monte carlo simulations of the pulse-height response function of ...

Zbornik radova 52. Konferencije za ETRAN, Palić, 8-12. juna 2008. Proc. 52nd ETRAN Conference, Palić, June 8-12, 2008

MONTE CARLO SIMULATIONS OF THE PULSE-HEIGHT RESPONSE FUNCTION OF GERMANIUM DETECTOR Milijana Steljić, Miodrag Milošević, Petar Beličev Centre for Nuclear Technologies and Research, Vinča Institute of Nuclear Sciences P.O. Box 522, 11001 Belgrade, Serbia Abstract – This paper describes a Monte Carlo simulations used to calculate the pulse-height response function of coaxial germanium detector for photon energies from 0.1 MeV to 1.7 MeV. The calculations allow the uncertainty estimation due to inadequate specifications of source positioning and to variations in the detector’s physical components. A detailed Monte Carlo model was developed using the MCNP-4C code. This analysis has indicated that Monte Carlo methods can represent a valuable tool for the quantitative uncertainty analysis of radiation spectrometers. They can furthermore be instrumental in the gamma-ray detector quantitative calibration and benchmarking processes, thus minimising the need for deployment of radioactive sources.

1. INTRODUCTION Depending on the application for germanium detectors in gamma ray spectrometry, knowledge of peak efficiency over a specific energy range and the response function (or pulse height distribution) of the detector are important [1]. For optimal efficiency calibration of detectors, radioactive standard sources with known properties are being used. Furthermore, multinuclide standard sources with gamma rays in the energy range of interest are used to compose an energy-dependent efficiency curve by fitting a function through the calibration points. While these methods are very reliable, they are also time and money consuming, ultimately increasing the amount of radioactive waste. The modern mathematical calibration software [2,3], based on the well-known MCNP Monte Carlo modelling code [4], brings a new level of capabilities to germanium gamma sample assay by eliminating the need for traditional calibration sources during the efficiency calibration process. These simulations have indicated good agreement when compared to empirical efficiency of Ge detectors and were validated for efficiency calculations of various sourcedetector geometries and source shapes [5]. They were also applied in construction of various detector models for efficiency interpolation in gamma ray spectrometry [6]. In this paper the Monte Carlo method was used to simulate the pulse-height response function of coaxial highprecision germanium (HPGe) detector for photon energies from 100 keV to 1.7 MeV. The calculations based on the detailed geometry model of germanium detector and the MCNP-4C Monte Carlo code allow the uncertainty estimation due to inadequate specifications of source positioning and to variations in the spectrometer angular positioning.

2. MONTE CARLO SIMULATIONS The specifications given by manufacturer on the physical dimensions of the detector’s crystal and position are frequently insufficient. To perform the proper calibrations using the Monte Carlo method, all the components have to be defined accurately. The estimated uncertainties due to improper specifications in detector dead layer thickness, source positioning, angular position of the spectrometer and the absence of aluminium holder and their effects on detector response, should be assessed. We developed a model to predict the uncertainty due to inadequate specifications of the detector’s properties, using the Monte Carlo radiation transport code MCNP-4C applying the standard MCNP pulse-height (energy deposition) estimates. The described MCNP model is results of the uncertainty estimation due to the model’s geometrical variations. The complete set-up of the portable Canberra coaxial Ge detector (Figure 1) presented in this work, is schematically represented in Figure 2. It was found that germanium detector crystal is a cylinder with 40 mm height and 50 mm diameter, including the cylindrical hole with 12 mm diameter and 30 mm height. The Ge crystal is located into an aluminium cylindrical holder with inner diameter of 50 mm and thickness of 1 mm. The cryostat (Al cylinder) has a thickness of 1.5 mm, and external diameter of 70 mm. The system is supplied with 50 mm lead collimator. The standard MCNP Gaussian broadening of the pulseheight response was used to define the FWHM (MeV) of the Ge crystal and to obtain a realistic spectrum FWHM(E)=0.00102828+0.00148606 E , (E in MeV)

The parameterised fitting function FWHM(E) represents an experimental fit of three points with ΔE1 =0.00145 MeV at 0.0810 MeV, ΔE 2 =0.00224 MeV at 0.6617 MeV and ΔE 3 =0.00274 MeV at 1.3325 MeV. The validation of this model is carried out by comparison of the calculated activities with the known activities of calibration sources 133 Ba, 137 Cs and 60 Co . The precision of the obtained results, for the energy range from 0.1 MeV to 1.7 MeV, is better than ±3%. The measured and calculated energy spectra of standard calibration gamma rays sources are presented in Figures 3, 4 and 5. All simulations were executed on a Pentium IV 3.2 GHz processor taking ~2 days of CPU time on the Linux cluster of California University at Berkeley. At least 2⋅109 histories were processed to obtain the statistical uncertainty in the peaks

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of 137 Cs and 60 Co pulse-height response below 0.3%.

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Fig. 4. Comparison of measured and calculated energy spectra of 137 Cs (137 m Ba ) source Fig. 1. Photo of the mobile coaxial Ge detector

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Fig. 5. Comparison of measured and calculated energy spectra of 60 Co source

Fig. 2. A schematic presentation of MCNP geometry, used to model the standard calibration sources (horizontal cross section) 5000 4500 4000

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Fig. 3. Comparison of measured and calculated energy spectra of 133 Ba source

3. RESULTS

Described MCNP geometry model was used for calibration of coaxial Ge detector for activity measurements of small bottles with water samples taken from the water pools or stainless steel containers with the RA reactor spent fuel elements. Due to different specific activity of the samples, each one was measured at various distances from detector base selected to preserve the detector dead time less than 1%. The efficiencies of the Ge detector for each nuclide in the sample were determined with the MCNP-4C simulation. The MCNP geometry model of Ge detector used to simulate the response function of samples with the water from the pools of the RA reactor spent fuel storage, is presented in Figure 6. The efficiency curves obtained for gamma rays emitted from 137 Cs and 60 Co in water samples are presented in Figure 7. The results of water samples activity measurements, given in Table I, show a good agreement with the referent (Extended Range) Ge detector GX5020 calibrated with the LabSOCS Calibration Software.

4. CONCLUSION

Having validated the MCNP-4c model of the portable Ge detector on the standard calibration sources, we used it to calculate the efficiencies for samples taken from the pools of the RA reactor spent fuel storage. A good agreement with standard calibration sources activities as well as with the results obtained by the referent Ge detector GX5020 for the water samples indicated that this model is valuable tool for numerical calibration of presented portable Ge detector.

ACKNOWLEDGEMENT

Fig. 6. The MCNP geometry of germanium detector used to model the response of the water sample (vertical cross section) Results for distance from detector base equal to 10.9 cm 8.0 7.5 7.0 -4

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This work was funded by the Ministry of Science and Environmental Protection of Serbia in frame of the Vinča Institute Nuclear Decommissioning Programme (VIND) under the VIND project "Safe Removal of Spent Fuel of the RA Reactor", and supported by the International Atomic Energy Agency (IAEA) under the IAEA-TC project No. SCG/4/003 "Safe Removal of Spent Fuel of the Vinča RA Research Reactor". The authors are grateful to the IEAE Department of Safeguards for permission to use their multichannel analyser MCA166 and to professors E. Greenspan and J. Vujić from the University of California at Berkeley (USA) for allowance to use the MCNP-4C code at the Berkeley Linux Clusters.

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REFERENCES

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Fig. 7. The efficiency of coaxial Ge detector as a function of weight of the water sample

Table I. Results of water samples activity measurements

Water sample from Pool 1 Pool 2 Pool 3 Pool 4 Channel in Room-141 Channel in Reactor hall

Activity [Bq⋅ml-1] Referent Ge Portable Ge detector detector 116.3 114.8 (± 3%) (±4.7%) 118.7 115.0 (± 3%) (± 4.7%) 118.8 112.5 (± 3%) (± 4.6%) 121.2 115.1 (± 3%) (± 4.7%) 114.0 119.1 (± 4.7%) (± 3%) 126.5 119.5 (± 5.1%) (± 3%)

[1] G.F. Knoll, Radiation Detection and Measurement, Third Ed., Wiley, New York, 2000. [2] F. Bronson, ″Ge Gamma Spectroscopy Characterisation Tools for Contaminated Materials in Buildings, Boxes and Dirt,″ American Nuclear Society and European Nuclear Society International Winter Meting, Washington, DC (United States), November 10-14, 1996. [3] F. Bronson, B. Young, V. Atraskevich, ″ISOCS Mathematical Calibration Software for germanium Gamma Spectrometry of Small and large Objects,″ American Nuclear Society Annual Meeting,. Boston, MA (United States), June 26, 1997. [4] J.F. Briesmeister, Editor, ″MCNPTM – A General Monte Carlo N-Particle Transport Code, Version 4C,″ LA– 13709–M Report, Los Alamos National Laboratory, April 2000. [5] F.L. Bronson, L. Wang, ″Validation of the MCNP Monte Carlo code for germanium detector gamma efficiency calibrations,″ Proceedings of the Conference on Waste Management, Tucson, AZ (United States), February 28, 1996. [6] P. Maleka, M. Maučec, ″Evaluation of Problem 7: Peak Efficiencies and Pulse Height Distribution of a Photon Germanium Spectrometer in the Energy Range below MeV,″ Report to QUADOS Collaboration, Kernfysisch Versneller Instituut, Groningen, The Netherlands, 2003.