Abstract This study describes the use of a neutron irradiator system based on a plutonium-beryllium neutron source for MnSO4 solution activation for use to determine the MSB system efficiency. Computational simulations using Monte Carlo code were performed to establish the main characteristics of the irradiator system. Among the simulated geometries and volumes, semi-cylindrical shape with 84.5 cm3 of MnSO4 solution yielded the best option to be built. Activity measurements were performed with a high-pure germanium detector to validate the new irradiation system. Results showed an average efficiency and uncertainty of the experimental standard deviation of the mean: 5.742 × 10−4 ± 0.036 × 10−4 (coverage factor k = 1), for MSB system. Efficiency value obtained shows good correlation to other published methods (i.e. a relative difference of 0.07%). This alternative metrological method demonstrated the utility of neutron sources for the irradiation of solutions in metrology laboratories providing a cost-efficient alternative to nuclear reactors or particle accelerators. INTRODUCTION Among several difficulties encountered by neutron metrology laboratories, one of them is the availability of a neutron flux intense enough to produce activity levels in a manganese sulfate (MnSO4) solution that are high enough to be used for calibrating a manganese bath system. The manganese sulfate bath (MSB) system(1) is a fundamental method in neutron metrology, which was implemented to measure the neutron emission rate by a neutron source(2). The neutron source to be calibrated is inserted in the center of a bath that contains the MnSO4 solution. By measuring the decay rate of 56Mn, the neutron emission rate of the source can be determined(3). Currently, irradiation of the MnSO4 solution for use in the MSB efficiency measurement is performed at an extramural research reactor site as the facilities required to generate the intensely high neutron flux needed for this activation are not available in-house(4). To mitigate this issue, the group at the Neutron Metrology Laboratory (LN) developed an irradiator system based on a 238PuBe 1.85 TBq (5.75 mm diameter and 3.33 mm length) neutron source to produce the intense flux of neutrons required. It was chosen for having important characteristics such as a long half-life (i.e. ~87.74 years) and high neutron emission rate 8.198 × 107 ± 0.85% cm−2 s−1 on 11 August 2016. Thus, it becomes feasible to discard the use of sources such as 252Cf, whose abundance has dramatically been reduced leading to an increase in market value. In addition, a 252Cf neutron source loses its neutron emission rate in <10 years due to its half-life (i.e. around 2.64 years), requiring frequent replacement of the source and recalibration of the system. The irradiator system described in this article was built as an alternative to nuclear reactors and particle accelerators, seeking to be a refinement of the procedure used in LN and also as a more cost-effective method for radiation efficiency measurements. MATERIAL AND METHODS Computational simulations The initial step in the development of the irradiator system consisted of performing mathematical modeling studies using Monte Carlo MCNPX (version 2.7.0)(5). The modeling compared several geometries and volumes of the MnSO4 1.3939 ± 0.010 g cm−3 to achieve highest 55Mn capture of the manganese solution per neutron. All manganese containers geometries were arranged close to the 238PuBe source, hermetically sealed and positioned in the center of a polymethyl methacrylate (PMMA) polymer cube, filled with ~27 000 cm3 of distilled water (Figure 1). The hydrogen present in water acts as moderator and a reflector medium for neutrons emitted by the 238PuBe source, facilitating their capture by 55Mn nuclei(6). Figure 1. View largeDownload slide Front view of simulated geometries in this study. (A) Semi-cylindrical; (B) cubic; (C) cylindrical and (D) double parallelepiped. Figure 1. View largeDownload slide Front view of simulated geometries in this study. (A) Semi-cylindrical; (B) cubic; (C) cylindrical and (D) double parallelepiped. Semi-cylindrical geometry inputs were configured with the container that has the manganese solution involving the 238PuBe (α, n) source in a semi-cylindrical format (Figure 2). This geometry aims to homogenize the activation of the manganese sulfate solution, eliminating distance points between the neutron source and the solution. Figure 2. View largeDownload slide Detailed views from the constructed irradiator system drawn in AutoCAD™. (A) The drawing on the left shows a 3D view of the open system in the semi-cylindrical geometry. (B) The drawing on the right represents a front view of the irradiator system. Figure 2. View largeDownload slide Detailed views from the constructed irradiator system drawn in AutoCAD™. (A) The drawing on the left shows a 3D view of the open system in the semi-cylindrical geometry. (B) The drawing on the right represents a front view of the irradiator system. The size of the manganese container should allow for sufficient volume of the manganese solution to perform activity measurements and in addition to the MSB system for the decay measurements. As the high-purity germanium (HPGe) method does not require large volumes of solution, the simulations were configured to 84.5, 100 and 169 cm3 for all geometries. All simulations used an MCNP tool named Tally Multiplier Card, or FM, which is a multiplier record associated with a type 4 record (Tally F4). Tally F4 determines the average fluence in the cell under study but does not consider neutron absorption reactions unless it is associated with FM, converting this reaction so that the capture of neutrons by manganese can be determined by the following capture reaction: 55Mn (n, γ)56Mn. The study used the following cross section libraries: endf60, endf66b, endf70a, endf70b, endf70j, tmccs, rmccs and rmccsa(5). Irradiation of MnSO4 solution by the alternative irradiator system The experimental procedure for the use of the irradiator system was developed after selection of the best geometry provided by the Monte Carlo simulations. A certain amount of MnSO4 solution was removed from MSB and inserted into the container of the alternative radiator system to be irradiated by the 238PuBe (α, n) source. Then, the neutron source is introduced in the system to irradiate the solution for 25 h, which is the equivalent time to ~10 half-lives (56Mn decay) required for the solution to reach its saturation point(6). Next, an aliquot is removed from the alternative radiation system and sent to the Spectrometry Laboratory where its activity is measured by a HPGe detector(7). The remainder of the irradiated solution is poured into the MSB so that the efficiency analysis of the MSB system can be determined. Irradiator efficiency calculations Efficiency is calculated using Equation 1, where C represents the average counting rate detected by the MSB system(8), mV is the mass of solution poured into the MSB, mA is the mass of solution used by the Spectrometry Laboratory and Ac is the activity determined by the HPGe method described in Conti et al.(7). ε(C,Ac,mv,mA)=Cmv·mAAc (1) Uncertainty calculations In accordance with Pereira et al.(8), Equation 2 describes the uncertainty for each efficiency measurement for MSB system. [μεε]2=[μCC]2+[−μ.mvmv]2+[−μAcAc]2+[μmAmA]2 (2) where με = efficiency uncertainty; μC = MSB count uncertainty; μmA = uncertainty in mass determination of the irradiated solution sent to Spectrometry Laboratory; μAc = uncertainty in activity determination by the HPGe method corrected to reference time of the MSB; and μmV = uncertainty in mass determination of the irradiated solution poured into the MSB system. RESULTS AND DISCUSSION Geometry simulation results Simulation results are described in Table 1. Testing of the Monte Carlo simulations revealed that the semi-cylindrical geometry produced higher activity values than three other geometries tried. Table 1. Activity values in Bq mg−1 for different geometries and volumes of the irradiator system. 84.5 cm3 (Bq mg−1) 100 cm3 (Bq mg−1) 169 cm3 (Bq mg−1) Semi-cylindrical 5.20 4.93 4.16 Cubic 3.19 3.16 2.88 Cylindrical 3.24 3.15 2.63 Double parallelepiped 2.89 2.88 2.72 84.5 cm3 (Bq mg−1) 100 cm3 (Bq mg−1) 169 cm3 (Bq mg−1) Semi-cylindrical 5.20 4.93 4.16 Cubic 3.19 3.16 2.88 Cylindrical 3.24 3.15 2.63 Double parallelepiped 2.89 2.88 2.72 Table 1. Activity values in Bq mg−1 for different geometries and volumes of the irradiator system. 84.5 cm3 (Bq mg−1) 100 cm3 (Bq mg−1) 169 cm3 (Bq mg−1) Semi-cylindrical 5.20 4.93 4.16 Cubic 3.19 3.16 2.88 Cylindrical 3.24 3.15 2.63 Double parallelepiped 2.89 2.88 2.72 84.5 cm3 (Bq mg−1) 100 cm3 (Bq mg−1) 169 cm3 (Bq mg−1) Semi-cylindrical 5.20 4.93 4.16 Cubic 3.19 3.16 2.88 Cylindrical 3.24 3.15 2.63 Double parallelepiped 2.89 2.88 2.72 Irradiator design requirements The irradiator frame was built with PMMA, due to its properties that minimally interfere in neutrons emitted by the neutron sources and its translucent characteristic, which allows complete visualization of the system helping to identify failures. The wall of the PMMA cube is 10 mm thick and the sides are 30 cm × 30 cm. The system has an internal duct located in the center of the PMMA cube where the 238PuBe source is housed. The semi-cylindrical vessel that contains the manganese sulfate solution was positioned around the duct where the 238PuBe source was placed. The vessel had a volume of 121 cm3 (Figure 3). Figure 3. View largeDownload slide Final construction of the irradiator with the manganese solution and 238PuBe source. Figure 3. View largeDownload slide Final construction of the irradiator with the manganese solution and 238PuBe source. Experimental results The gamma counts from activated 56Mn induced by capture of neutrons in the manganese sulfate solution were acquired using a NaI(Tl) detector with the Genie-2000’s Gamma Acquisition & Analysis™ program, which provided a report with values of decay counts of 56Mn as a function of time (500 s in this study(8)). A comparative analysis of the average efficiency value obtained from triplicate measurements in our MSB system and the average values from work described by Leite et al.(6) is shown in Table 2. Efficiency uncertainty determined with the use of the irradiator developed in this study was 0.63%. The efficiency value obtained in this study compared with the value previously described by Leite et al.(6), which used a 252Cf neutron source, showed a relative difference of 0.07%. Table 2. Efficiency measurements results of the alternative irradiator and comparisons. Average activity Average efficiency of this study Efficiency by Leite (2010) Comparative difference Relative difference (%) Uncertainty Leite (2010) This study 4.46 Bq mg−1 5.742E-04 5.738E-04 −4.05E-07 −0.07 0.36% 0.63% Average activity Average efficiency of this study Efficiency by Leite (2010) Comparative difference Relative difference (%) Uncertainty Leite (2010) This study 4.46 Bq mg−1 5.742E-04 5.738E-04 −4.05E-07 −0.07 0.36% 0.63% Table 2. Efficiency measurements results of the alternative irradiator and comparisons. Average activity Average efficiency of this study Efficiency by Leite (2010) Comparative difference Relative difference (%) Uncertainty Leite (2010) This study 4.46 Bq mg−1 5.742E-04 5.738E-04 −4.05E-07 −0.07 0.36% 0.63% Average activity Average efficiency of this study Efficiency by Leite (2010) Comparative difference Relative difference (%) Uncertainty Leite (2010) This study 4.46 Bq mg−1 5.742E-04 5.738E-04 −4.05E-07 −0.07 0.36% 0.63% Efficiency values for each irradiation performed by the irradiator using a 238PuBe source, average efficiency and reference value are shown in Figure 4. Figure 4. View largeDownload slide Efficiency values for each irradiation performed by the irradiator system with 238PuBe source. Figure 4. View largeDownload slide Efficiency values for each irradiation performed by the irradiator system with 238PuBe source. CONCLUSION This study demonstrates that it is possible to utilize more suitable radioisotope neutron sources such as 238PuBe and still have a reliable measurement system to determine the efficiency of MSB system. Furthermore, the system can replace the use of nuclear reactors or particle accelerators in this type of measurements, reducing the cost of radiation efficiency measurement procedures. REFERENCES 1 Szilard , L. , Anderson , H. L. and Fermi , E. Neutron production and absorption in uranium . Phys. Rev. 56 , 284 – 286 ( 1939 ). Google Scholar CrossRef Search ADS 2 O’Neal , R. D. and Sharff-Goldhaber , G. Determination of absolute neutron intensities . Phys. Rev. 69 , 368 ( 1946 ). Google Scholar CrossRef Search ADS 3 Park , H. , Choi , K.-O. , Lee , I. M. , Lee , K. B. , Hahn , M. S. and Kralik , M. Absolute measurement of the neutron emission rate with a manganese sulphate bath system . J. Korean Phys. Soc. 47 ( 4 ), 603 – 609 ( 2005 ). 4 Roberts , N. J. et al. . International key comparison of measurements of neutron source emission rate (1999-2005)—CCRI(III)-K9.AmBe . Metrologia 48 ( Tech. Suppl. 06018 ), ( 2011 ). 5 Pelowitz , , D. B. (Ed). MCNPX User’s Manual, Version 2.7.0. Los Alamos National Laboratory Report LA-CP-11-00438. pp. 1–645 ( 2011 ). 6 Leite , S. P. , Pereira , W. W. , Silva , A. X. , Fonseca , E. and Patrão , K. C. S. Alternative irradiation system for efficiency manganese bath determination . Radiat. Meas. 45 , 1499 – 1501 ( 2010 ). Google Scholar CrossRef Search ADS 7 Conti , C. C. , Salinas , I. C. P. and Zylberberg , H. A detailed procedure to simulate an HPGe detector with MCNP5 . Prog. Nucl. Energy 66 , 35 – 40 ( 2013 ). Google Scholar CrossRef Search ADS 8 Pereira , W. W. and Leite , S. P. Neutron primary standard metrology. In: Ionizing Radiation Metrology , 1st edn . (Brazil: Ionizing Radiation Metrology) pp. 103 – 115 ( 2016 ). © The Author(s) 2018. Published by Oxford University Press. All rights reserved. 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Radiation Protection Dosimetry – Oxford University Press
Published: Aug 1, 2018
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