Neutron Spectrometry in Mixed Fields: Characterisation of the RA-1 Reactor Workplace

Transcription

1 Neutron Spectrometry in Mixed Fields: Characterisation of the RA-1 Reactor Workplace Gregori, B.N.; Carelli, J.L; Cruzate, J.A.; Papadópulos, S. and Kunst, J.J. Presentado en: Second European of IRPA (International Radiation Protection Association). París, Francia, mayo 2006

2

3 Neutron Spectrometry in Mixed Fields: Characterisation of the RA-1 Reactor Workplace B. Gregori; J. Carelli; J.A. Cruzate; S. Papadópulos; Kunst, J.J. Autoridad Regulatoria Nuclear Argentina The characterisation of the neutron spectra in workplaces of the Argentine reactor No.1 (RA-1) has been carried out by using a neutron spectrometric system based on a set of moderated spheres with 3 He detector. The neutron response matrix was calculated using the MCNPX 2.5e code and ENDF/B-VI library in the energy range between thermal neutron and 100 MeV. The neutron spectrum was unfolded by using the UMG suit code. The validation of the spectrometric system was performed at Commissariat à l Energie Atomique, (CEA)-Cadarache (France) with of 252 Cf and AmBe sources. The spectral fluence of the field in the selected points of the facility (RA-1) is presented. The ambient dose equivalent -H*(10)- and the personal dose equivalent -Hp(10)- have been derived from the neutron fluence, by applying ICRP-74 recommended fluence to dose conversion factors. Introduction The characterisation of the neutron spectrum of a workplace is an essential dosimetric tool to improve the assessment of the personal equivalent dose and the ambient dose equivalent [1]. In addition, if the operational conditions of the facility are well defined, the set of field spectra obtained may be used as a reference to compare the performance of different types of neutron detectors [2-3]. Despite of difficulty of doing both measurements at the same time, the irradiations conditions can be evaluated by independent methods such as the control of the reactor power, the identical material distribution in the place, and to assure the same measurement position. Neutron survey meters and personal dosimeters calibrated using reference fields such as 252 Cf and AmBe sources do not always yield correct measurements in a workplace neutron field if this spectrum is significantly different from that of the radiation reference. The aim of this work is to present the characterization of the workplace neutron spectra of the RA-1 research reactor facility applying the Bonner Sphere Spectrometry System (BSS) and the LOHUI82 [4], GRAVEL [5] and MAXED [6] unfolding codes. The ambient dose equivalent and the personal dose equivalent are calculated using the ICRP74 [7] conversion factors. The results of the spectrometric system s validation carried out CEA- Cadarache (France) with of 252 Cf and AmBe sources are also shown. Materials and Methods The Bonner sphere spectrometric system (BSS) The spectrometric system used is composed of nine polyethylene moderating spheres (density range: to g.cm -3 ) with diameters ranging from 3 to 12 inches. The laboratory has designed its own set of spheres and commissioned its fabrication. The calibration of the BSS s sphericity and density were performed at the Instituto Nacional de Tecnología Industrial (INTI). The thermal neutron detector, 3 He-filled spherical SP9 type proportional counter with 1.25 diameter and 400 kpa nominal pressure (Centronic), is placed in the centre of each sphere. The electronic chain is used to acquire and analyse the output spectrum. The count rate M i from the sphere (i) in a given neutron field is obtained by integration of the product of the spheres response function F i (E) with the neutron fluence spectral distribution φ E (E). The use of a

4 number of spheres of different diameters in an unknown neutron field leads to the following set of equations: E E ( E) Φ E ( E) de i = 1,... Nd max M i = Fi, min where N d is the number of spheres. The resolution of the system with respect to φ E (E) was performed by applying the following unfolding codes: LOUHI and UMG suit (MAXED and GRAVEL). The dosimetric quantities, Hp(10) and H*(10) were calculated from the neutron spectral distributions according to ICRP74. In the case of MAXED solutions, the uncertainty of both these values and the solution spectrum were calculated. The BSS matrix response The BSS matrix response was performed applying MCNPX 2.5e code [8] with the cross sections library ENDF/B-VI [9]. The response was calculated for 312 monoenergetic energies in the range of thermal neutron to 100 MeV in parallel beam geometry. The simulation was done for the commercial detector used in the system; neither air gap between the detector and the sphere nor additional Krypton gas was considered. The density s variation of the spheres was included in the calculation. The BSS verification The BSS was irradiated in CEA - Cadarache with 252 Cf and AmBe sources. The sources were placed at 75 cm from the sphere centre and at 3.2 m from the floor. The shadow cone technique was used in order to account for the scattered radiation. Each sphere-detector set, the bare detector and the bare detector under Cd filter, were irradiated five times, for a time period long enough to get very low uncertainties (<2%). The measurements of bare detector and bare detector under Cd filter were not considered because of its low statistics. The experimental irradiation conditions of the BSS, with the real pressure value (128 kpa) [10], were simulated with the MCNPX 2.5e code. These results agree well with the BSS measurements carried out for the aforementioned radiation sources. The relationship in the response and the uncertainty obtained, considering the complete BSS, are: AmBe, 0.99 ± 14% (N=9, k=2), 252 Cf, 1.03 ± 10% (N=9, k=2). where N is the number of spheres used and k is the coverage factor. The standard deviations on the dosimetric quantities were less than 10% for both sources.

5 Figure 1 presents the comparison between the ISO 8529 [11] and the unfolded spectrum for AmBe respectively. Neutron Fluence rate per lethargy [cm -2 s -1 ] AmBe Energy [MeV] Figure 1. The experimental neutron spectrum (dotted line) unfolded by MAXED code, and the ISO standard spectrum for AmBe (line) as function of neutron energy. Figure 2 shows the comparison between the ISO 8529 and the unfolded spectrum for 252 Cf respectively. Neutron Fluence rate per lethargy [cm -2 s -1 ] Cf Energy [MeV] Figure 2. The experimental neutron spectrum (dotted line) unfolded by MAXED code, and the ISO standard spectrum for 252 Cf (line) as function of neutron energy. The measured workplace fields The measurements described in this work have been performed in Argentine research reactor RA-1. This open tank reactor works at 40kW with 20% enriched uranium fuel. The two measurement points selected are located inside the containment, in areas only used by personnel occupationally exposed. The points have been chosen in order to have a representative set of different typical neutron spectra at the workplace of this facility.

6 The detector s centre, inside each sphere of the BSS, was placed at the same marked position, 1 m from the floor. At each point, measurements have been performed with different commercial neutron monitors. The records of the control room have been evaluated and the power was practically constant. Results and Discussion The neutron spectra obtained in the present work are shown in Figure 3. Neutron Fluence rate per lethargy [cm -2 s -1 ] Energy [MeV] Figure 3. Neutron spectra distribution for the point 6 (line) and 12 (dotted line) of the RA-1 reactor facility, by applying MAXED unfolding code as function of neutron energy. The mean energy is 0.28 MeV and 0.13 MeV for points 6 and 12 respectively. The quantities evaluated, using three different unfolding codes are presented in Table 1. Position Unfolding code Fluence rate [n/s] H*(10) [µsv/h] Hp(10) [µsv/h] point 6 LOUHI GRAVEL MAXED 877±19% 173±38% 182±37% point 12 LOUHI GRAVEL MAXED 544±20% 73±40% 77±39% Table 1: The dosimetric quantities (fluence rate, ambient equivalent dose rate and personal equivalent dose rate) unfolded by different codes (LOUHI, MAXED and GRAVEL) for the point 6 and point 12 of the RA-1 reactor facility. As may be observed, the maximum difference due to the unfolding codes in the fluence rate and in the ambient dose equivalent rate is less than 20%, meanwhile in the personal dose equivalent rate is less than 40%. This value of the discrepancy depends on the selection of the guess spectrum used in the first step for iterative adjustment in unfolding codes (1/Energy, in our case) and agrees with results from other authors [12].

7 Conclusions Investigations of dosimetric characteristics of the RA-1 neutron fields have been carried out by BSS measurements. Three unfolding codes have been applied to the neutron measurements. The dosimetric quantities have been calculated with uncertainties less than 40%, which are low according to radiation protection requirements. The BSS has been tested in well-known neutron fields. The overall results obtained with the AmBe and 252 Cf are satisfactory. The neutron spectra unfolded from the spheres measurements agree well with ISO standards. Acknowledgment The authors are grateful to Dr H. Muller and Dr T. Bolognese of the Institut de Protection et Sûreté Nucléaire, Commissariat à l Energie Atomique, Cadarache (France) for the irradiation and to the working group team at RA-1 reactor of Comisión Nacional de Energía Atómica (Argentina) for their cooperation. References [1] Aroua, A.; Boschung, M.; Cartier, F.; Grecescu, M.; Pretes, S.; Valley, J.F.; Wernli, Ch. Characterisation of the Mixed Neutron-Gamma Fields inside the Swiss Nuclear Power Plants by Different Active Systems. Radiat. Prot. Dosim. 51 (1) 17-25(1994). [2] Chartier, J.L.; Posny, F.; Buxerolle, M.; Experimental Assembly for the Simulation of Realistic Neutron Spectra. Radiat. Prot. Dosim. 44 (1-4) (1992). [3] Chartier, J.L.; Jansky, B.; Kluge, H.; Schraube, H.; Wiegel, B. Recent Development in the Specification and Achievement of Realistic Neutron Calibration Fields. Radiat. Prot. Dosim. 70 (1-4) (1997). [4] J. T. Routti and J. V. Sandberg, General purpose unfolding program LOUHI78 with linear and nonlinear regularizations, Comput. Phs. Commun, 21(1980) [5] M. Reginatto, P. Goldhagen. and S. Neumann, "Spectrum unfolding, sensitivity analysis and propagation of uncertainties with the maximum entropy deconvolution code MAXED", Nucl. Instr and Meth. A 476,242 (2002) [6] M. Matzke, "Unfolding of Pulse Height Spectra: The HEPRO Program System", Report PTB-N-19, October [7] International Commission on Radiological Protection. Conversion Coefficients for Use in Radiological Protection against External Radiation. ICRP Publication 74 (Oxford: Pergamon Press) (1997). [8] Los Alamos National Laboratory: TPO-E83-G-UG-X MCNPX TM User s manual. Version Laurie S. Waters Editor. (1999). [9] Evaluated Nuclear Data File (ENDF), National Nuclear Data Center, Brookhaven National Laboratory. [10] Internal Communication. National Regulatory Authority (ARN). Buenos Aires [11] ISO Neutron Reference Radiations for Calibrating Neutron Measuring Devices Used for Radiation Protection Purposes and for Determining their Response as a Function of Neutron Energy. International Standard ISO :1998(E) International Organization for Standardization, Geneva, Switzerland (1998). [12] Alevra, A. V. and Thomas, D. J., Radiat. Prot. Dosim. 107 (1-3) (2003).