone as aggregate and the white Portland cement produced in specific places. The most lowering concentrations of this limestone concrete were 0.16, 0.049 and 0.060 ppm, respectively. The limestone concrete was excellent as biological shielding concrete, because the neutron shielding effect was excellent a little compared with ordinary concrete. If this concrete used for biological shielding concrete, concrete waste will be able to handle as follows. Usage of this limestone low-activated concrete makes almost all concretes satisfy the clearance level for ⁶⁰Co after 20 yr cooling from decommissioning. In respect of ¹⁵²Eu, radioactivation quantity in the biological shielding concrete is reduced up to a half of the average value or less. With regard to ¹³⁴Cs, all concrete satisfies the clearance level.     

Pulsed Neutron Experiment in Fast Natural Uranium Systems

Separation of fission products was carried out by solvent extraction using tri-n-butyl phosphate (TBP). ^{144, 144}Ce, ⁹¹Y and ⁹⁵Zr and ⁹⁵Nb were extracted with TBP from a freshly prepared nitric acid-potassium bromate solution. These nuclides in TBP were successively back-extracted with various aqueous solutions: ^{144, 144}Ce with hydrogen peroxide and nitric acid solution, ⁹¹Y with hydrochloric acid solution, and ⁹⁵Zr-⁹⁵Nb with oxalic acid solution. The other nuclides were not extracted by the TBP and remained in the nitric acid and potassium bromate solutions.     

Shielding performance of newly developed boron-loaded concrete for DT neutrons

In order to enhance the neutron-shielding performance of concrete in neutron-generation facilities from the viewpoint of reduction of effective dose rates in operation and radioactive wastes in decommission, we developed concrete with boron of more than 10 wt%. We performed a neutron-shielding experiment using the mockup of the newly developed boron-loaded concrete and Deuterium Tritium (DT)neutrons at the Fusion Neutronics Source in the Japan Atomic Energy Agency, and measured the reaction rates of ⁹³Nb(n,2n)^92mNb and ¹⁹⁷Au(n,γ)¹⁹⁸Au reactions in the mockup. The calculations were conducted using MCNP-5.14 and FENDL-2.1. The calculation results agreed well with the measured ones, and we confirmed that the accuracy was very good on the atomic composition data of boron-loaded concrete and their nuclear data. In addition, we calculated the neutron and photon effective dose rates and reaction rates of ⁵⁹Co(n,γ)⁶⁰Co and ¹⁵¹Eu(n,γ)¹⁵²Eu reactions, which produce critical radioisotopes in decommission, in boron-loaded concrete and other concretes with DT neutrons, and the experimental condition presently used. The dose rates and reaction rates were drastically reduced by using boron-loaded concrete and it was concluded that boron-loaded concrete had very good shielding performance for DT neutrons.     

Separation of Zirconium-95, 97 and Niobium-95, 97 from Fission Product-Mixtures and Irradiated Uranium by an Extraction Method using 8-Quinolinol

An 8-quinolinol-extraction method has been applied to separate ⁹⁵Zr-⁹⁵Nb or ⁹⁵Nb from fission product mixtures and/or irradiated uranium. Radiochemically pure nuclides can be separated simply and rapidly from either origin. In the separation of ⁹⁵Zr-⁹⁵Nb, the extraction is made from a solution of pH 4.5 to 5.5, containing acetate as masking agent. ⁹⁵Nb is extracted from a solution of pH 4.8 to 5.2, containing fluoride as masking agent, in the presence of uranium. ⁹⁵Nb can be saparated from ⁹⁵Zr and other fission products by extraction from a solution of pH 5.3 to 8, containing EDTA as masking agent. In these cases, after the first extraction, the chloroform solution should be washed with water to remove any remaining extraneous uranium or fission products. The addition of a small quantity of sulfite is required to prevent the extraction of ¹⁰³Ru.

The method can be also used for the separation of the short-lived nuclides, ⁹⁷Zr-⁹⁷Nb or ⁹⁷Nb, from irradiated uranium.     

A new HVE 6 MV AMS system at the University of Cologne

CologneAMS is the new Centre for Accelerator Mass Spectrometry (AMS) at the University of Cologne. It will operate a dedicated AMS system designed to measure all standard cosmogenic nuclides (¹⁰Be,¹⁴C,²⁶Al,³⁶Cl, ⁴¹Ca,¹²⁹I) and which uses a 6 MV Tandetron™ accelerator equipped with an all solid-state power supply, foil and gas stripper. The system also enables a sensitive detection of heavy ions up to ²³⁹U and ²⁴⁴Pu. The high-energy mass-spectrometer consists of a 90 degree magnet with a radius of 2 m and a mass-energy product of 351 AMU MeV to allow the detection of ²⁴⁴Pu⁵⁺ up to the maximum terminal voltage of 6 MV. This magnet is followed by an electrostatic energy analyzer and a switching magnet that can transport the rare isotope beam into various beamlines. The switching magnet forms a third analyzing element which is needed especially for the sensitive detection of heavy elements. So far two beamlines are equipped with their own detection system. One of these lines is used for suppression of isobaric background in the case of the analysis of e.g. ³⁶Cl. This is accomplished by an absorber foil which generates a Z-dependent energy loss in combination with a momentum/charge-state selection via a 120 degree magnet that features up to 30 mrad acceptance for efficient beam transport.In this contribution we will introduce the new Centre, the layout and specific characteristics of the AMS system as well as the main topics of the future scientific work to be performed at CologneAMS.     

The Cost of Transmutation of Fission Products in Nuclear Reactors

The operating regime of a VVÉR reactor in which the most important long-lived fission products ⁹⁹Tc and ¹²⁹I are transmuted is investigated. Estimates are presented for the decrease in the fuel burnup and decrease in the run time as a result of transmutation. Two methods for inserting the nuclides to be transmuted are examined – by adding to the nuclear fuel or the coolant. It is established that ⁹⁹Tc and ¹²⁹I transmutation with the rate of accumulation in a reactor decreases burnup by 5.1 GW·days/metric ton, i.e., by 12.7% of the standard burnup. This corresponds to electricity underproduction 110 GW·days per run or 37 GW·days per year of operation. This result is independent of the method used to insert the nuclides to be transmuted. These energy losses are the price to be paid for transmuting nuclides without removing them during reactor operation.     

Characterization of Solidified Products Yielded by Plasma Melting Treatment of Simulated Non-Metallic Radioactive Wastes

As a part of an advanced volume reduction program for low level radioactive wastes in JAERI, melting tests of non-metallic solid wastes have been conducted with the aim of establishing the optimum melting condition for preparation of a stable solid that is suitable for disposal. Simulated non-metallic wastes containing ⁶⁰Co, ¹³⁷Cs and ¹⁵²Eu tracers were melted with a plasma torch of a non-transferred type, and the examination was carried out on chemical composition and physical properties of solidified products and the distribution of tracers in each product. The products were almost homogeneous and possessed satisfactory mechanical strength for disposal. The radioactive nuclides were almost uniformly distributed in the products. The amount of ¹³⁷Cs remaining in the products depended on the basicity of slag, while that of ⁶⁰Co was not. A small amount of fine metal particles containing ⁶⁰Co was observed in a slag. This was attributable to partial reduction of metal oxides under the relatively reductive atmosphere.     

Radioactivity of fission product and heavy nuclides deposited on soil in Fukushima Dai-Ichi Nuclear Power Plant accident

Based on periodically performed radioactivity measurements on soil samples in the site of Fukushima Dai-Ichi Nuclear Power Station, activity ratios to ¹³⁷Cs of fission product and heavy nuclides were obtained for Sr, Nb, Mo, Tc, Ru, Ag, Te, I, Ba, La, Pu, Am, and Cm isotopes. By exponentially fitting or averaging, the activity ratios at the core shutdown were estimated. Using correlations of activity ratios of ¹³⁴Cs to ¹³⁷Cs, and ²³⁸Pu to the sum of ²³⁹Pu and ²⁴⁰Pu against fuel burnup, burnup of the fuel sourcing the deposited activity of the soil was estimated. The activity ratios to ¹³⁷Cs of each nuclide on the deposited activity were divided by those calculated on the fuel at the shutdown to obtain the deposited activity fraction of each nuclide as a relative value to ¹³⁷Cs, which also corresponds to the deposited fraction of each element as a relative value to Cs. The obtained deposited fractions relative to Cs are the orders of 10^?4 to 10^?2 for Sr, 10^?5 to 10^?3 for Nb, 10^?2 to 10^?1 for Mo, 1 to 10 for I, 10^?3 to 10^?2 for Ba, 10^?2 for La, 10^?6 to 10^?3 for Pu, 10^?6 to 10^?4 for Am, and 10^?7 to 10^?5 for Cm. The deposited fractions for Tc, Ag, and Te were not estimated due to the lack of the calculated inventories in the fuel for the relevant measured radioactive nuclides.     

Nuclear Data Sheets for A = 50

The experimental nuclear spectroscopic data for known nuclides of mass number 50 (Cl,Ar,K,Ca,Sc,Ti,V, Cr,Mn,Fe,Co,Ni) have been evaluated and presented together with Adopted properties for levels and γ rays. This evaluation has been carried out about 15 years after the previous one by Thomas Burrows (1995Bu29). Except for ⁵⁰Sc and ⁵⁰V, extensive new data have become available for all the other nuclides in the intervening years. The data for ⁵⁰Sc and ⁵⁰V have also been checked again in detail and several changes made. No data are yet available for excited states in ⁵⁰Cl, ⁵⁰Ar and ⁵⁰Ni. This work supersedes earlier evaluations (1995Bu29, 1990Bu18, 1984Al29, 1976Au07) of A=50 nuclides.     

Nuclear Data Sheets for A = 163

The available data from the various reaction and decay studies leading to nuclides having mass number A=163 have been reviewed. These data are summarized and presented, together with adopted level schemes and properties, for nuclides from Eu (Z=63) through Os (Z=76). ¹⁶³Eu represents a new addition to these nuclides. Level structures are now known for ¹⁶³Ta and ¹⁶³W through high-spin studies and recoil-decay tagging techniques. Several radioactive decays in this mass chain are not known at all or poorly established. No β or γ-ray data are available for ¹⁶³Re-¹⁶³W-¹⁶³Ta-¹⁶³Hf ε decay chain. Those for ¹⁶³Hf-¹⁶³Lu-¹⁶³Yb ? decay chain and ¹⁶³Eu-¹⁶³Gd-¹⁶³Tb β⁻ decay chain are very sketchy. This evaluation is an update and revision of the previous one (2000Si01).     

Feasibility of Fast Fission System Confining Long-Lived Nuclides

The feasibility of fast fission system confining long-lived nuclides without other supporting system as synergetics for fuel sustainment and waste incineration was studied from the aspects of nuclear material balance and neutron economy. The continuous utilization of fast fission system which confines all actinides in the reactor but discharges all FP will lead to huge accumulation of radioactive wastes such as ¹²⁹I, ¹³⁵Cs, ¹⁰⁷Pd, ⁹³Zr, ⁹⁹Tc, ¹²⁶Sn and ⁷⁹Se in the far future. Then we studied the feasibility of the system that these long-lived seven FP are also confined in the reactor with actinides. In this scheme, all the long-lived nuclides to be disposed of were exposed with neutrons in the reactor and removed as different nuclides after nuclear transmutation. As the wastes stored in the repository was composed of only shorter-lived nuclides, total amount of radioactive wastes in the repository was suppressed to be less than a few tons per 3 GWt reactor.     

Simplified Method for Calculation of Neutron Capture Transformation Effects of Fission Products on Decay Power

A simplified method is proposed for the calculation of the effects of neutron capture transformations of fission products (FPs) on the decay power of FPs. The decay power of FPs after shutdown changes by the neutron capture transformations of FP nuclides during reactor operation. It is proposed to calculate the neutron capture transformation effects considering the production of the following 7 nuclides ¹⁰³Ru, ¹³⁴Cs, ¹³⁶Cs, ^148mPm, ¹⁴⁸Pm, ¹⁵⁴Eu and ¹⁵⁶Eu by the neutron capture reaction of the direct mother nuclide alone giving a cumulative fission yield for the mother nuclide. The present method was assessed by com-paring the calculation results with the rigorous calculation results for the thermal-neutron fission of ²³⁵U irradiated between 1 and 5 yr in a light water reactor with thermal-nentron flux between 3 x 10¹³ and 6 x 10¹³ n/cm²·s and for the fast-neutron fission of ²³⁹Pu irradiated between 1 and 5 yr in a fast breeder reactor with total neutron flux between 3 x 10¹⁵ and 6 x 10¹⁵ n/cm²·s. It has been clarified that the present method can calculate the neutron capture transformation effects within the accuracy of ±1% of the decay power for the irradiation of 1yr and cooling time less than 10⁹s irrespective of fission type and neutron flux. The accuracy varies little with neutron flux but considerably with irradiation time. For a irradiation of 5 yr the present method can calculate the capture effect within the accuracy of +1% and -5% of the decay power. The accuracy can be improved to ±1% of the decay power with the simple correction factors.